Text

Proposed Tech Specs Re Administrative Controls
	ML20065L497
Person / Time
Site:	Grand Gulf
Issue date:	04/15/1994
From:	; ENTERGY OPERATIONS, INC.
To:	;
Shared Package
ML20065L496	List: ML20065L492; ML20065L497;
References
	GNRO-94-00056, GNRO-94-56, NUDOCS 9404210109
	Download: ML20065L497 (183)
	v • d • e

-. - -

.n';:,rr-a 5

GRAND GULF CHAPTER 1

Revised PSTS Pages 4

9404210109 940415 ADOCK0500g6 DR

Enc'ssure 2 to GNRO-94/00050 Definitions Chapter 1 Page 2 1.1 Definitions

'l CORE ALTERATION monitors, traversing incore probes, or special (continued) movable detectors (including undervessel replacement) is not considered a CORE ALTERATION.

In addition, control rod movement with other than the normal control rod drive is not considered a CORE ALTERATION provided there are no fuel assemblies in the associated core cell.

Suspension of CORE ALTERATIONS shall not preclude completion of movement of a component to a safe position.

)

CORE OPERATING LIMITS The COLR is the unit specific document that REPORT (COLR) provides cycle specific parameter limits for the current reload cycle.

These cycle specific limits

{

shall be detennined for each reload cy l operation within these limits isb.1.0.g e in 1

accordance with Specification 5.^

Plant ddre ed in pc ot-individual Spectfications, 1

r3,g,5 gj/on R1 DOSE EQUIVALENT I-131 DOSE EQUIVALENT I-131 shall be that concentration of I-131 (microcuries/gro) that alone would produce the same thyroid dose as the quantity and i

isotopic mixture of I-13'i, I-132, I-133, I-134, and I-135 actually present.

The thyroid dose i

conversion factors used for this calculation shall be those listed in { Table III of TID-14844, h

AEC, 1962, " Calculation of Distance Factors for Power and Test Reactor Sites' Or thc;c listcd in '

Thic E-7 of lit NRC,1977T." gulatcry Cuide 1.100, "e.

1, o 8-AVERAGE E shall be the average (weighted in proportion DIS 1NTEGRATION ENERGY to the concentration of each radionuclide in the reactor coolant at the time of sampling) of the g

Q disintegration (ge beta and ganna energies per

'p sum of the avera in MeV) for isotopes, other than iodines, with half lives > [15] minutes, making up at least 95% of the total noniodine activity in the coolant.

EMERGENCY CORE COOLING The ECCS RESPONSE TIME shall be that time interval SYSTEM (ECCS) RESPONSE from when the monitored parameter exceeds its ECCS TIME initiation setpoint at the channel sensor until the ECCS equipment is capable of performing its safety function (i.e., the valves travel to their (continued)

BWR/6 STS 1.1-3 Rev.

O, 09/28/92

Definitions 1.1 1.1 Definitions q/

k1 SNI.UU.7" PRESSURE AND shall be determined for each I ce eriod in TEMPERATURE LIMITS accordance with Specification 1 '.

Plant

^

REPORT (PTLR) operation within these operating imits is (continued) addressed in LCO 3.4.11, "RCS Pressure and Temperature (P/T) Limits."

RATED THERMAL POWER RTP shall be a total reactor c re heat transfer (RTP) h rate to the reactor coolant o

[3833fMWt.

REACTOR PROTECTION The RPS RESPONSE TIME shall be that time interval SYSTEM (RPS) RESPONSE from when the monitored parameter exceeds its RPS TIME trip setpoint at the channel sensor until de-energization of the scram pilot valve solenoids.

The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured.

SHUTDOWN MARGIN (SDM)

SDM shall be the. amount of reactivity by which the reactor is suberitical or would be subcrit3 cal assuming that:

a.

The reactor is xenon free; b.

The moderator temperature is 68'F; and All control rods are fully inserted except for c.

the single control rod of highest reactivity worth, which is assumed to be fully withdrawn.

With control rods not capable of being fully Qcq inserted, the reactivity worth of these control rods must be accounted for in the determination of SDN.

--->l STAGGERED TEST BASIS A STAGGERED TEST BASIS shall consist of the testing of one of the systems, subsystems, channels, or other designated components during the interval specified by the Surveillance Frequency, so that all systems, subsystems, channels, or other designated components are tested during n Surveillance Frequency intervals, where n is the total number of systems, subsystems, channels, or other designated components in the associated function.

(continued)

BWR/6 STS 1.1-7 Rev.

O, 09/28/92

, _ ~ _.

Enclosura 2 to Ct4RO-94@056.,

Chapter 2 Page 1 i

GRAND GULF CHAPTER 2 Revised PSTS Pages e

Enclosurs 2 to GNRO-94/000M

$($

Chaptst2 Page 2 20 2.0 SAFETY LIMITS (SLs) 2.1 SLs 2.1.1 Reactor Core SLs 2.1.1.1 With the reactor steam dome pressure < 785 psig or core flow < 10% rated core flow:

THERMAL POWER shall be s 25% RTP.

2.1.1.2 With the reactor steam done pressure e 785 psig and core flow e 10% rated core flow:

I c<o MCPR shall be a [+,rei for two loo recirculation OOI operationora{t:;4&}@.or single c

operation.

l.o')

oop recirculationf 2.1.1.3 Reactor vessel water level shall be greater than the top of active irradiated fuel.

2.1.2 Rtictor Coolant Systes Prerrure St.

h Reactor steam dome pressure shall be s 1325 psig.

2.2 SL Violations i

With any SL violation, the following actions shall be completed:

2.2.1 Within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , notify the NRC Operations Center, in accordance with 10 CFR 50.72.

2.2.2 Within 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months s:

2.2.2.1 Restore compliance with all SLs; and 2.2.2.2 Insert all insertable control rods.

Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , notify thkcener[hanagerE"ur!e 2.2.3 W

G;; rioid;nt T;;;;;r 0;n~--=} and theg[tffsi te...,;_;:r: A

!?00f#f?d 4" S?::ific.ii;; 5.5.2, "[0ff:it: P:';ier&ndAudii.-]."-

Nb cor po qkt. be.bW 2

r 93/tso R I cup @oe % ge#\\

f\\n n a c \\ g f *'t (continued)

BWR/6 STS 2.0-1 Rev.

O, 09/28/92

. - _. -. _. - __---_ _ -___-_--___--_ _ to GNRO-94/00050 Chapter 2 Page 3 5Ls-2.0 2.0 SLs 2.2 SL Violations (continued) 2.2.4 Within 30 days, a Licensee Event Report (LER) shall be prepared PC A Jursuant to 10 CFR 50.73.

The LER shall be submitted to the NRC, q g ig gi the t - :i z :.::_;r:; stcr"t ed 5;;;inceden ;.5.2]. esiG the

'a h "

Gen;5[QltinagerC.f :S an and M c

= f :::

r ::a?

"" ~

c g f o m %. s.* e_c.v W s su po n s'. 'bk 'y g oe.c o,\\\\ plen; noe.br-Mdy.

2.2.5 Operation of the unit shall not be resumed until authorized by the NRC.

I l

BWR/6 STS 2.0-2 Rev.

O, 09/28/92 i

ra.nne."t-j j

GRAND GULF SECTION 3.0 Revised PSTS Pages

)

gfy,g'*gg i " LCO Appiicabil 3.0 LCO APPLICA8ILITY be specifiedconditionsintheApplicabilitythatarereaut@redt LC0 3.0.4 (continued) to comply with ACTIONS th j

s]4 p c5 % sLMc, Exceptions to this Specification are stated in the individual Specifications.

These exceptions allow entry into MODES or other specified conditions in the

-l 3

Applicability when the associated ACTIONS to be entered allow unit operation in the MODE or other specified condition in the Applicability only for a limited period of time.

l LCO 3.0.5 Equipment removed from service or declared inoperable to comply with ACTIONS may be returned to service under administrative control solely to perform testing required to demonstrate its OPERA 8ILITY or the OPERABILITY of other equipment. This is an exception to LC0 3.0.2 for the system returned to service under administrative control to perform the testing reqvfred to demonstrate OPOtABILITY.

LCO 3.0.6 When a supported systes LC0 is not met solely _ due to a support system LC0 not being met, the Conditions and a

Required Actions associated with this supported system are not required to be entered. Only the support system LC0 ACTIONS are required to be entered.

This is an exception to.

LC0 3.0.2 for the supported' system.

In this event, 5'. 5'q additional evaluations and limitations may be required in accordance with $pecification

" Safety Function Detensination Program (SFDP)."@If a loss of safety funct i

is determined to exist by this program, the appropriate PCo L Conditions'and Required Actions of the LC0 in which the' loss 63)li Al of safety function exists are required to be entered.

When a support system's Required Action directs a supported system to be declared inoperable or directs entry into' Conditions and Required Actions for a supported system,'the applicable Conditions and Required Actions shall be entered in accordance with LC0 3.0.2.

(continued)

BWR/6 STS 3.0-2 Rev.

O, 09/28/92

$$pI.EIP e

LCO Applicability B 3.0 BASES

@CO L

.0.6 declared inoperable or direct entry into Conditions and (continued)

Required Actions for the supported system.

This may occur imediately or after some specified delay to perfom some other Required Action.

Regardless of whether it is immediate or after some delay, when a support system's Required Action directs a supported system to be declared inoperable or directs entry into Conditions and Required Actions for a supported system, the applicable Conditions and Required Actions shall be entered _in accordance with LCO 3.0.2.

g cg,,,,

Specification

" Safety Function Detemination Program" (SFDP), ensures loss of safety function is detected and appropriate actions are taken.

Upon failure to meet two or more LCOs concurrently, an evaluation shall be made to detemine if loss of safety function exists. Additionally, other limitations, remedial actions, or compensatory actions may be identified as a result of the support system inoperability and corresponding exception to entering supported system Conditions and Required Actions.

The SFDP implements the requirements of LCO 3.0.6.

Cross division checks to identify a loss of safety function for those support systems that support safety systems are required.

The cross division check verifies that the supported systems of the redundant OPERABLE support system are OPERABLE, thereby ensuring safety function is retained.

If this evaluation determines that a loss of safety function exists, the appropriate Conditions and Required Actions of the LCO in which the loss of safety function exists are required to be entered.

LC0 3.0.7 There are certain special tests and operations required to be perfomed at various times over the life of the unit.

These special tests and operations are necessary to demonstrate select unit performance characteristics, to perform special maintenance activities, and to perform special evolutions.

Special Operations LCOs in Section 3.10 allow specified TS requirements to be changed to permit performances of these special tests and operations, which otherwise could not be performed if required to comply with the requirements of these TS.

Unless otherwise specified, all the other TS requirements remain unchanged.

This will (continued)

BWR/6 STS B 3.0-8 Rev.

O, 09/28/92

. =

Endmure 2 to GNRO-94/C0056 Chapter 3.3 Page 1 GRAND GULF i

SECTION 3.3 l

Revised PSTS Pages i

I

Enebsura 2 to GNRO-94/CD056 Chapter 3.3 Pap 2 PAM Instrumentation 3.3.3.1 3.3 INSTRUMENTATION 3.3.3.1 Post Accident Monitoring (PAM) Instrumentation LCO 3.3.3.1 The PAM instrumentation for each Function in Table 3.3.3.1 1 shall be OPERABLE.

APPLIC.BILITY:

MODES 1 and 2.

A ACTIONS

..................................---N0TES--..-.--.......--..-................

I.

LCO 3.0.4 is not applicable.

2.

Separate Condition entry is allowed for each Function.

CONDITION REQUIRED ACTION COMPLETION TIME A.

One or more Functions A.1 Restore required 30 days with one required channel to OPERABLE channel inoperable.

status.

B.

Required Action and 0.1 Initiate act g Immediately associated Completion enec u

Time of Condition A Specifica 6 fP" ed not met.

5.0.2..

sobmh -

Qu:j. o't @Pof

. NOTE..-...)

C.

C.1 Restore one required 7 days No appi ble t channel to OPERABLE hy en itor]

status.

anne

.........]

One or more Functions with two required channels inoperable.

(continued)

BWR/6 STS 3.3-18 Rev.

O, 09/28/92

$$pI.", lins"."I" 2

PAM Instrumentation 3.3.3.1 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME i

'\\iperable.'03 Two [refu' ed hydrog D.1 Restbreone[requireh 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Ob nitor ch nels hydroge nitor]

\\

channel to PERABLE N

status.

l

.) Required Action and Enter the Condition Imediately 9

associated Completion referenced in

(

Time of Condition C Table 3.3.3.1-1 for

- % W not met, the channel.

7

\\

. As required by 1)

Be in MODE 3.

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Required Action E.1 and referenced in Table 3.3.3.1-1.

h

%. As required by D

Initiate lamediately RequiredAction11 QL -.

fSp:i#heuv7

-SL-SL-and referenced in Table 3.3.3.1-1.

Q.".25 PCsL.

<13/o JI R 1 9"

~

%%m% a &&

h d.

f BWR/6 STS 3.3-19 Rev.

O, 09/28/92

Encbsuro 2 to GNRO 94@056 Chapter 3.3 Pags 4 PAM Instrumentation B 3.3.3.1 BASES ACTIONS function of the instruments, the operator's ability to (continued) diagnose an accident using alternate instruments and methods, and the low probability of an event requiring these instruutents.

OCq A Note has also been provided to modify the ACTIONS related to PAM instrumentation channels.

Section 1.3, Completion Times, specifies that once a Condition has been entered, MM56ns.

subsequenf 6, subsystems, components, or variables expressed in the Condition, discovered to be inoperable or not within limits, will not result in separate entry into the Condition. Section 1.3 also specifies that Required i

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 i

inoperable PAM instrumentation channels provide appropriate compensatory measures for separate inoperable functions. As i

such, a Note has been provided that allows separate Condition entry for each inoperable PAM Function.

6d When one or more Functions have one required channel that is inoperable, the required inoperable channel must be restored to OPERABLE status within 30 days.

The 30 day Completion Time is based on operat ience and takes into account h

the remaining OPERABLE hanne in the case of a Function that has only one require

, other non-Regulatory Guide 1.97 instrument channels to monitor the Function), the passive nature of the instrument (no critical automatic action is assumed to occur from these instruments), and the low probability of an event requiring PAM instrumentation during this interval.

gchnd h U

P Co l P"PM If a channel has not been restored to OPERABLE status in 43/0##

d 5hN a 30 days, this Required Action specifies initiation of RI ne-g i 5pWei OkpN-I'r ' -",;" n ' :-" J. "; '". fn!!!"""" """" *JE + h* \\'

_..+s...

a......,s.....m e----

-n

'" " - ~<

/.

Mm.. :.:' _m. m"m_. n..:.:,,, ".' '..:-';...; m.. : *. "to the NRC.

ra This ro report discusses the results of the root cause evaluation of the inoperability and identifies proposed restorative L_ Action is appropriate in lieu of a shutdown This actions.

k %. < PRc'4 \\ A <p.t-Y A ll Q g l., N-Ph)

'd kV a cc ck h e a J 4k Ib c F R Gb, t-l QWm _N Mw5 J echc 's n g CAN! n d i

BWR/6 STS B 3.3-59 Rev.

u, 0 8/92

Encbsurs 2 to GNRO-94/00056 Chxpttr 3.3 Paga 5 PAM Instrumentation B 3.3.3.1 BASES Q:0 ACTIONS (continued)

Condition referenced in the Table is Function dependent.

Each time an inoperable channel has not met any Required Action of Condition C or D, as applicable, and the associated Completion Time has expired, Condition E is entered for that channel and provides for transfer to the appropriate subsequent Condition.

For the majority,'of Functions in Table 3.3.3.1-1, if any Required Action fand associated Completion Time of Condition C dEiGD is not met, the plant must be placed in a MODE in which the LCO does not apply.

This is done by placing the plant in at least MODE 3 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

The allowed Completion Times are reasonable, based on optrating experience, to reach the required plant condition free full power conditions in an orderly manner and without challenging plant systems.

S -%

3 Since alternate means of monitoring V::::cr vro @h 1-#- C:::: s0 primary containment area radiation have been developed and tested, the Required Action is not to shut

,, gg down the plant but rather to t:ll w th: c: r;;ti c.-E D S- -

n t_

' E;;;j f k.t ;;r. 5.9.O.

These alternate means may be q3/g o dloo"5 g*

temporarily installed if the normal PAM channel cannot be M

restored to OPERABLE status within the allotted time.

The b

pec foM

' gQ 'y report provided to the NRC should discuss the alternate

~

M g ggh means used, describe the degree to which the alternate means are equivalent to the installed PAM channels, justify the Yo M g areas in which they are not equivalent, and provide a schedule for restoring the normal PAM channels.

SURVEILLANCE The following SRs apply to each PAM instrumentatio Function REQUIREMENTS in Table 3.3.3.1-1.

' In omebu w:% ID C.FR %.4 JAh\\n Q6P 4 <n+yc % Condl W n inued)

BWR/6 STS B 3.3-61 Rev.

O, 09/28/92

Encbure 2 to GNRO 04/00056 '

Chapter 3.4 Page 1 1

GRAND GULF CHAPTER 3.4 b

Revised PSTS Pages

EUpI.E$'S",7**

RCS P/T Licits 3.4.11 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.11 RCS Pressure and Temperature (P/T) Limits LCO 3.4.11 RCS pressure, RCS temperature, RCS heatup and cooldown rates, and the recirculatio temperature requirements shall be maintained within limits.W 52e e: " TLC.

PCOL13/ou

~

R1 APPLICABILITY:

At all times.

ACTIONS i

CONDITION REQUIRED ACTION COMPLETION TINE A.

NOTE-----.---

A.1 Restore parameter (s) 30 minutes Required Action A.2 to within limits.

shall be completed if this Condition is M

entered.

A. 2-Detemine RCS is 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months acceptable for Requirements of the continued operation.

LCO not met in MODES 1, 2, and 3.

1 l

B.

Required Action and B.1 Be in MODE 3.

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months

)

associated Completion Time of Condition A E

not met.

B.2 Be in MODE 4.

36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months (continued) i BWR/6 STS 3.4-25 Rev.

O, 09/28/92

. - ~.

l Encbsuro 2 to GNRO-94/000!4 Chaplic 3.4 Paga 3 RCS P/T Licits 3.4.11 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME i

C.

NOTE...--..-.

C.1 Initiate action to Immediately Required Action C.2 restore parameter (s) shall be completed if to within limits.

this Condition is entered.

81 @

C.2 Determine RCS is Prior to Requirements of the acceptable for entering MODE 2 LCO not met in other operation.

or 3

)

than MODES 1, 2, and 3.

]

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.11.1

.. NOTE-.-..---.---.-----.-

Only required to be performed during RCS A'_heatup and cooldown operationC-f "e9 pcQ.

sira:-7 ice - un :.t riidreiRu c - tening.

i R1)

Veri fy dCS e ra-

, "" + - -- --+ -

= = A9-30 minutes RCS heatup and cooldown_ rates are M :Mo v.,,..,, seesii'ici i., n.: - RE. f b lo6*k n any 2 6dM w

~

SR 3.4.11.2 Verify RCS pressure and RCS temperature are Once within within the criticality limits specified in 15 minutes the PTLR.

prior to control rod withdrawal for the purpose of achieving criticality (continued)

BWR/6 STS 3.4-26 Rev.

O, 09/28/92

Encbsure 2 to GNRO-94/00056 Chapter 3.4 Page 4 RCS P/T Li::its 3.4.11 SURVEILLANCE RE@ IREMENTS (continued)

SURVEILLANCE FRE@ENCY l

SR 3.4.11.3

NOTE------------.--..-.-

Only required to be met in MODES 1, 2, 3, and 4 [with reactor steam done pressure a25psig].

Verify the difference between the botton Once within head coolant temperature and the reactor 15 minutes pressure vessel (RPV) coolant temperaturey prior to each is N 'frit: 0;;;;44f;d in Q,e.~TL.,.

startup of a PC recirculation 93/)1

- jas F pump 4

o R1 SR 3.4.11.4

..... -- - -. - - -. - - - -NO T E-. --------------.

Only required to be met in MODES 1, 2, 3, and 4.

Verify the difference between the reactor Once within coolant temperature in the recirculation 15 minutes i

loop to be started and the RPV coolant _

prior to each temperature 4:

te.i., tr.e : tens a = ft:)

startup of a A6. th: "n.".

~

recirculation pm 93//s - -

_ _ 6 5o F P""P

$3 SR 3.4.11.5

-.-..---.--..---..-NOTE.--.----...-..------

Only required to be perfoneed when tensioning the reactor vessel head bolting studs.

Verify reactor vessel flange and head 30 minutes flange temperatures are 4ifthir. the 14mitD9-A6 ::i' s 4.-...i TTLE.

NC)

(continued)

C'\\ 3)/I L

\\

BWR/6 STS 3.4-27 Rev.

O, 09/28/92

s

...m Enclosurs 2 to GNRO-04/00056 chapi., s.4 e.g. s RCS P/T Licits

'3. 4.11 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.4.11.6

.--------NOTE-----.-.------.-.---

Not required to be performed until 30 minutes after RCS temperature s 80*F in MODE 4.

Verify reactor vessel flange and head 30 minutes flange temperatures arem:::

tr.

=ttJP 9 Cot A 5::;; fin 1; ::: rPuy.

. 9.lt 1y-

.h*10 '

SR 3.4.11.7

-.....-------------N0TE--..-.-..---.--..--.

Not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after RCS temperature s 100'F in MODE 4.

Verify reactor vessel flange and head 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Aflange temperatures are- ::: tM P-* *#

:: in = nut.

R 760F Tn$c.c4 297A '- PD DL

'G)ll R.2 BWR/6 STS 3.4-28 Rev.

O, 09/28/92

a En& sum 2 io GNR494Ma%

Chapter 3 4 Page 6 INSERT 28A SR 3.4.11.8

NOTE-------------

Only required to be met in single loop operation with the operating recirculation pump not on high speed and THERMAL POWER

< 36% of RTP.

Verify the difference between Once within the bottom head coolant 15 minutes temperature and t_he RPV coolant prior toLan temperatureis@100F.

increase in.

J THERMAL i

POWER or an l ECD' increase in 9 3/11 loop flow Al SR 3.4.11.9

NOTE-------------

Only required to be met in single loop operation with the operating recirculation pump not on high speed, and THERMAL' POWER

< 36% of RTP, and the idle recirculation loop not isolated from the RPV.

Verify the difference between ~

Once within

'l the reactor coolant temperature 15 minutes in the recirculation loop not in prior to an operation and the RPV coolant.

increase in temperature is @ 50*j.

THERMAL

\\

POWER or an

' ptet increase in 43/in loop flow R1 l

i 1

SR 3.4.11.10

NOTE-------------

{

Only required to be performed pq)(

during RCS inservice leak and

_j hydrostatic testing.

A S/ / l

-l R)

~

Verify the RCS pressure, RCS 30 minutes temperature, and RCS heatup and cooldown rates are within the limits specificled in the PTLR.

INSERT 3.4-28

Encbsurs 2 to GNRO-94/00050 Chapter 3.4 Pags 7 i

RCS P/T Limits B 3.4.11 B 3.4 REACTOR COOLANT SYSTEM (RCS)

B 3.4.11 RCS Pressure and Temperature (P/T) Limits BASES

I l '

I I

I-I

'I BACKGROUND All components of the RCS are designed to withstand effects of cyclic loads due to system pressure and temperature changs.

These loads are introduced by startup (heatup) and shutdown (cooldown) operations, power transients, and reactor trips.

This LCO limits the pressure and temperature changes during RCS heatup and cooldown, within the design assumptions and the stress limits for cyclic operation.

[P M The PTLR contains P/T limit curves for heatup, cooldown. and inservice leak and hydrostatic testinvo...

q3/lI '

{ Q.a rna of-ctr,p " %: ::::=t termt=[u

... m The g )I ' -

heatup curve provides limits for both heatup and criticality.

l Each P/T limit curve defines an acceptable region for normal operation.

The usual use of the curves is operational guidance during heatup or cooldown maneuvering, when pressure and temperature indications are monitored and compared to the applicable curve to determine that operation is within the allowable region.

The LCO establishes operating limits that provide a margin to brittle failure of the reactor vessel and piping of the reactor coolant pressure boundary (RCPB).

The vessel is the component most subject to brittle failure.

Therefore, the LCO limits apply mainly to the vessel.

10 CFR 50, Appendix G (Ref.1), requires the establishment.

of P/T limits for material fracture toughness requirements of the RCPB materials.

Reference 1 requires an adequate margin to brittle failure during normal operation, anticipated operational occurrences, and system hydrostatic tests.

It mandates the use of the American Society of Mechanical Engineers (ASME) Code,Section III, Appendix G (Ref.2).

The actual shift in the RTm of the vessel material will be established periodically by removing and evaluating the irradiated reactor vessel material specimens, in accordance with ASTM E 185 (Ref. 3) and 10 CFR 50, Appendix H (Ref. 4).

The operating P/T limit curves will be adjusted, (continued)

BWR/6 STS B 3.4-51 Rev.

0, 09/28/92

Encbaurs 2 to GNRO-94MXX)50 Chaptu 3 A Pagi 8 RCS P/T Limits B 3.4.11 BASES j

ACTIONS C.1 and C.2 (continued)

Operation outside the P/T limits in other than MODES I, 2, and 3 (including defueled conditions) must be corrected so that the RCPB is returned to a condition that has been verified by stress analyses.

The Required Action must be initiated without delay and continued until the limits are restored.

Besides restoring the P/T limit parameters to within limits, an evaluation is required to determine if RCS operation is allowed.

This evaluation must verify that the RCPB integrity is acceptable and must be completed before approaching criticality or heating up to > 200'F.

Several methods may be used, including comparison with pre-analyzed transients, new analyses, or inspection of the components.

ASME Section XI, Appendix E (Ref. 6)~, may be used to support the evaluation; however, its use is restricted to evaluation of the beltline.'

SURVEILLANCE SR 3.4.11.1 REQUIREMENTS

-Q Verification that operation is within@ limits is required every 30 minutes when RCS pressure and temperature conditions are undergoing planned changes.

This Frequency NN is considered reasonable in view of the control room indication available to monitor RCS status.

Also, since T3/ )

temperature rate of change limits are specified in hourly increments, 30 minutes permits assessment and correction of g3 minor deviations.

g Surveillance for he

> wanam== atu ooldown = 9 /k may discontinued when the cMteria q

given in the relevant plant procedure for ending the 1

activity are satisfied.

This SR has been modified by a Note that requires this Surveillance to be performed only during system hea Cooldown oDerationsfaM s icayh.c 164:g: Gu isydi vii.a

~

Q^ ? t 4 "" k~

y (continued)

BWR/6 STS B 3.4-56 Rev.

O, 09/28/92

$$Ell0$,"?"

RCS P/T Liaits B 3.4.11 BASES SURVEILLANCE SR 3.4.11.2 REQUIREMENTS (continued)

A separate limit is used when the reactor is approaching criticality. Consequently, the RCS pressure and temperature must be verified within the appropriate limits before withdrawing control rods that will make the reactor critical.

Perfonning the Surveillance within 15 minutes before control rod withdrawal for the purpose of achieving criticality provides adequate assurance that the limits will not be exceeded between the time of the Surveillance and the time of the control rod withdrawal.

)

SR 3.4.11.3 and SR 3.4.11.4 f

Differential temperatures within the applicab limits ensure that thennal stresses resulting from the startup of g

an idle recirculation pump will not exceed design allowances.

In addition, compliance with these limits ensures that the assumptions of the analysis for the startup of an idle recirculation loop (Ref. 8) are satisfied.

Performing the Surveillance within 15 minutes before starting the idle recirculation pump provides adequate assurance that the limits will not be exceeded between the i

time of the Surveillance and the time of the idle pump start.

An acceptable means of demonstrating compliance with the temperature differential requirement in SR 3.4.11.4 is to compare the temperatures of the operating recirculation loop and the idle loop.

SR 3.4.11.3 has been modified by a Note that requires the Surveillance to be met only in H0 DES 1, 2, 3, and 4t{with reactor steam dome pressure a 25 psig]t In MODE 5, the overall stress on limiting components is lower; therefore, AT limits are not required.

SR 3.4.11.5. SR 3.4.11.5. and SR 3.4.11.7 Limits on the reactor vessel flange and head flange temperatures are generally bounded by the other P/T limits (continued)

BWR/6 STS B 3.4-57 Rev.

O, 09/28/92

$$pf.UIp$.To" RCS P/T Limits B 3.4.11 BASES SURVEILLANCE SR 3.4.11.5. SR 3.4.11.6. and SR 3.4.11.7 (continued)

REQUIREMENTS during system heatup and cooldown.

However, operations approaching MODE 4 from MODE 5 and in MODE 4 with RCS temperature less than or equal to certain specified values require assurance that these temperatures meet the LCO limits.

The flange temperatures must be verified to be above the limits 30 minutes before and while tensioning the vessel head bolting studs to ensure that once the head is tensioned the limits are satisfied. When in MODE 4 with RCS temperature s 80*F, 30 minute checks of the flange temperatures are required because of the reduced margin to the limits. When in MODE 4 with RCS temperature s 100*F, monitoring of the flange temperature is required every Oc.o t 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months to ensure the temperatures are within the limits 43/fl H _ ND I

"^

The 30 minute Frequency reflects the urgency of maintaining the temperatures within limits, and also limits the time M

that the temperature limits could be exceeded.

The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months TN Frequency is reasonable based on the rate of temperature 65 *5 A change possibie at these temperatures.

REFERENCES 1.

10 CFR 50, Appendix G.

2.

ASME, Boiler and Pressure Vessel Code,Section III, bhbdEeA Pcadke Erb

^EE'"

ASTM E 185-82, July 1982.h dome.ilec, 71.h3 St. L6M-

~

3.

astse Gok A pue.Ar- %e 4.

10 CFR 50, Appendix H.

q% doc yesds, 5.

Regulatory Guide 1.99, Revision 2, May 1988.

ASME, Boile' anLfressure Vessel Code.Section X 6.

r T E 4feusace be A M ;

' EcWM Appendix E.

NEDO-21778-A,hw%4,ne.s Je34%(M s e r Dec r 1978 7 7.

h

8. i.AFSAR, Section {15.

5)

BWR/6 STS B 3.4-58 Rev.

O, 09/28/92

(nclosure 2 to GNRO-94/00056 1

Chapter 3.4 Page 11 t

INSERT B58A SR 3.4.11.8 and SR 3.4.11.9 FCo ed/U

-1 Differencial temperatures within the applicable limits ensure that thermal stresses resulting from increases in THERMAL POWER g

or recirculation loop flow during single. recirculation loop operation will not exceed design allowances.

Performing the Surveillance within 15 minutes before beginning such an increase in power or flow rate provides adequate assurance that the limits will not be exceeded between the time of the Surveillance and the time of the change in operation.

An acceptable means of demonstrating compliance with the temperature differential requirement in SR 3.4.11.9 is to compare the temperatures of the operating recirculation loop and the idle loop.

Plant specific test data has determined that the bottom head is not subject to temperature stratification with natural circulation at power levels as low as 36% of RTP or with any single loop flow rate when the recirculation pump is on high speed operation.

Therefore, SR 3.4.11.8 and-SR 3.4.11.9 have been modified by a Note that requires the Surveillance to be met only under these conditions.

The Note for SR 3.4.11.9 further-limits the requirement for this Surveillance to exclude comparison of the idle loop temperature if the. idle loop is isolated from the RPV since the water in the loop can not be introduced into the remainder of the reactor coolant system.

SR 3.4.11.10 Fcot Verification that operation is within PTLR limits is required 93/II every 30 minutes when RCS pressure and temperature conditions Rl are undergoing planned changes.

This Frequency is considered reasonable in view of the control room indication availsble to monitor RCS status.

Also, since temperature rate of change limits are specified in hourly increments, 30 minutes permits assessment and correction of minor deviations.

Surveillance for inservice leakage and hydrostatic testing may be discontinued when the criteria given in the relevant plant procedure for ending the activity are satisfied.

This SR has been modified by a Note-that requires this Surveillance to be performed only during inservice leakage and.

hydrostatic testing.

INSERT B 3.4-58

. -.. to GNRO-94/00056 Chapter 3.6 Page 1 GRAND GULF SECTION 3.6 l

l Revised PSTS Pages i

i

-~

Encbsure 2 io GNRO44/00056 Chapter 3 8 Paga 2 Orywell 3.6.5.1-3.6 CONTAINMENT SYSTEMS 3.6.5.1 Orywell LCO 3.6.5.1 The drywell shall be OPERABLE.

APPLICABILITY:

MODES 1, 2, and 3.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

Drywell inoperable.

A.1 Restore drywell to I hour OPERABLE status.

B.

Required Action and 8.1 8e in MODE 3.

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months a

associated Completion Time not met.

Agl 8.2 Be in MODE 4.

36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months SURVEILLANCE REQUIREMENTS FREQUENCY SURVEILLANCE h SR 3.6.5.1.1 Verify bypass leakage is 10hofthe 003 W#

bypass leakage limit.

3 L 5 u--t d

54 A SR 3.6.5.1.2 Visually inspect the exposed accessible N40}Fmonths' j

h interior and exterior surfaces of the drywell.

BWR/6 STS 3.6-54 Rev.

O, 09/28/92

Encbsurs 2 to CNRO-94/00056 Chapter 3.0. Pagt. 3 INSERT 54A i

18 months following 2 consecutive tests with bypass leakage

> 10% of the bypass leakage

'l limit until 2 consecutive tests are s 10%

of the bypass leakage limit AUR 36 months fallowing a-test with bypass leakage

> 10% of the bypass leakage limit AND 120 months to CNRO.94/00056 Chapter 3 6 Page 4 Drywell Air Lock 3.6.5.2 i

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY

.6.5. b \\-.---------------h 0TE-----------K--fter L SR 0lyrequiredtobeyerformedonce ea closing.

s Verify s I leakage rate s [200] scfh 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months when the between the d r seals is pressurized o a [11.5] psig

)

SR 3.

5.2.2 Verify drywell air lock seal air fl k

7 days ressure is a [90]

-ig.

\\

/k i

SR 3.6.5.

) -.---.-.---------. NOTE-------------------

Only required to be performed upon entry

\\

into drywell.

QCo Verify only one door in the drywell air 18 months G, /13,

lock can be opened at a time.

3

\\

s 2.

SR 3.6.5. S(

---..-----------.-NOTE----...----..-.----

An inoperable air lock door does not invalidate the previous successful performance of the overall air lock leakage test.

Verify ovgrall drywell air lock leakage 18 months overall air l@ock leakage test at rate is N 2(

scfh by performing an Og)

I i*fil.5)*psig.

1 (continued)

BWR/6 STS 3.6-58 Rev.

O, 09/28/92 i

i

Enc 6 cure 2 to CNRO-24/00056 Chapter 3.6 Page 5 Orywell Air Lock 3.6.5.2 SURVEILLANCE REQUIREMENTS - (continued)

SURVEILLANCE FREQUENCY

[1 months

[90] y, from an i 'tial pressure SR 3.6.

.5 Ver ig, the dryw 1 air lock sea pneumat'c system pres re does not de y f L. C) (-

at a rat equivalent to

[30] psig fo-eriod of 0] days.

g i

')

BWR/6 STS 3.6-59 Rev.

O, 09/28/92

.I l

1

Enebeurs 2 to GMRO-94/03050 Chapter 3.0 Poe 6 Drywell B 3.6.5.1 BASES ACTIONS A.1 (continued) commensurate with the importance of maintaining the drywell OPERABLE during MODES 1, 2, and 3.

This time period also ensures that the probability of an accident (requiring drywell OPERABILITY) occurring during periods when the drywell is inoperable is minimal.

Also, the Completion Time i

is the same as that applied to inoperability of the primary containment in LCO 3.6.1.1, " Primary Containment."

B.1 and 82 If the drywell cannot be restored to OPERABLE status within the required Completion Time, 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 at least MODE 3 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and to MODE 4 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

CI SURVEILLANCE SR 3.6,5.1.1 o,9 b

REQUIREMENTS The analyses in Reference 2 are based on a maximum drywell bypass leakage.

This urveilla e ensures that the actual c>

drywell bypass ka e ssl10 of the acceptable A/ d gar M '^"' g design value o ft assumed in the_ safety analysis.

2 b o, ' p

" The leakage test is performed'"~" '""

-~P' consistent with the difficulty of performing the test, risk of high PCoC-c Dn radiation exposure, and the remote possibility that a q 3/,3 componer,t failure that is not identified by some o her a

drywell or arimary containment SR might occur..

per ng I^f*

e la h

o t

nh rea cy i

(6IlO

ereTorn the Frequency was concluded to be acceptable from a reliability standpoint.

SR 3.6.5.1.2 The exposed accessible drywell interior and exterior surfaces are inspected to ensure there are no apparent (continued)

BWR/6 STS B 3.6-110 Rev.

O, 09/28/92

Enc 65ure 2 to CMtO 94/C0050 Chapter 3.0 Page 7 INSERT B110A This Surveillance is performed at least once every 10 years.

If during the performance of this required Surveillance the drywell bypass leakage rate is > 10% of the drywell bypass leakage limit the Surveillance Frequency is increased to every 36 months.

If during the performance of the subsequent consecutive Surveillance the drywell bypass leakage rate is s 10% of the drywell bypass leakage limit the 10 year Frequency may be resumed.

If during the performance of two consecutive Surveillances the drywell bypass leakage is > 10% of the drywell bypass leakage limit the Surveillance Frequency is increased to at least once every 18 months.

The 18 month Frequency is maintained until during the performance of two consecutive Surveillances the drywell bypass leakage rate is s 10% of-the drywell bypass leakage limit, at which time the 10 year Frequency may be resumed.

For two Surveillances to be considered consecutive the Surveillances raust be performed at least 12 months apart.

Since the Frequency is performance based,

$$I."rYPSo".T Orywell Air Lock B 3.6.5.2 BASES ACTIONS D.1 and 0.2 (continued) conditions from full power conditions in an orderly manner and without challenging plant systems.

SURVEILLANCE SR 3.6.5.2.1 REQUIREMENTS pofthSR requir s a test be hi erformed to verify seal leakage res a [11.5] \\ 'g.

drywell r lock doo at pres p

A seal I kage rat limit of s 00] scf has been esta ished Koensur the inte ity of the eals.

e Surveillanc is o ly requi ed to be erformed on after ch closing.

he Fr uency o 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months s based on seratin experience and is c nsidere adequate view of t other 'ndications favail le to p nt opera 'ons person 1 that e seal is intact.

pcot

'NSR 3.6.5.

2 W3 Eve 7 days e dr ell ir lock al air fl k pres re is veri ed to be

[9 psig o ensure that the s I sys m remain viable.

t must be hecked b ause it c id ble d down dur or foi owing acce through he air lo

, whi curs reg arly, e 7 day Fr uency ha been sho to be ac ptable, sed on erating e erience, nd is con dered tadeq ate in vi of th other indi tions to he plant opera ' ns perso el tha the seal

'r flask p essure is

' (ow.

t (SR 3.1i. 5. 2. 3 The air lock door interlock is designed to prevent simultaneous opening of both doors in the air lock.

Since both the inner and outer doors of the air lock are designed to withstand the maximum expected post accident drywell pressure, closure of either door will support drywell OPERABILITY.

Thus, the door interlock feature supports drywell OPERABILITY while the air lock is being used for personnel transit in and out of the drywell.

Periodic testing of this interlock demonstrates that the interlock will function as designed and that simultaneous inner and outer door opening will not inadvertently occur.

Due to the (continued)

BWR/6 STS B 3.6-117 Rev.

O, 09/28/92

Encbsurs 2 to GNRO 94/0005G Chaoter 3.6 Paas 9 Drywell Air Lock B 3.6.5.2

,t'C O L

~

kO BASES ce kD U[g

^

SURVEILLANCE SR 3.6.5.

(continued) cenc+.'fiq o.

REQUIREMENTS hat h - W-A mnhenic$natureofthisinterlock,andgiven f

the interlock mechanism is only challenged whenndrywell is J

OCG-'

cyd (6tT5D, P' months.The918fmonthFrequen )is based on the this test is only required to be perfomed once y everyM18 need to perform this Surveillance under the# conditions that apply during a plant outage and the potential forWccrfc c. At) h' - '

'tmpufmeG trensient-M---the _Surai U =ef wer:

~

zthe rcactor-+t-cower-f Tperating experience has shown these components usually pass the Surveillance when perfomed at h

the118}"Thonth Frequency, which is based on the refueling cycle.

Therefore, the Frequency was concluded to be acceptable from a reliability standpoint.

The Surveillance is modified by a Note requiring the Surveillance to be performed only upon entry into the

[

drywell.

SR 3.6.5..

dq *-

bo q,

This SR requires a test to be performed to verify overall air lock leakage of the drywell air lock at pressures t*{11.57psig.

The*T18}* month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage a_nd the potential for en-Eurvet=nea-=cr: -cerf;md witt)

-; ann:c :ransient i; bherating experience has shown these

_ reactor at nnwar /

components usually pass the Surveillance when performed at the [18] month Frequency, which is based on the refueling cycle. Therefore, the Frequency was concluded to be acceptable from a reliability standpoint.

This SR has been modified by a Note indicating that an inoperable air lock door does not invalidate the previous successful performance of an overall air lock leakage test.

This is considered reasonable, since either air lock door is capable of providing a fission product barrier in the event of a DBA.

KDb sa 3.s5.2.s w 93/l3 N

s alpneu\\m.ic his R ensuTgs tha he dr 1 air lock system pressuresdoes n t decay an unaccep ble rate. N e Th (continued)

BWR/6 STS B 3.6-118 Rev.

O, 09/28/92 i

S'a7.U$ Y[$7 "

Drywell Air Lock-B 3.6.5.2 BASES l

SURVEILLANCE SR 3.6.5:2.5 (contin d)

)

REQUIREMENTS or drywell PERABILI down to a will supp[90]

air ock sea i

re of ig.

Si e the ai lock seal pneu tic pres air fl k pressu is verifi in SR 6.5.2.2-be l

gg n [90] p *, a dec rate s [3 psig o r [10] d s is ceptable.

The [10 day intery is bas on engi ering 93[l3

,' ju nt, co idering at there i no post ated DB where J

the rywell is till pr surized[10 days af r the e nt.

The [

month F quency based on t need t erform this Su veillance der the onditions t t apply uring a lant ou e and the otentia for an unp nned tra ient i e Survei ance were rfonne ith the re tor at p er.

Op ating ex rience has hown th these co nents us 11y pass the Surve'llance when erforme at the [1 month Frequ cy.

The fore, the.

quency as conclude to be

_accepta le from a eliability andpoi REFERENCES 1.

10 CFR 50, Appendix J.

LIFSAR, Chaptersk6 and 15[

2.

O BWR/6 STS B 3.6-119 Rev.

O, 09/28/92

._.._.m....

Encbsurs 2 to G4RO-94/00056 Chaplet 5 Page 1 i

.l 4

)

GRAND GULF CHAPTER 5

101.7;#."i"

ATTACHMENT 1 CTS - PSTS COMPARISION DOCUMENT 1A: MARKUP OF CTS 1B: DISCUSSION OF CHANGES 1C: NO SIGNIFICANT HAZARDS CONSIDERATIONS

a,.

w..

a,.

.m4 ea-4<

=A A

a'.9 as n.

4 s

x x

s

_,_4 a-0 ATTACHMENT 1A CTS - PSTS COMPARISION DOCUMENT MARKUP OF CTS to GNRO-94/0005(

Chapter 5 Page 4 NL-90/06 6.0 ANINISTRATIVE CONTROLS

6. / 4-5.Q 5, /

6.1 RESPONSIBILITY h.,

The@el Manager. Plant Operation)shall be responsible for overall 5.1' 1 6.1.1 unit operation and shall delegate in writing the succession to this responsi-bility during his absence.

g, f, q 6.1.2 The Shift Superintendent or, during his absence from the Control Room, a designated individual shall be responsihte_ for the Control Roos command

_ unction.fA nagement91rective this effe K signed by t Vice President, OS2 Opeqatihos GGNS hallbeKissuedto 11 station 'nersonnel on _

g, a 6.2 ORGANIZATION annual bhis.

6.2.1 0FFSITE AND ONSITE ORGANIZATIONS y 7,;

Onsite and offsite organizations shall be established for unit organization and corporate management, respectively.

The onsita and offsite organizations shall include the positions for activities affecting the safety of the nuclear power plant.

5. 2. J. a a.

Lines of authority, responsibility, and consunication shall be established and defined for the highest management levels through intermediate levels to and including all operating organization positions.

These relationships sha 1 be documented and updated, as appropriate, in the fore of organization charts, functional descriptions of departmental responsibilities and relationships, and job descriptions for key personnel positions, or in equivalent forms of documentation.

These requiremnts shall be doctmented in the

, UFSAR gnd_ updated at least annusy P y.2.J,b b.

Thegioneral Manaaer. Plant Doerations]shall be responsible for over-all unit safe operation and shal' have control over those onsite pH activities necessary for safe operation and maintenance of the plant.

5. Q. l. c The Vice President onerations GGNOshall have corporate responsibility c.

for overall plant nuclear safety and shall take any esasures needed to ensure acceptable performance of the staff in operating, maintaining, and providing technical support to the plant to ensure nuclear safety.

d.

The individuals who train the operating staff and those who carry 9, p,, 4 out health physics and quality assurance functions may report to the 4

appropriate onsite manager; however, they shall have sufficient organizational freedom to ensure their independence from op ating' pressures.

Q.,pwg;,.S Q % k g & h g g UNIT STAFF h'Q, Awe y &.&.n %. LT5 A A

$. a. 2 6.2.2 The unit organization shall be subject to the following; f

~ ~b::.r er, itj ;t.'ft :P.11 t,e ;g:::d ;f e le=*t tha -sintam-ehh.

a p

_ ~.

: n::tti;; -:tra " Tcle ".. *=1.

E b.

At least one licensed Reactor Operator shall be in the control roos

6. R. A b when fuel is in the reactor.

In addition, while the. reactor is in 1

OPERATIONAL CONDITION 1, 2 or 3, at least one licensed Senior Reactor Operator shall be in the Control Room.

GRAND GULF-UNIT 1 6-1 Amendment No. 76

m.

Encbsure 2 to GNRO-94/00056' Chaptzt 6 Page 6 INSERT 1A l

A non-licensed operator. shall be on site when fuel is in the reactor and an additional'non-licensed operator shall be on site while the unit is in MODE 1, 2,

or 3.

i l

l I

l l

i I

INSERT 6-1

%; %."?"

ADMINISTRATIVE CONTROLS UNIT STAFF (Continued) 5 23,f c.

A health physics technician

shall be onsite when fuel is in the reactor.

d.

11 CORE ALTE NS shall e observed'a d directly s rvised by D

,e her a icensed for Rea r Operator Senior Reac Operator Lim d to 1 Hand who ha o other con rent responsibili ie durin this ope etion. f (DELETED) L g

f.

' ' 7' 7, q Administrative procedures shall be developed and implemented to limit the working hours of unit staff who perform safety-related functions; e.g., senior reactor operators, reactor operators, health physicists, auxiliary operators, and key maintenance personnel.

Adequate shift coverage shall be maintained without routine heavy use of overtime.

However, in the event that unforeseen problems 1

require substantial amounts of overtime to be used, h g the following guidelines shall be followed:

g 1.

An individual should not be permitted to work more than 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months straight, excluding shift turnover time.

2.

An individual should not be permitted to work more than 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months in any 24-hour period, nor more than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months in any 48-hour period, nor more than 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months in any seven-day period, all excluding shift turnover time.

3.

A break of at least eight hours should be allowed between work periods, including shift turnover time.

4.

Except during extended shutdown periods, the use of overtime should be considered on an individual basis and not for the entire staff on a shift.

Any deviation from the above guidelines shall be authorized by the A

Veneral Manaaer, Plant operations)or his designee, or higher levels of management, in accordance with established procedures and with documentation of the basis for granting the deviation.

Controls shall be included in the procedures such that individual overtime shall be 5,2,2.)fThe number of health physics technicians may be less than the minimum require-ments for a period of time not to exceed 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months in order to accommodate unexpected absence provided immediate action is taken to fill the required positions.

GRAND GULF-UNIT 1 6-2 Amendment No. 82

.~.

-.. ~.. ~..

.. ~ ~

-.. - - =,.

' Endosure 2 to GNRO-94/00050 -

Chapter 5 Page 7 i

L i

INSERT 2A or during extended periods of shutdown for refueling, major maintenance, or major plant modification,.on a temporary basis b

h n

k l

i INSERT 6-2 i

l

~

=.~.-

Enclosurs 2 to CNRO-94/0005C Chapter 5 Page 8 ADMINISTRATIVE CONTROLS UNIT STAFF (Continued) reviewed month 1y'by thekneral Manager, Plant onoratM or his designee to assure that excessive hours Nive not been assigned.

Routine deviation from the above guidelines is not authorized.

.o,

5. 2. '2.T /g.

The Operations Superintendent, Shift Superintendents, Operations p1 Assistants,-and Shift Supervisors shall each hold a Senior Reactor Operator License.

h.

The Manager Plant Operations must have been a Senior Reactor Operator

.or have been certified on a plant of this ty

- [ 6. 2. 3 INDEPENDENTSAFETYENGINEERINGGROUP(ISEGN N

N FUNCTION s

l N

N 6,2.3.1 The ISEG shall

}NRCissuances,industrya nction to examine unit operating characteristics,.

series, Licensee Event Reports, and other sources of p'lant desi and operatin experience information, including plants of similar fdesigrt which indicate are for improving planbgafety.

COMPOSI N

6.2.3.2 Th ISEG shal be composed of ulti-disciplined cated onsite, group with a inimum assigned complement o five engineers or a)propr,iate

! specialists.

GRAND GULF-UNIT 1 6-2a Amenhnt No. ' 75

Encbsurs 2 to G!!tO-94/00056 Chapt $r 5 Pr08 9 I

i DELETED l

l i

i i

i l

GRAND GULF-UNIT 1 6-3 Amendment No. 45

. to GNRO 04/00056 Chapter 5 Page 10 P

DELETED i

1 GRAND GULF-UNIT 1 6-4 Amendment No. 45

~ to GNRO-04!00056 Chaptor 5 Page 11 7

MINIMUM SHIFT CREW COMPOSITT0tf

,NUMBEROFih0!VIDUALSREQUIREDTOFIkPOSITION POSITION C'ONDITIONS 1,'2 CONDITIO 5,4 & 5 s

e (ao?'

1 "1"W l

g.,. 7, g 2___

1)

,mg

- _. m s

ABLE NOTATION LA3 SS -ShiftSuperintendentwMi a Senior Reactor 0 ators License on Unit 1.

SRO 'Iodividul with a Senior Rhetor Operators Lice e on Unit 1.

R

- Init idual h a Reactor Opikators License on Un t 1.

A0 - Auxil ry Oper or.

STA - hift echnical visor._

Except for the Shift Superintendent, the Shift Crew Composition may be one y; 2.2. c less than the minimum requirements of Table 6.2.2-1 for a period of time not to exceed 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months in order to accommodate unexpected absence of onduty shift crew members provided immediate action is taken to restore the Shift Crew Composition to within the minimum requirements of Table 6.2.2-1.

This provision does not permit any shift crew position to be unmanned upon shift change due to an oncoming shif t crewman being late or absent.

During any absence of the Shift Superintendent from the Control Room while the Cl 1 unit is in OPERATIONAL CONDITION 1, 2 or 3 an individual with a valid SRO license l shall be designated to assume the Control Room command function.

During any absence of the Shift Superintendent from the Control Room while the unit is in OPERATIONAL CONDITION 4 or 5, an individual with a valid SRO or R0 license shall be designated to assume the Control Room command function.

Tr crici.her 20 c. shifUhc.aeEthedilifintions ivr i.it ceeined)

O'D hbcsition._gecified-for__00trion4MRommissiWs Polic 7 ement

,- on_ Engineering Expertise on Shift, may also serve as th_e STA.

this a shirw. tnen tnt-Sep,arate STA position,,may be elimina d.

u GRAND GULF-UNIT 1 6-5 Amendment No. 64

m.

Encbsure 2 to CNRO.94/00056 Chapter 5 Page 12 NL-90/06 ADMINISTRATIVE CONTROLS OLA5JsCINDEPENDENT SAFETY ENGINEERING GROU x

s.

RE5PONSIBILMIES 6.2.3.

The IS shall be resp sible for maint ning surveil nce of unit activit s to pro de independent verification" th these activ ties are perYorme correctly nd that human rrors are reduce as much as actical.

AUTH0 TY 6.2.3.4 The IS shall na detailed rec ndations for evised proce res, equipment diff tions, ma tenance activi ies operations tivities or ther (means of improving unit safet to the Vice President, Operations GGNS.

6.2.4 SHIFT TECHNICAL ADVISOR 6.2.4.1 The Shift Technical Advisor shall provide technical support to the 5* 1 D Shift Superintendent in the areas of thereal hydraulics, reactor engineering-l and plant analysis with regard to safe operation of the unit.

6.3 UNIT STAFF QUALIFICATIONS 5,2 7,$

6. 3.

6.3.1 Each member of the unit staff shall meet or exceed the minimum qualifica-tions of ANSI N18.1-1971 Q

Radiation Control and thift Technical Adviso'r7 do shall meet or exceed th 3

education and experience requirements of ANSI /ANS 3.1-1981 as endorsed by Regul-s M

utory Guide 1.8, Rev_ision 2.JS8f7a licensed sonnel ~

snaTNeeet or Cexcese,thecriter,iaoftheaccredW lic train mr

~and thos members, ref tne inaspenc safety Engi r'ng Group u for asett ein comple '

ified in tion 6.2.3., each of hall have a chelor f Scfince Oc.AL'mentsp de' e or registere as a Profes nel Enginee and shall h e at'le t two ears xperie e in their field, at le t one year f which expo ionce sh 11 br /

(in he clear isld.

~8 4 =551

'N 6.4)1,Aretr i

replacement 4 raining p' rae for unit staf hall.be'

- ma nta ed unde he d ion of the ager, N

.r Traini shall mee r-exc dI acc tion ria and 1 R Part 5, nd shall.

lude-famil riz ion with evant dustry opera onal exper ence.

6.5 REVIEW A E AUO5T \\

I

/

~ - - _

3 s'

i 6. 5.1 PLANTSAFETKREVIEWCDWITTEE(PSR

~

4L __.

N FUNCTION

\\\\

N 6>s5.1.1MheP shall f ction to a ise the Gener Manager, Plant erations ) l'

/ on\\1l matters r tad to n lear safet j

~

f

~

fAfeT,ct GRANO GULF-UNIT 1 6-6 Amendment No. 76

=

Encbsure 2 to GNRO 94/CD56 Chapter 5 Page 13 ADMINISTRATIVE coltTROLS Pt,)yITSAFETYREVIEWCom!TTEE(PSRC)(Continued)

C0 ITION

6. 5.1. 2 The PSRC shall be composed of nine seabers of GGN5 organizat n management at the Superintendent level or above.r g

ALTERNATES

6. 5.1. 3 All al rnate seabers shall be appointed in Manager, Plant op two alternates shalattons to serve on a temporary b iting by the General one time.

participate as voting members is; however, no more than:

L.S n PSRC activities at any NEETING FREQUENCY 6.5.1.4 The PSRC shall ese at least once by the PSRC Chairman or Vice hairman.

r calendar month and as convened QUCR\\nt 6.5.1.5 The quorum of the PSRC n saary for the performance of the PSRC.

responsibility and authority prov i s of these Technical Specifications shall-consist of the Chairman or Vice 41 and four members' including alternatas, i

RESPONSIBILITIES

6. 5.1. 6 The PSRC shall be esponsible for view of:

a.

Station admini rative procedures a hanges thereto, b.

The safety a luations for (1) procedu

, (2) changes to' procedures,-

equipment o systems, and (3) tests or risents completed under the previs on of Section 50.59, 10 CFR, to erify that such actions.

did not stitute an unreviewed safety ques ten and all programs-requi by Specification 4.8 and changes the te, c.

Propos procedures and changes to procedures, ipment or systaes.

whic may involve an unreviewed safety question a defined in 5

on 50.59, 10 CFA.

d.

P sed' tests er experiments which may involve an viewed safety _

stion as defined in Section 50.59, 10 CFR.

Proposed changes to Technical Specifications or the Ope ting License.-

e.

GRAND GULF-UNIT 1 67 Amendeont No. 76

l Encbsure 2 to CNRO-94/000$6 Chapter S Pese 14 ADMINISTRATIVE CONTROLS RESPONSIBILITIES (Continued) f.\\

Reports of violations of codes, regulations, orders, Tec cal Specifications, or Operating License requirements havi nuclear safety ignificance or reports of abnormal degradation of s t contain radioactive material.

taas designed g.

Rep ts of significant operating abnormalities deviations from norma and expected performance of plant. equi nt that affect nuclear safety.

h.

Review o all REPORTABLE EVENTS.

.l 1.

All recogni ed indications of an unan icipated deficiency in some aspect of de gn or ;,peration of saf ty related structures, systems, or components, j.

The plant Secur y Plan and chan s thereto.

k.

The Emergency Plan nd changes thereto.

1.

Items which may cons tute potential nuclear safety hazard as i

identified during revi of facility operations.

m.

Investigations or analys of special subjects as requested by the-Chairman of the Safety R iew Committee.

i n.

Changes to the PROCESS ONT L PROGRAM, OFFSITE DOSE CALCULATION MANUAL, and radwaste systems.

u 1

o.

Fire Protection Pro ram and chan es thereto.

l.

AUTHORITY 6.5.1.7 The PSRC shall-

a. Recommend 1 writing to the General Mana er, Plant Operations approval or disappr al of items considered under 5.1.6.a. c, d, e, j, and k, above.

b.

Render terminations in writing to the Gener 1 Manager, Plant Opera-tions th regard to whether or not each itse nsidered under-6.5.1

.a. c and d, above, constitutes an unrev d safety question.

c.

Pro de written notification within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to th SRC of disagreement be ween the PSRC and the. General Manager, Plant Ope tions; however,.

e General Manager, Plant Operations shall have res nsibility for esolution of such disagreements pursuant to 6.1.1 ab e.

RECORDS 6.5.1.

The PSRC shall maintain written minutes of each PSRC meetin ' that, at a min aus, document the results of all PSRC activities performed unde the resp nsibility and authority provisions of these Technical Specificati s.

Cop es shall be provided to_the SRC.

GRAND GULF-UNIT 1 6-8 Amendment No. 82

l Endosure 2 to GNRO44/00050 Chapter 5 Page 15 ADMINISTRATIVE CONTR01.5 6.5>2_ SAFETY REVIEW COMMITTEE (SRC)

FUNCTib 6.5.2.1 Th SRC shall function to provide independent review an& udit of designated ac ivities in the areas of:

a.

nuclea power plant operations b.

nuclear e ineering c.

chemistry an radiochemistry d.

metallurgy A

e.

instrumentation an control f.

radiological safety g.

mechanical and electrica engir erirg h.

quality assurance practices COMPOSITION

6. 5. 2. 2 The SRC shall be compose 'of a C irman and nine voting members.

Two or more additional voting member shall be onsultants to Entergy Operations, Inc., consistent with the reco endations o the Advisory Committee on Reactor Safeguards letter, Mark to Pa adino dated Oc ober 20, 1981.

The Chairman and memb s shall meet the qu ifications below.

The SRC Chairman sh I hold a Bachelor's deg e in an engineering or physical science field an equivalent and a minim m of six years of professional level man erial experience in the powe field.

In addition, the Chairman shall have t necessary overall nuclear bac ground to determine when to call consultants d contractors for dealing with c mplex problems beyond the scope of Enter Operations, Inc., expertise.

l1 The SRC memb rs shall hold a Bachelor's degree in an engineering or physical science field or equivalent experience and a minimum of fi e years of technical experience of w ich a minimum of three years shall be in on or more of the disciplines of 6.5.2.la through h.

In the aggregate, the me bership of the committee sha i provide specific practical experience in the ajority of tne disciplines f 6.5.2.la through h.

ALTERNATES 6.5.2.3 11 alternate members shall be appointed in writing by t e SRC Chairma to serve on a temporary basis; however, no more than two Iternates shall articipate as voting members in SRC activities at any one ti e.

GRAND GULF-UNIT 1 6-9 Amendment No. 65

,- to CtmOM00056 I

Chapter 5 Page 16 ADMINISTRATIVE CONTROLS COSSULTANTS 6.5.2.

Consultants, in addition to those required in Specification 6.5.'.2, shall be ilized as determined by the SRC Chairman to provide expert a ice to the SRC.

N more than three consultants 3 hall participate as voting m bers in SRC activi es at any one time.

MEETING FREQUENC 6.5.2.5 The SRC sha 1 meet at least once per calendar quarter ring the initial year of unit operation following fuel loading and at least onc per six months thereafter.

QUORUM 6.5.2.6 The quorum of the SR necessary for the perform ce of the SRC review and audit functions of these Te hnical Specifications s all consist of the Chairman or his designated alter te and at least 7 S voting members including l j alternates.

No more than a m16ori of the quorum s 11 have line responsibility for operation of the unit.

REVIEW 6.5.2.7 The SRC shall review:

a.

The safety evaluations for (1) c ges to procedures, equipment or systems and (2) tests or experim ts completed under the provision of Section 50.59, 10 CFR, to verif t t such actions did not constitute an unreviewed safety question b.

Proposed changes to procedu s, equip nt or systems which involve an unreviewed safety ques on as defin-in Section 50.59, 10 CFR.

c.

Proposed tests or exper.ents which invo ve an unreviewed safety question as defined in ection 50.59, 10 FR.

d.

Proposed changes to ppendix A Technical Sp cifications or this Operating License.

e.

Violations of des, regulations, orders, Tech ical Specifications, license requi ments, or of internal procedures or instructions having nuclear saf y significance, f.

Significa operating abnormalities or deviations rom normal and expected performance of unit equipment that affect uclear safety.

g.

All Rr ORTABLE EVENTS.

h.

All.ecognized indications of an unanticipated deficie cy in some as ect of design or operation of structures, systems, o components t at could affect nuclear safety.

i.

Reports and meetings minutes of the PSRC.

Written reports from audits of the ALARA program.

GR 0 GULF-UNIT 1 6-10 Amendt nt No,10 l

..__.m.

2.

..,_.m

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

l Encbsuto 2 to GNRO-44/CCD56 -

Chapter 6 Page 17 ADMINISTRATIVE CONTROLS

,/

AUDITS

/

6.

2.8 Audits of unit activities shall be performed under the cogniz ce of the RC.

These audits shall encompass:

a.

The conformance of unit operation to provisions contain within the Appendix A Technical Specifications and applicable 11 nse conditions t least once per 12 months.

q b.

Th erformance, training and qualifications of e entire unit staff at 1 st once per 12 months.

6 c.

The res ts of actions taken to correct defi encies occurring in unit equi ent, structures, systems or met d of operation that affect nuclear sa ty at least once per 6 months d.

The performanc of activities required y the Operationai quality Assurance Progra to meet the criter of Appendix "B", 10 CFR 50, at least once per 4 months.

A e.

The Emergency Plan an implementi g procedures at least once per 12 months, f.

The Security Plan and imp me ting procedures at least once per 12 months.

g.

Any other area of unit op ati coraidered appropriate by the SRC or the Vice President, O ratiom, i GGNS.

l h.

The Fire Protection Pr ram and i

'esenting procedures at least once per 24 months, i.

An independent fir protection and loss revention. inspection and audit shall be pe ormed at least once p 12 months utilizing either qualified offsit licensee personnel or an utside fire protection firm.

j.

An inspection d audit of the fire protectio and' loss prevention program shall e performed by an outside quali ed fire consultant 4

at intervals no greater than 36 months.

k'.

The' radio gical environmental monitoring program d the results thereof-t least once per 12 months.-

~

1.

The O SITE DOSE CALCULATION MANUAL and implementing p cedures at leas once per 24 months, m.

.T PROCESS CONTROL PROGRAM'and implementing procedures fo olidification of. radioactive wastes at least once per 24 ths.

n.

The performance of activities required by the Quality.Assuranc Program to meet the criteria of Regulatory Guide 4.15, February 979, at least once per 12 months.

G D GULF-UNIT 1

.6-11 Amendment No. 74

~.

l

- Encbsure 2 to GNRO 94/00056 Chapter 5 Page 18 ADMINISTRATIVE CONTROLS

/

A.UTHORITY T

6.5y.9 The SRC shall report to.and advise the Vice President, Operat ns GGNS on t se areas of responsibility specified in Sections 6.5.2.7 and 6

. 2. 8.

-l RECORDS 6.5.2.10 ecords of SRC activities shall be prepared, approv d and distributed as indicate below:

Min tes of each SRC meeting-shall be prepared, proved and. forwarded a.

to t e Vice Preide % Operations GGNS within 4 days following each meeti b.

Reports of reviews encompassed by Section

.5.2.7 above, shall be prepared, approved and. forwarded to the ice President, Operations GGNS within 14 ays following completion of he review.

c.

Audit repor encompassed by Sectio

.5.2.8 above, shall be forwarded to he Vice President, O rations GGNS and to the management positions resp sible for the are s audited within 30 days after completion of th audit by the diting organization.

6. 5. 3 TECHNICAL REVIEW AND NTROL ACTIVITIES 6.5.3.1 Activities which affect clear safety shall be conducted as follows:

a.

Procedures required by e nical Specification 6.8 and other proce-dures which affect p1 tn ear safety. and changes thereto, shall be prepared, reviewe and app ved.

Each such procedure or procedure-change shall be re owed by an dividual/ group other than the individual / group ich prepared t procedure or procedure change, but who say be f on the same organ ation as the individual / group which prepared he procedure or proc re change.

Procedures other than Administ tive Procedures shall b approved as delineated in writing by t General Manager, Plant ations.

The General Manager, P1 t Operations shall approve a inistrative procedures, i

security i lamenting procedures and emerge plant implementing procedure.

Temporary approval to procedures ich clearly do not.

change t intent of the approved procedures' be made by two members of the ant annagement staff, at least one of holds a Senior Reactor Operator's License.

Temporary changes sha be reviewed by the re Jewing authority within 14 days of being iss For changes to p edures which may involve a change in intent of he approved proc ures, the person authorized above to approve the rocedure shall app ve the change.

b.

Pro osed changes or modifications to plant nuclear safety-lated st uctures, systems and components shall be reviewed as des nated b the Vice President, Operations GGNS.

Each such modificat n shall reviewed by an individual / group other than.the individual /

up hich designed the modification, but who say be from the same organization as'the individual / group which designed the modifica ions.

ITmplementationofproposedmodificationstoplantnuclearsafety-5')*l related structures, systems and components shall be approved by the neral Manager, Plant Operations.

GRAND GULF-UNIT 1 6-12 Amendment No. 75

._~

.-.u to GNRO-94/00056 -

Chapter 5 Page 19 NL-90/06 ADMINISTRATIVE CONTROLS ACTIVITIES (Continued)

/

j Proposed tests and experiments which affect plant nuclear are not addressed in the Final Safety Analysis Report s ety and i be reviewedt an individual / group other than the individual /gr roposed. test or experiment.

which prepared th d.

Events oportable pursuant to Section 50.73 10 CFR Part 50 shall be vestigated and a report prepa which evaluates the event.

and which.p ides recommendations to event recurrence.

Such report shall be appro by the General Ma ger, Plant Operations.

e.

Individuals respons le for revi s performed in accordance with 6.5.3.1.a. 6.5.3.1.b, 5.3.1 and 6.5.3.1.d shall meet or exceed the qualification requir s of Section 4.4 of ANSI 18.1, 1971, as previously designated y Vice President, Operations GGNS or General Manap r, Plan peratio as applicable.

Each such review i

shall include a det ination'of er or not additional cross-disciplinary revi is necessary.

If a

shall be perfo d by the review personne' d necessary, suc,h review the appropriate discipline, f.

Each re ow shall include a determination of what or.not an 1

unre owed-safety question is involved.

g.

ecords of the above activities shall be provided to t neral Manager, Plant Operations, PSRC and/or as necessary for fred reviews.

~6.6 RE' PORTA 8LE EVENT ACTION N-6.6.1 The fo s shall be takeDJoMttPOR EVENTS:

The f c ie. ion snall be notified pu to the requirements of a.

- d tion 50.72 to 10 CFR Part 50, and a repor tted pursuant a the requirements of Section En 71 to 10 CFR Part kh b----EeMeTaaLE EVENT-sh=11~ " r ieewed by the fSRC and submitted

~

to the SRC and thg._Eic: 77.e;uens, upsi.M r F 5 ht i

SAFETY LIMIT VIOLATION 6.7.1 The following actions shall be taken in the event a Safety Limit _is violated:

n1nd b

a.

The NRC Operations Center shall be notified by telephone as soon as

' g g, a possible and in all cases within one hour.

The Vice. President, Operations GGNS and the SRC shall be notified within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

b.

A Safety Limit Violation Report shall be prepa_ red Q.he report shal1%

% reviewed by the PSRD." This' report'sha11 describe (1) app 11 cable

~

' circumstances preceding the violation, (2) effects of the violation upon unit components, systems or structures, and (3) corrective

' actions taken to prevent recurrence.

g LA1 GRAND GULF-UNIT 1 6-13 Amendment No. 76

Enc 6sure 2 to GNRO-94/0C05C Chapter 5 Page 20 ADMINISTRATIVE CONTROLS

$AFETY LIMIT VIOLATION (Continued) rwl c.

The Safety Limit Violation Report shall be submitted to the Nuclear Regulatory Commission pursuant to Section 50.4 of 10 CFR Part 50,

+,

h Lthe SRCjand the Vice President, Operations GGNS within 14 days of 9,-",-

the violation.

d.

Critical operation of the unit shall not be resumed until authorized by the Commission.

G 'I

6. 8 PROCEDURES AND PROGRAMS g,9, ;

6.8.1 Written procedures shall be established, implemented and maintained covering the activities referenced below:

The applicable procedures recommended in Appendix "A" of Regulatory g.ylq a.

Guide 1.33, Revision 2, February 1978.

I.

Refueling operations k b

.g, y, /, s

@O.

Surveillanceandtestactivitiesofsafetyrelatedequipmenk

'~

c._

d Security Plan implementation.

L o sctr h 14 h,. - M l OAa

~

n te.

Emergency Plan implementation.

5.4. ) d f.

Fire Protection Program implementation.

@- 4 PRBM5~5ENT20L PRMDAM imnle?9ntat40&

5, y, /, e ( h.

OFFSITEDOSECALCULATIONMANUALimplementation) g,9 % ~ i.

Quality Assurance Program for effluent and environmental; onitorings.

^Q. 4using the guidance in Regulatorv Guide 4.15, February _1979h

. e. z Each girocedure ut 6.8.1 abov nd changes thereto, shaIT Te reviewed';

s p4 - as equire4 by'6Q, aboveNrior to i sentatfo(and shall beNgvieyed

/

er icallhas set forth iriNdministrat procediha s.-

5,5 6.8.3 The following programs shall be established, implemented, and maintained:

a.

Primary Coolant Sources Outside Containment h 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:

1.

RCIC system outside containment containing steam or water, except the drain line to the main condenser.

2.

RHR system outside containment containing steam or water, except the line to the LRW system and headers that are isolated by manual valves.

3.

HPCS system.

4.

LPCS system.

5.

Hydrogen analyzers of the combustible gas control system.

GRAND GULF-UNIT 1 6-14 Amendment No.80

.r

..~.....m~.

Enebaure 2 to GNRO-94/00056 Chapter 5 Page 21 INSERT 14A b.

The emergency. operating procedures required to implement the requirements of NUREG-0737 and to NUREG-0737, Supplement 1, as stated in Generic Letter 82-33; All programs specified in Specification 5.5.

e.

4 i

l INSERT 1

6-14 Y

9 w

w

$$lPpSITSf" ADMINISTRATIVE CONTROLS PROCEDURES AND PROGRAMS (Continued)

5. 5. a 6.

Feedwater leakage control system.

7.

Post-accident sampling system.

8.

Su?pression pool level detection portion of the suppression pool maeup system.

The program shall include the following:

1.

Preventive maintenance and periodic visual inspection requirements and 2.

Integrated leak test requirements for each system at refueling cycle intervals or less,

b. NIn-Plant \\ Radiation Monitoring \\

N N

A regram ich will ensdv the capab' ity to accu telydetermine) th airborne 'odine concen tion in v1 1 areas und accident con 'tions.

T 's program sh include t following:

1.

Tra' ing of pe onnel,ing, and 2.

Proc ures for m itor (3.

Provis' ns for mal tenance of sam ing and an sis equi ent.

Post-acclentSampling w

5, 6. 3 c.

A program which will ensure the capability to obtain and analyze reactor coolant, radioactive iodines and particulates in plant gaseous effluents, and containment atmosphere samples under accident conditions.

The program shall include the following.

1.

Training of personnel, 2.

Procedures for sampling and analysis 3.

Provisions for maintenance of samplin,g and analysis equipment.

5. 5. 4 d.

Radioactive Effluent Controls Prooram A program shall be provided conforming with 10 CFR 50.36a for the control of radioactive effluents and for maintaining the doses to MEMBERS OF THE PUBLIC from radioactive effluents as low as reasonably achievable Theprogram(1)shallbecontainedinthe (2)shallbeimhlementedbyoperating)rocedures,and ODCM shallincluderemedialactionstobetakenwienevertheprogr(am) 3 limits are exceeded.

The program shall include the following elements:

1.

Limitations on the operability of radioactive liquid and caseous monitoring instrumentation including surveillance tests ind setpoint determination in accordance with the methodology in the 00CM, GRAND GULF-UNIT 1 6-15 Amendment No. 80, 87 to GNRO 04/00056 Chapter 5 Page 23 ADMINISTRATIVE CONTROLS PROCEDURES AND PROGRAMS (Continued)

I. y 2.

Limitations on the concentrations of radioactive material released in liquid effluents for UNRESTRICTED AREAS conforming to ten times the concentration values in 10CFR Part 20.1001 -

20.2401, Appendix B, Table 2, Column 2.

3.

Monitoring, sampling, and analysis of radioactive liquid and gaseous effluents in accordance with 10 CFR 20.1302 and with I

the methodology and parameters in the 00CM, 4.

Limitations on the annual and quarterly doses or dose commitment to a MEMBER of the PUBLIC from radioactive materials in liquid effluents released from the unit to UNRESTRICTED AREAS conforming to Appendix I to 10 CFR Part 50, 5.

Determination of cumulative and projected dose contributions from radioactive effluents for the current calendar quarter and current calendar year in accordance with the methodology and parameters in the ODCM at least every 31 days, 6.

Limitations on the operability and use of the liquid and gaseous effluent treatment systems to ensure that the appropriate portions of these systems are used to reduce releases of radioactivity when the projected doses in a 31-day period would exceed 2 percent of the guidelines for the annual dose or dose commitment conforming to Appendix I to 10 CFR Part 50, 7.

Limitations on the dose rate resulting from radioactive material released in gaseous effluents from the site to areas at or beyond the SITE BOUNDARY shall be limited to the following:

a.

For noble gases:

Less than or equal to 500 mrem / year to the total body and less than or equal to 3000 mrem / year to the skin, and, b.

For lodine-131, for Iodine-133, for tritium, and for all radior.uclides in particulate form with half-lives greater than 8 days:

Less than or equal to 1500 mrem / year to any organ, 8.

Limitations on the annual and quarterly air doses reculting from noble gases released in gaseous effluents from the unit to areas beyond the SITE BOUNDARY conforming ta Appendix I to 10 CFR Part 50, 9.

Limitations on the annual and quarterly doses to a MEMBER OF THE PUBLIC from Iodine-131, lodine-133, tritium, and all radionuclides in particulate form with half-lives greater than 8 days in gaseous effluents released from the unit to areas b

beyond the SITE BOUNDARY conforming to Appendix I to 10 CFR Part 50, and GRAND GULF-UNIT 1 6-15a Amendment No. 817 111

[

$@.",U N."$"

a ADMINISTRATIVE CONTROLS I

PROCEDURES AND PROGRAMS (Continued) 10.

Limitations on the annual dose or dose commitment to any MEMBER OF THE PUBLIC due to releases of radioactivity and to radiation from uranium fuel cycle sources conforming to 40 CFR Part 190.

dioloaical Environmental Monitorina Proaral

~

A program shall be pro ded to monitor the radiation and t

\\

radibnuclides in the en rons of the plant.

The program shall provide (1) representativ measurements of r ioactivity in the highest'p athways, and (2) rification of the kaccuracy 'otential exposure LM f the effluent mon toring program an modeling of environmen 1 exposure pathwa The program sh 11 (1) be contained

's ih,the ODCM,- 2) conform to the uidance of Appen ix I to 10 CFR Part 50, and include the fol wing:

\\

1.

Nonitoring, ampling, analysi and reporting radiation and dionuclide in the environme in accordance th the me odology an parameters in th ODCM, 2.,

A Lan Use Census o ensure that c 'nges in the use f areas s

at and eyond the TE BOUNDARY are dentified and t t

\\ modifica ions to the onitoring progr are made if r uired by the re ults of this census, and 3.

P ticipatio in an Interlaboratory Compa ison Program to ens re that iMependent chhks on the prec ion and accura y of thq measurerhents of radio'hGtive materials in. environment 1 sample matrices'are performed as part of the ality assurance program for environmental monitoring.

REPORTING RE0VIkMENTS N{-ggg fg{" S 5'. (o 6.9 ROUTINE REPORTS

~- - ]m 6.9.1 In addition to the applicable reporting requirements of Title '10, Code of Federal Regulations, the following reports shall be submitted to the Nuclear Regulatory Commission pursuant to Section 50.4 of 10 CFR Part 50.

O/N[STARTVPREPORTS( --.

l 6.9 1 1 A su ary re rt of plan startup and ower escalatio testing shall be submitted follo ng (1) ceipt of operating

' ense (2) ame dment to the licenseNovolving planne nerease i ower level, (3) installa on of fuel 1

l tirat has aMifferent (esign o has-been m ufactured a different uel supdkier, and (4) modiT(cations % hat may ha significaliNy altered t e nuclear, s

therma'1, or hydraulic performance o l

GRAND GULF-UNIT 1 6-15b Amendment No. 111 j

l j

j Enctaoure 2 to CNRO-94C356 -

Chapter 5 Page 25 INSERT 15bA

[ Add new programs per NUREG-1434 markup.]

l

-5.5.9 Safety Function Determination Program (SFDP) 5.5.10 Technical Specifications (TS) Bases Control-j l

INSERT 6-15b

m_

.~ - - _

Enc 6sure 2 to GNRO-94/00056 Chapter S Page 26 DESIGN FEATURES 5.5 METEOROLOGICAL TOWER LOCATION 5.5.1 The meteorological tower shall be located as shown on '

5.6 FUEL STORAGE CRITICALITY ~

3 5.6.1 The spent fuel storage racks are desipr 9

A maintained with:

E A'k,ff equivalent to less than r f,d when flooded with a.

unborated water, including al" as described in Section 9.)

g re,1 uncertainties and biases l

b.

A nominal 6.26-inch car' I

.h distance between fuel assemblies placed in the storage g

5.6.1.2 The k,ff for new f'

<st core loading stored dry in the spent fuel storage racks h i eed 0.98 when aqueous foaa moderation is assumed.

g DRAINAGE Y

q 5.6.2 The sper'

, pool is designed and shall be maintained to prevent-inadvertent v b

, pool below elevation 202'5 1/4".

i CAPACITY y

5.6.3 9

.41 storage capacity is designed and shall be maintained with a s'

/ II'It'd **:

6 e

e a re than 2324" spent fuel assemblies in the spent fuel pool, and.

p No more than 800 spent fuel as'semblies in the upper containment pool.

Ant of fuel in the upper containment pool is limited to temporary storage

.uel during refueling operations.

Prior to return to reactor criticality.

.1 spent fue shall be removed free the upper containment pool.

5. 7. COW ONENT CYCLIC OR TRANSIENT LINIT r.cr.s 5.7.1 The -

nents identifi'd N = : 1. $ 5

~

e anintained wi n the cyclic or transient limita. m._.: _ _ _ _ 4

.: u ihe pttysical-limit is 4348. The 2324 limit reflects the number of spent fuel assemblies that can be stored in the spent fuel pool without excessive reliance on RHR supplement cooling; i.e., for a time period in excess of a normal refueling duration.

'u A) } ct.,55(A L CePe.c % "Q dgn Mom sRAle GULF-UNIT 1 Amendment No.19 l' nsa G-Is b (Q

4 s

N TABLE 5.7.1-1

E l._ M2 N

E COMPONENT-CYCLIC OR TRANSIENT LIMITS x

.o S

T CYCLIC OR DESIGN CYCLE i

ch -

COMPONENT TRANSIENT LIMIT OR TRANSIENT

'N

\\

5

.i Reactor.

120 heatup and cooldown cycles 70 F to 560 F to 70 F w

80 step change cycles Loss of feedwater heaters 1

200 reactor' trip cycles 100% to 0% of RATED THERMAL POWER /

- 40 hydrostatic pressure or leak Pressurized to > 930 psig and tests 5 1250 psig f

'y

J e

M

- n d

. J.

O k

. i i'

?I

+

4 '

-$8 m

U

=

2...

.r

. E c

+

2 u

m t-v-

. ~.

~

.- to GNRO 94/0005e Chapter 6. Pags 28 APDLICABILITY 5.6,(o SURVEILLANCE REQUIREMENTS 4.0.1 Surveillance Requirements shall be met during th"

.t0NS er other conoitions specified for individual Limitin' unless otherwise stated in an individual Surveill-

' < ration 4.0.2 Each Surveillance Requirement shall P surveillance interval with a maximum all" 6ne specified the specified surveillance interval.

Lk. O to exceed 25% of 4.0.3 Failure to perform a Surv

b within the allowed surveillance interval, defin v 4

p11ance with the OPERABIL P p

4.0.2. shall constitute noncom-The time limits of the '

Qo a Limiting Condition for Operation.

jg

. are applicab'e at the time it is identified that a P 7

..nent has not been performed. The ACTION requirements mav ie surveillance co 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to pennit the completion of the are less *

. outage time limits of the ACTION requirements dga

<eillance Requirements do not have.to be perfonned on ine hh

.. OPERATIONAL CONDITION or other specified ap

.iot be made unless the Surveillance Requirement (plicables) as

.cing Condition for Operation have been perfonned within the

< surve111ance interval or as otherwise specified. This provision

.iot prevent passage through or to OPERATIONAL CONDITIONS as required to

.. ply with ACTION requirements.-

4.0.5 Surveillance Requirements for inservice inspection and testing of ASME '

Code Class 1. 2. & 3 components shall;be applicable as follows:

Inservice inspection of ASME Code Class 1. 2.-and 3 components and a.

inservice testing of ASME Code Class 1. 2. and-3 pumps and valves LA9 shall be perfonned in accordance with Section XI of the ASME Boiler d

and Pressure Vessel Code and applicable Addenda as required by 10 CFR

50. Section 50.55a(g), except where specific written relief has been-granted by the Commission pursuant to 10 CFR 50. Section 50.55a(g)

H (6)(1).

/

1 b.

Surveillance intervals specified in Section XI of the ASME Soiler and Pressure Vessel Code and applicable Addenda for the inservice inspection arid testing activities required by the ASME Boiler and' i

Pressure Yessel Code and applicable Addenda shall be applicable as

_ follows in.these Technical Specifications:

.l h5d (o-I5.b OD i

cu,nc c,utr-unit 1 4/4 0-b Amendment No. 77 -

ap[er 5 P e 29 APPLICABILITY SURVEILLANCE REQUIREMENTS (Continued)

ASME Boiler and Pressure Vessel Required frequencies Code and applicable Addenda for performing inservice terminology for inservice inspection and testing insoection and testino activities activities Weekly At least once per 7 days Monthly At least once per 31 days Quarterly or every 3 months At least once per 92 days Semiannually or every 6 months At least once per 184 days 5bq Every 9 months At least once per 276 days Yearly or annually At least once per 366 days The provisions of Specification 4.0.2 are applicable to the above required frequencies for performing inservice inspection and testing activities.

d.

Performance of the above inservice inspection and testing activities shall be in addition to other specified Surveillance Requirements.

Nothing in the ASME Boiler and Pressure Vessel Code shall be e.

construed to supersede the requirements of any Technical Specification.

f.

The inservice inspection program for piping identified in NRC Generic Letter 88-01 shall be performed in accordance with the NRC staff positions on schedule, methods, personnel, and sample expansion included in Generic Letter 88-01 or in accordance with alternate measures approved by the NRC staff.

~

- C n s e. x &

Co -I 5 b l'O GRAND GULF-UNIT 1

-3f4+3-Amendment No. 69,104

- - _ -.. to CNRO-94/00056 Chapter 5 Page 30 Mro m CON'a!NMENT SYSTEM 5 5, g, 4 CT5 c/,G, G,3 SURVEILLANCE REQUIREMENTS (Continued)

[At least once per 18 months or (1) after any structural maintenance c.

on the HEPA filter or charcoal adsorber housings, or (2) following 5~ G (o painting, fire or chemical release in any ventilation zone Mossunicatina with the__ subsysten bys

~

1.

Verifying that the subsystes satisfies the in place testing acceptan:e criteria and uses the test procedures of Regulatory 5.5.6.9 Positions C.5.a. C.5.c and C.5.6 of Regulatory Guide 1.52, c 10'.. '" 2' "*"Ch II70'

""d *** system flow rate is 4000 cfm

"'*I'i

5. 5.(o b h' 2MVerifyino within 31 davs after remova0that a ia. oratory anaiysis of a representative caroon sasole obtained in accordance with Regulatory Position C.6.6 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.

I'F 6

3.

Verifying a subsystem flow rate of 4000 cfm 15 during system s' M b operation when tested in accordance with ANSI k510-1975.

N h

E. /Af ter every 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months of charcoal adsorber operation)by verifying

~

witnin n days after renoval that a laboratory analysis of a repre-sentative carcon sample obtained in accordance with Regulatory T. 5. 6. C, Position C.6.b of Regulatory Guide 1.52 Revision 2, March'1978, meets the lateratory testing criteria of Regulatory Position C.6.a l

of Regulatory Guide 1.52 Revision 2, Mar:h 1978.

5~. 6 6 (d At least once per la months byp --@

/QW 1.

Perfoming a system functier.a1 test which includes simulated y;g automatic actuation of the system througnout its'energency operating secuence for the:

9.c. Y.,.e n a)

LOCA, and 3,L b)

Fuel handling accident.

2.

Verifying that the pMasure drop across the combined HEPA filters 5 '. T 6 d and charcoal adsorter banks is lass than 9.2 inches Water Gauge while operating the filter train at a flow rate of 4000 cfs 2-12.

3 Verifying that the filter train and isolation despers receive the Abs 5d appropriate actuation signal by each of the following test w%

conditions. For at least one of these test conditions, verify that the filter train starts and isciation despers open an

$ d on receipt of the actuation signal.

3.3 6.'A a.

Drywell pressure - high, i

b.

Reacter vessel water level - low low, level 2 I

c.

Fuel handling area ventilation exhaust radiation - high high, d.

Fuel handling area pool sweep exhaust radiation - high high, and e.

Manual initiation from the Control Room.

4.

Verifying that the fan can be manually started.

s, s.4 e 5.

Verifyin~g that the heaters dissipate 44

5.0 kW when tested l

in accoreence with ANSI M510-1975 (except for the phase balance criteria stated in Section 14.2.3).

SRAND GULF-UNIT 1 3/t 10 Amendment No.11, l ww +

(o-15 b C 5)

. =

cE$i7,U N"."$*"

CONTAINMENT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

I' g e.~ (AftereachcomofeteorpartialreplacementofaHEPAfilterbank]by verifying that the HEPA filter banks ove greater than or equal to g'g'6,q 99.95% of the 00P when they are teste q place in accordance with ANSI N510-1975 while operating the system at a flow rate of 4000 cfm i 10%.

f.

Afte n ach complete or partial replacement of a charcoal adsorbe g g, (o bank)by verifying that the charcoal adsorbers remove greater an 9.95% of a halogenated hydrocarbon refrigerant test gas when 5'.E. (o. h they are tested in place in accordance with ANSI N510-1975 while operating the system at a flow rate of 4000 cfm i 10%.

M 7 aid

'~

T}w p ros; 5 2 o n S R 3. 0. 2 gnJ $ R 3.D. 3 a e r

are 1 'c o bk t, tw VFTP 4ts q_ % ;q,3 l

1 l

O N=

(,- 151 M

GRAND GULF-UNIT 1 3/4-6 T. T. 6

$$I"rl PI.Y PLANT SYSTEMS 3/4.7.2 CONTROL ROOM EMERGENCY FILTRATION SYSTEM LIMITING C)NDITION FOR OPERATION f

3.7.2 fwo independent control room emergency filtration syster

.ia l i be OPERABLE.

APPLICABILITY:

All 0PERATIONAL CONDITIONS and *.

1 b

ACTION:

a.

In OPERATIONAL CONDITION 1, 2 or 3 with c d\\

..n emergency y

filtration subsystem inoperable, restor

,le subsystem to OPERABLE status within 7 days or be i C

. SHUTOOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTP

?

., following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

b.

In OPERATIONAL CONDITION 4, 5 o-u 1.

With one control room e" ation subsystem inoperable, restore the inoperab1' yp OPERABLE status within 7 days or initiate and mai-

.i of the OPERABLE subsystem in s

the isolation mot g[8} s 2.

With both cont ency filtration subsystems inoperable, suspend CORF nandling of irradiated fuel in the primary or

.tainment and operations with a potential for dra4 3

,or vessel.

c.

The provie

.ication 3.0.3 are not applicable in Operatir SURVEILLANCE r 4.7.2 Ea

.n emergency filtration subsystem shall be demonstrated OPERABl"

, once per 31 days on a STAGGERED TEST BASIS by initiating,

.ne control room, flow through the HEPA filters and charcoal srbers and verifying that the subsystem operates for at least s continuous hours with the heaters OPER_ABLE.

[leastonceper18monthsor(1)afteranystructuralmaintenancel b.

y on the HEPA filter or charcoal adsorber housings, or (2) following T. 6. 6 painting, fire or chemical release in any ventilation zone i

,.tcommunicating with the subsystem by:

/

1.

[0ELETE0]

Ahswh in &c% n W,7 f

When irradiated fuel is being handled in the primary or secondary containment.

InSCk C.-Is b (f7)

GRAND GULF-UNIT 1 4 /" 7 Amendment No. 69

.~_

Encbsure 2 to GNRO-94/CD56 -

Chapter 5 Page 33 PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) 2.

Verifying that the subsystes satisfies the in place testing 5

9 acceptance criteria and uses the test procedures of Regulatory 6.5.6,b 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 4000 cfs

1=.

gp _ _(3Terifyina within 31 days after remavahthat a laboratory a 3.

5.6. b c of a representative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, 4

March 1978, meets the laboratory testing critaria of Regulatory Position C.6.a of Regulatory Guide 1.52, Revision 2, March 1978.

i b'",

4.

Verifying a subsystem flow rate of 4000 cfm t 10% during subsystem operation when tested in accordance with ANSI-N510-1975.

Q~ fter everv 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months of charcoal adsorber operatjo3 by verifying 6,5. lo witnin 31 days af ter removal tnat a 1anorar.ory analysis of 'a repre-

~

sentative carbon sample obtained in accordance with Regulatory g'g,(o,C.

Positon C.6.h of Regulatory Guide 1.52, Revision 2, March 1978, meets the laboratory testing criteria of Regulatory Position C.6.a h

of Regulatory Guide 1.52, Revision 2 March 1978.

5~ T. Q T d.

Atleastonceper18monthsby) 1.

Verifying that the pressure drop across the combined MEPA filters y'g* 4 ' g and charcoal adsorber banks is less than 7.2 inches Water Gauge.

while operating the subsystes at a flew rate of 4000 cfm a 105.

E Verifying that the subsystes' receives an appropriate isolation actuation signal by each of the following test conditions. For McOOL5M at least one of the test conditions, verify that the subsystem 9%

automatically switches to the isolation mode of operation and 4,

the isolation valves close within 4 seconds.

3N (a) High high radiation in the outside air intake duct.

(b) High drywell pressure.

(c) Low low reactor water level, 'and (d) Manual initiation from the Control Room.

3 Verifying tha't the heaters dissipate 20.7 t 2.1 kW when tested e g' 6, q, in accordance with ANSI N510-1975 (except for the phase balance criteria stated in Section 14.2.3).-

ffter each complete or partial realme--t af a Maa ** t=e band by e.

vertrying snat tne nera rilter banks remove greater than or equal to y' y' ( ' 9 l99.95%ofthe00Pwhentheyaretestedinplaceinaccordancewith i ANSI N510-1975 while operating the systes at a flow rate of 4000 cfm

105.

g _ JAfter each complete or partial replac----t of a charcoal adsorbe f.

l brdk ey verirying that sne charcoal adsorbers remove n.-_.nas.d hydrocarbon ban of a -

1 s 5.6. b in place in accordance with AN5! M510-1975 while operating the system j

at a flow rate of 4000 cfm i 10%.

5 W+

fo - I Q ; ( R D 1/4,

Amendment No. 25 l

j GRAND GULF-UNIT 1

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

~

Enc 6sure 2 to GNRO44100056 Chapter 5 Page 34 3/4.11 RADI0 ACTIVE EFFLUENTS 3/4.11.1 LIOUID EFFLUENTS LIOUID HOLDUP TANKS LIMITING CONDITION FOR OPERATION y-g, 7 U.11.1.4 The quantity of radioactive material contained in any outside E 5.7.b ' temporary tank, not including liners for shipping radwaste, shall be limited

, to less than or equal to 10 curies, excluding tritium and dissolved or y

gtrained noble gases.

APPLICABILITY: At all times.

[ ACTION:

a.

With the quantity of radioactive material in any of the above specified tanks exceeding the above limit, immediately suspend all-additions of radioactive material to the tanks and within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months dl reduce the tank contents to within the limit, and describe the events leading to the condition in the next Annual Radioactive l

Effluent Release Report.

b.

The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE REQUIREMENTS 4.11.1.4 The quantity of radioactive material.:ontained in each of the above g g,3,3 specified tanks shall be determined to be withn the awve limit by analyzing arepresentativesampleofthetank'scontectJfatleastonceper7dayswhen)

Cracioactive materiars are oeing acaea to the tap Lwd p%vi si o n s o9 S R 3. O.1 cia d S R 3, o. 3 h f *5N t b5 M

SN oragt l

m p p )!c-e h h i b h %

L,x

% A cM MM w%n % ',g Pcqvom s-u m

%e s.

hsb (b G )%

GRAND GULF-UNIT 1 4/4 - l --

Amendment No. 59,87,111

$$ ice'Pge 5

RADI0 ACTIVE EFFLUENTS 3/4.11.2 GASEOUS EFFLUENTS

'l i

EXPLOSIVE GAS MIXTURE LIMITING CONDITION FOR OPERATION 5 6. T a 3.11.2.6 The concentration of hydrocen in the main cana amar offgas treatment systemshallbelimitedgolessthanorequalto4%byvolu]me -Q

[ APPLICABILITY: Whenever the main condenser offgas treatment systea fs in

~

operation.

ACTION:

a.

With the concentration of hydrogen in the sain condenser offgas treatment systes exceeding the limit, restors the concentration to within the limit within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months .

b.

The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE REQUIREMENTS 5'. 5.'7. q 4.11.2.6 The concentration of hydrown in the main condenser offgas treatment syctes shall be detenmined to be wit:11n the above limits by monitorina the' waste gas in the main condenser off-aan treat =nt tvst=Gw 'th%ydrogerd gnitor urtRABLE as requirea oy specif cation 3.3.T.1277 1

~fo ssN G-J5b (led GRAND GULF-UNIT 1 3/4 11 Amendment No. 87

Encbsure 2 to GtJRO 94/00050 Chapter 5 Pagt 30 ELECTRICAL POWER SYSTEMS hm bRVEILLANCEREQUIREMENTS(Continued)

US V.T",/,/,Q c,JAtleastonceper92 day)andfromnewoilpriortoadditiontothe E 5, E storage tanks by _veritVing that a sample obtained in accordance with ASTMfUuu-nos(reapproved1980Khasawaterandsedimentcontentof oLg7/

fless than or equal to.05 volume percent (and a kinemat_ic viscosity ___

j e,@ 40*C of greater than or equal to 1.9 but less tnan or equal to 4.1 I,when tested in accordance with_ ASTM-0975-77/ and an impurity leve

5. M c s of less than 2 mg. of insolubles_per 100 mT. when tested in accordance

)

\\with ASTM-02274-7f0xcept snat the test of new fuel for impuri i

1 snati oe performed within 7 days after addition nf the new_ fue j

to._the storace tanL/

l d.

At least once per 18 months, during shutdown, by:

1.

Subjecting the diesel to an inspection in accordance cedures prepared in conjunction with its manufactu-mendations for this class of standby service.

2.

Verifying the diesel generator capability to r' sf greater than or equal to 1200 kW (LPCS Pump)

.nerator 11, greater than or equal to 550 kW (RHR B/'

eesel generator 12, and greater than or equal tr

.5 Pump) for diesel generator 13 while maintainir e equal to i

75% of the difference between nominal '

4 overspeed trip setpoint, or 15% above nominal, less.

b' ect a load of 3.

Verifying the diesel generator capr

\\

s at least 5450 kW but not to exces of diesel generators 11 and 12 and 3300 kW for diese' s without tripping.

)

The generator voltage shall nr

(

, volts during and following the load rejection (O

Af'[

tself, and:

4.

Simulating a loss of offsi a)

For Divisions 1 anc' 1)

Verifying d-of the emergency busses and load shedr'

,[

emergency busser,.

2)

Verifyi' generato* stu ts on the auto-start signal 9

.ne emergency busses with permanently' conn sthin 10 seconds, energizes the auto-g swn loads through the load sequencer and

~

cor c

e

,reater than or equal to 5 minutes while its C

loaded with the shutdown loads. After ener-

.ne steady state voltage and frequency of the busses shall be maintained at 4160 1 416 volts p

. 1.2 Hz during this test.

b) an 3:

tfying de-energization of the emergency bus.

,erifying the diesel generator starts on the auto-start signal, energizes the emergency bus with the loads within 10 seconds and operates for greater than or equal to 5 minutes while its generator is loaded with the shutdown loads.

After energization, the steady state voltage and frequency of the emergency bus shall be maintained.at 4160 t 416 volts and 6011.2 Hz during this, test.

GRAND GULF-UNIT 1 3 /' O 2 --

L5M b16b (,ll)

Encbsure 2 to GNRO-94/00056 Chapter 5 Page 37 ADMINISTRATIVE CONTROLS

)

STARTUP REPORTS (Continued) f%2 The startup report shall address each of the tests identified in e

FSAR arnQhall include a description of the measured values of the op ing conditions or 4 aracteristics obtained during the test program an comparison x

of these values w'Tt esign predictions and specifications.,Any' corrective a

actions that were requi to obtain satisfactory operaMon shall also be described. Any additional s Qfic details requirf k id license conditions based on other commitments shall be incNed in th.s-re' port.

6.9.1.3 Startup reports shall be-subliiitte thin (1) 90 days following completion of the startup test' program, (2) 90 following resumption or commencement of co dal power operation, or (3) nths following initial criticality, whi ver is earliest.

If the Startup Repo oes not cover all three events

.e.,

initial criticality, completion of startu st program, and resumpti or commencement of commercial operation, supplementary orts shall

, be s tted at least every three months until all three events have b c

eted.

s /

1 ANNUAL REPORTS' 3g -

5.G. ) 6.9.1.4 Annual reports covering the activities of the unit as described below for the orevious calendar year shall be submitted prio_r to Marc _ of each year.

he initial report shall be suDmitted prior to March 1 of the year fo_llowing]

initial criticality.

y 6.9.1.5.1 Reports shall include a tabulation on an annual basis of the number g'g }

of station, utility, and other personnel, including contractors, receiving exposuresgreaterthan100 mrem /yragdtheirassociatedmanremexposure according to work and job functions, e.g., reactor operations and surveil-i lance, inservice inspection, routine maintenance, special maintenance (describe maintenance), waste processing, and refueling.

The dose assignments to various duty functions may be estimated based on pocket dosimeter, TLD, or film badge measurements.

Small exposures totalling less than 20 percent of the individual total dose need not be accounted for.

In the aggregate, at least 80 percent of the total whole body dose received from external sources should be assigned to specific major work functions.

y,G,d 6.9.1.5.2 Reports shall include documentation of all challenges to safety and relief valves.

D

[6.9.1.5.3 Reports shall include the results of specific activity analyses In3 which the primary coolant exceeded the limits of Specification 3.4.5.

The fol-I s

lowing information shall be included:

(1) reactor power history starting l

48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months prior to the first sample in which the limit was exceeded; (2) results,

' of the last isotopic analysis for radioiodine performed prior to exceeding the !

limit, results of analysis while the limit was exceeded, and results of one Qnalysis after the radioiodine activity was reduced to less than the limit FA single submittal may be made for a multiple unit station.

The submittalT

~

OA3-houldcombinethosesectionsthatarecmmtontoallunitsatthestation.)

5,g, \\

This tabulation supplements the requirements of 10 CFR 20.2206.

l 2/

GRAND GULF-UNIT 1 6-16 Amendment No. 287 111

SE'Ip'III*'eSo"[ Sa "

ADMINISTRATIVE CONTROLS ANNUAL REPORTS (Continued) 4 with each result including date and time of sampling and the ra 'oiodine s

con ntrat' ns; (3) cle nup syste Qlow history rting 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months rior to the firs ple which the imit was e). eeded; (4) g ph of the I-11 concentra-tion an gne et radiciodi isotope centration microcuries er gram

( as functf of t for the ation of e specific tivity above he Istea -state el; a (5) the t duratio when the sp ific activity of i the pr ary coo ant exceeded the ta oiodine l'mit.

L k

Aa l

\\

GRAND GULF-UNIT 1 6-16a Amendment No. 28 to GNRO-94/00056 Chapter 5 Page 39 ADMINISTRATIVE CQNTROLS ANNUAL RADIOLOGICAL ENVIRONMENTAL OPERATING REPORT 6,G. 2 6.9.1.6 The Annual Radiological Environmental Operating Report covering the operation of the unit during the previous calendar year shall be submitted before M (D of each year.

--5 6.9.1.7 he report shall include summaries, interpretations, and analysis of S G.~2 trends of the results of the Radiological Environmental Monitoring Program for the reporting period. The material provided shall be consistent with the objectives outlined in (1) the ODCH and (2) Sections IV.B.2, IV.B.3, and IV.C of Appendix I to 10 CFR Part 50.

dL seci Mk..

ANNUAL RADIOACTIVE EFFLVENT RELEASE REPORT T.(o.3 6.9.1.8 The Annual Radioactive Effluent Release Report covering the operation of the unit during the previous 12 months of operation shall be submitted M

_,uoO_3 of each year.

m -o. m.

g, L J 6.

The report shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the unit.

The material pro ided shall be (1) consistent with the objectives outlined in the ODCM and (2) in conformance with 10 CFR 50.36a and Section IV.B.1 of Appendix 1 td,10 CFR Part 50.

GRAND GULF-UNIT 1 6-17 Amendment No. 87 rlll

Enctosure 2 to GNRO-94/00056 Chapter 5 Page 40 INSERT 17A~

The Annual Radiological Environmental Operating Report shall include the results of analyses of all radiological environmental samples and.of all environmental radiation measurements taken during the period. pursuant to the locations specified in.the table and figures in the ODCM, as well as summarized and tabulated'results of these analyses' and measurements in the format of the table in the Radiological Assessment Branch Technical Position, Revision 1, November 1979.

The report shall identify the T L D -- r e s u l t s that represent collocated dosimeters in relation to the NRC TLD program and the exposure period associated with each result.

In the event that some individual results are not available for inclusion with the report, the report shall be submitted noting and explaining the reasons for the missing results.

The missing data shall.be submitted in a supplementary report as soon as possible.

ll 1

INSERT 6-17

ET.7eN 'P?le" k ADMINISTRATIVE CONTROLS SEMIANNUALRADI0ACTIVEEFFLUENTRELEASEREPORT(Continued) t i

Page intentionally left blank.

i l

GRAND GULF-UNIT 1 6-18 Amendment No. 87

$$.3 eIE."?f" ADMINISTRATIVE CONTROLS MONTHLY OPERATING REPORTS 6.9.1.10 Routine reports of operating statistics and shutdown experience, i

hq including documentation of all challenges to main steam system safety / relief valves, shall be submitted on a monthly basis no later than the 15th of each month following the calendar month covered by the report.

CORE OPE. RATING LIMITS REPORT (COLR) 6.9.1.11 Core operating limits shall be established prior to each reload 5,(o5g cycle, or prior to any remaining portion of a reload cycle and shall be documentedintheCOREOPERATINGLIMITSREPORT(COLR)fdor y

he tot 1 C The Average Planar Linear Heat Generation Rate (APLHGR) for Technical a.

Specification 3.2.1.

{l3 -

b.

The Minimum Critical Power Ratio (MCPR) for Technical Specification g

3.2.3.

The Linear Heat Generation Rate (LHGR) for Technical Specification c.

3.2 h

~

r,w The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC in the following documents.

The appropriate revision / supplement number for each document shall be identified in the Core Operating Limits Report.

1)

XN-NF-79-71(P), Exxon Nuclear Plant Transient Methodolooy for Boilina Water Reactors, Exxon Nuclear Company, Inc., Richland, WA. Approved by NRC letter dated October 24, 1986.

2)

XN-NF-80-19(P)(A), Volume 1, Exxon Nuclear Methodoloov for Boilina Water Reactors - Neutronic Methods for Desion and Analysis, Exxon Nuclear Company, Inc., Richland, WA.

3)

XN-NF-80-19(P)(A), Volume 1, Advanced Nuclear Fuels Methodoloov for Boilina Water Reactors: Benchmark Results for the CASMO-3G/MICR0 BURN-B Calculation Methodoloov, Advanced Nuclear Fuels Corporation, Richland, WA.

4)

XN-NF-80-19(P)(A), Volume 3, Exxon Nuclear Methodolooy for Boilino Water Reactors THERMEX: Thermal Limits Methodoloov Summarv Descriotion," Exxon Nuclear Company, Inc., Richland, WA.

5)

ANF-913(P)(A) Volume 1, GQTPASA?: A Comouter Procram for Boilina Water Reactor Transient Analysis, Advanced Nuclear Fuels Corporation, Richland, WA.

6)

ANF-ll25(P)(A), ANFB Critical Power Correlation, Advanced Nuclear Fuels Corporation, Richlar.d, WA.

7)

XN-NF-84-105(P)(A), Volume 1, XCOBRA-T: A Comouter Code for BWR Transient Thermal Hydraulic Core Analysis, Exxon Nuclear Company, Inc., Richland, WA.

GRAND GULF-UNIT 1 6-19 Amendment No. 80, 87, 106

~

- -.. - - ~. -.

Ericiosure 2 to GNRO-94/00056

"*S* d '

. ADMINISTRATIVE' CONTROLS' CORE OPERATING LIMITS REPORT (COLR) (Continued)

I 8)

XN-NF-573(P), EatiPEX Pellet-Clad Interaction Evaluation Code for Power Eggn, Exxon Nuclear Company, Inc., Richland,- WA. Approved by NRC letter dated August 28, 1990.

9)

XN-NF-81-58(P)(A),R00EX2:

Fuel Rod Thermal-Mechanical Resoonse Evaluation Model, E::xon Nuclear Company, Inc., Richland, WA.

10) XN-NF-85-74(P)(A), RODEX2A (BWR): Fuel Rod Thermal-Mechanical Resnonse Evaluation Model, Exxon Nuclear Company, Inc., Richland, WA.

11) XN-CC-33(P)(A) HUXY: A Generalized Multirod Heatuo Code with 10CFR50 Accendix K Heatuo Dotion, Exxon Nuclear Company, Inc., Richland,-WA.

j

12) XN-NF-825(P)(A), BWR/6 Generic Rod Withdrawal Error Analysis. MCPR_ for Plant Ooeration Within the Extended Operatino thin," Exxon Nuclear i

Company, Inc., Richland, WA.

I

13) XN-NR-81-51(P)(A), LOCA-Seismic Structural Resnonse of an Exxon Nuclear Comoany BWR Jet Pumo Fuel Assembly, Exxon Nuclear Company Inc., Richland, WA.

14) XN-NF-84-97(P)(A), LOCA-Seismic Structural Response of an ENC 9x9 BWR Jet Pumo Fuel Assembiv, Advanced Nuclear Fuels Corporation, Richland, WA.

15) XN-NF-86-37(P), Generic LOCA Break Snectrum Analysis for BWR/6 Plants, _

Exxon Nuclear Company, Inc., Richland, WA. Approved by NRC letter dated October 24, 1986.

16) XN-NF-82-07(P)(A), Exxon Nuclear comoany ECCS Claddina Swellina and Ruoture Model, Exxon Nuclear Company, Inc., Richland, WA.

17)

XN-NF-80-19(A), Volumes 2, 2A, 28, & 2C, Exxon Nuclear Methodoloav for -

Boilina Water Reactors EXEM BWR ECCS Evaluation Mode', Exxon Nuclear Company, Inc., Richland, WA.

18) XN-NF-79-59(P)(A), Methodoloav for Calculation of Pressure Dron in BWR Fuel Assemblies, Exxon Nuclear Company, Inc., Richland, WA.

The core operating limits shall be determined such that all applicable limits

_$45C (e.g., fuel thermal-mechanical limits, thermal-hydraulic limits, Emergency Core Cooling System (ECCS) limits, Nuclear limits such as' shutdown' margin, transient analysis limits, and accident limits) of the safety analysis are met.

n T4 The COLR, including any aid cycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC Document Control Desk with copies to the Regional Administrator and Residen* !Petat PECIAL REPORT OLA3 Special reports sha115 submitted to the Nuclear Regulatory Commission 6.9.2 pursuant to Section. 50.4 of 10 CFR Part 50 within the time period specified (foreachreport.

7 N

GRAND GULF-UNIT 1 6-19a Amendment No. 106

..~

,. -.. ~ -.

Encbsure 2 to GNRO44/00056 Chapter 5 Page 44

)

1 EMERGENCY CORE COOLING SYSTEMS LIMITING CONDITION FOR OPERATION (Continued)

CTS 3' 5' /

l ACTION:.(Continued)

For ECCS divisions 1 and 2, provided that P l

e.

OPERABLE and divisions 1 and 2 are other 1.

With one of the above require

,h

.able, restore the inoperable ADS valve +

sithin 14 days or be in at least HOT SHtF' g c, 4t 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and reduce

)

reactor steam done e 0

1g within the next 24

)

hours.

2.

With two o-c

. required ADS valves inoperable, s

e.g '

..,within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and reduce reactor i

be in s' stes-2 135 psig within the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

QC

, line " keep filled" pressure alan f.

With h

..inel inoperable, perfors Surveillance Requirement

,et once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

..S header delta P instrumentation channel inoperable,

, the inoperable channel to OPERABLE status with.72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months 'or.

,emine ECCS header delta P locally at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months

,(,herwise declare the associated ECCS inoperable.

H V n the e ont an ECCS system is ac ted and inj tswaterintothh R

tor C lant Syst a Special rt shall be pared and sub-03 tt to th Commissio ursuant to cification.

2 within

)

90 ay descri the ci stances of actuation d the total acc la d actua on cycles date. The nt value f the useag fac for e h affected fety inject n nozzle s 11 be i

govidedin.. his Spec 1 Report never its va exceeds 0.70.

i.

With an ADS accumulator low pressure alarm system instrunea' tion channel (s) inoperable, determine the associated ADS se"

{

pressure locally at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ; rest--

ble channel (s) to OPERABLE status within 7 days:

the associated ADS valves inoperable.

s\\ A

}\\O O

$a } s f

Whene -

. eas are inoperable, if unable to attain Cr d.

.,y this ACTION, maintain reactor coolant temperature g hy

., use of alternate heat removal methods.

at GRAN., GULF-UNIT 1 4/4-5 Amendment No.

21 Ls<.d Effective Date: GCT 2 019!i

(,-l% {0

Encbsure 2 to GNRO-94/C3066 Chapter 5 Page 45 a

ELECTRICAL POWER SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) 16.

Verifying that the following diesel generator locke prevent diesel generator starting and/or trip the xor only when required; a)

Generator loss of excitation.

b)

Generator reverse power.

c)

High jacket water temperature, d)

Generator overcurrent with volt b

e)

Bus underfrequency-(11 and 12 d ' 2 only).

@\\

f)

Engine bearing temperature

\\

12 only).

g)

Low turbo charger oil pre h)

High vibration (11 and 1)

High lube oil temper' C

.2 only).

j)

Low lube oil press-

\\O k)

High crankcase r' 1)

Generator groe (11 and 12 only).

l e.

At least once per 10 e any modifications which could affect diesel genc

[

endence by starting all three diesel generators sieu' g

ing shutdown, and verifying that the

,lerate to at least 441 rpe for diesel three diesel -

u p

generators

p A2 rps for diesel generator 13 in less than or equal o

f.

At le s years by:

4ch fuel oil storage tank, removing the accumulated 1

. and cleaning the tank using a sodium hypochlorite or

. lent solution, and srforming a pressure test of those portions of the diesel fuel oil system designed to Section III, subsection N0 of the ASME Code in accordance with ASME Code Section XI Article IWD-5000,

4. 8. J.1. 3 RepIrts-Alldieselgeneratorfailures,valdornon-vald,shal be re orted t the Commisslb in a SpechlReportpursuattoSpecifi tion 6.9.2 (thin 30 s.

Report f diesel benerator failure shall incl e the

informattu recomme ed in Regul ory Posit C.3.b of Reg atory Guide

.108, ReKsion1 ugust 19 If the n er of fa res in the la 100 valid tests on a Aer nuc ar unit ba

, is great than or al to 7, the eport sha 1 be supplemente to include e addition informat recommende in Regulatory (PositionT.3.bo Regulatory uide 1.108, evision 1.Jupt la?'

~_

LA3

~$ $scN (c,- I % b GRAND GULF-UNIT 1

-3/4 0 Amendment No. 6B l

Effective Date:/ Q er g i

S O :'"4.T " "

cts 3.3a.s TABLE 3.3.7.S-I (Continued)-

ACCIDENT MONLTORLNG INSTRUMENTATION ACTION STATEMENTS ACTION 80 -

a.

With the number of OPERA 8LE ace' s

d *9 instrumentation channels lest fed Number of Channels shown in Table 3.'

. the inoperable-channel (s) to OPERABLE s' aays or.be in at least HOT SHUTDOWN wit

. hours and be in COLD j[

SHUTDOWN within the-

. [

s.cident monitoring b.

With the number instrumentat' than the Minimum Channels OPERA 8LE r-(

sable 3.3.7.5-1, restore the i

s inopera>*

.o OPERA 8LE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months

^

or be

}

. SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months

l and

. 00WN within the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

ACTION 81 -

Wit' C

OPERABLE channels less than required by the

@ f

_d of monitoring.the appropriate parameter (s)

. OPERABLE requirements, initiate the preplanned r

..o aurs, and:

1

.ther restore the inoperable channel (s) to OPERA 8LE status within 7 days of the event, or b

p pare and bait a scial Ryort the C ission L43 pur ant.to Sp ificat 6.9.2 thin 4 day following ee t outlin the a ion ta

, the ause f the in era lity and plans nd sch ule fo rett ing the syst to PERABLE s tus, f

ACTION 82 -

For OPERATIONAL CONDITIONS 1, 2. 3 a.

With the number of OPERABLE accider' instrumentation channels less th eber of Channels shown in Table 3.3.'

inoperable channel (s) to OPERABLE st-or be in at 4.$

least HOT SHUTDOWN wit'

..ars and be in COLD i

SHUTOOWN within the b.

With the numb-gC

..: dent monitoring

$g instrumen+

,s' than the Minimum Channels OPERAB'"

iable 3.3.7.5-1, restore the ino" co OPERA 8LE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months V

.0T SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months

.,HUTDOWN within the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

a, t,0NDITIONS 4, 5

..aber of OPERABLE accident monitoring instrumentation

, less than required by the Minimum Channels OPERA 8LE

.rement, either restore the inoperable channel (s) to To sq.d ( 3)

V/%

GRAND GULF-UNIT 1

-3/4 MS -

Amendment No. 103

SEOpI.$ 'rSg"." f" "2

ADMINISTRATIVE CONTROLS 6710 RECORD RETENT,1Qfi In additioho4he applicable record retention requirements of e 20, Code of Federal Regullitiets, the following records shall be re ed for at least the minimum period indi ted.

6.10.1 The following records s a 1 be retal for at least five years:

a.

Records and logs of unit ra n covering time interval at each power level, b.

Records an s of principal maintenance ivities, inspections, repair rep acement of principal items of a ment related to n

ar safety.

/ All REPORTABLE EVENTS.

LAI 3

GRAND GULF-UNIT 1 6-19b Amendment No.106

}

Encbsure 2 to CNRO-94/C0056

" ' * ' ' "'" ' 8 ADMINISTRATIVE CONTROLS es10 RECORD RETENTION (Continued) d.

Records of surveillance activities, inspections and calibratio required by these Technical Specifications, e.

Records of changes made to the procedures required by Spec fication 6.8.1.

f.

R ords of radioactive shipments.

g.

Reco s of sealed source and fission detector leak te ts and result h.

Records o annual physical inventory of all seale source material of record.

6.10.2 The following re rds shall be retained for the dura ion of the Unit Operating License:

a.

Records and dra 'ng changes reflecting unit de ign modifications made to systems a equipment described in th Final Safety Analysis Report.

b.

Records of new and irr diated fuel invento y, fuel transfers and assembly burnup histori O

c.

Records of radiation expos e as requi d by 10 CFR Part 20.

d.

Records of gaseous and liquid adio tive material released to the environs.

e.

Records of transient or operatio cycles for those unit components identified in Table 5.7.1-1.

f.

Records of reactor tests and xperimen s.

g.

Records of training and q ification fo current members of the unit staff.

h.

Records of in-service spections performed ursuant to these Technical Specificati ns.

i.

Records of Quality rsurance activities require by the Operational Quality Assurance anual not listed in Section 6. 0.1.

j.

Records of revi ws performed for changes made to p 'cedures or equipment or views of tests and experiments pursu t to 10 LFR 50.59.

k.

Records meetings of the PSRC and the SRC.

1.

Recor of the service lives of all hydraulic and mechani al snub ers including the date at which the service life comm nces and as ciated installation and maintenance records.

m.

ecords of analyses required by the radiological environmenta monitoring program.

n Records of reviews performed for changes made to the OFFSITE DOS CALCULATION MANUAL and the PROCESS CONTROL PROGRAM.

GRAND GULF-UNIT 1 6-20 Amendment No. Eb B7rlli

"pf0 f "

ADMINISTRATIVE CONTROLS 6,11 RADI Al lur4 FR0itti10N PROGRAM 6.

1 Procedur for personne adiation prot tion shall be prepared consis nt with the equirements o O CFR Part 2 and shall be approved, maintain and adhere to for all ope 4tions involv 99 personnel radiation

}

1.12 HIGH RADIATION AREA 6.

.1 In lieu of the " control device" or " alarm signal" required by 10 20.1 1(a) each high radiation area in which the intensity of radiatio s

greate than 100 mrem /hr but less than 1000 mrem /hr shall be barrica d and conspic usly posted as a high radiation area and entrance theret shall be contro11e by requiring issuance of a Radiation Work Permit (R

.* Any individual group of individuals permitted to enter such a as shall be provided with or accompanied by one or more of the followi a.

A ra 'ation monitoring device which conti ously indicates the radiat'on dose rate in the area.

b.

A radiati monitoring device which ntinuously integrates the radiation se rate in the area an alarms when a preset integrated d e is received.

En y into such areas with this monitoring dev' e may be made a er the dose rate level in the area has been e ablished and ersonnel have been made gg knowledgeable of em.

c.

A health physics qua fi individual, i.e., qualified in radiation protection p edures, with a radiation dose rate monitoring device, wh responsible for providing positive control over the ac 'viti s within the area and shall perform periodic radiation surveil nce at the frequency specif;sd by the unit Health Phys ist in the Radiation Work Permit.

6.12.2 In addition to t requirements of

.12.1, areas accessible to personnai with radiati levels such that a or portion of the body could receive in one hour a ose greater than 1000 m m shall be provided with locked doors to pre nt unauthorized entry, and e keys shall be maintained under the adminis ative control of the Shift Supe intendent on duty and/or the unit Radiat' n Control Supervisor.

Doors shall main locked except duri g period of access by personnel under an approve RWP which shall n

specify th ose rate levels in the immediate work area d the maximum allowabl stay time for individuals in that area.

For in 'vidual areas accessi e to personnel with radiation levels such that a ma' r portion of the body ould receive in one hour a dose in Health Physics personnel or personnel escorted by Health Physics per nel shall be exempt from the RWP issuance requirement during the performan of their assigned radiation protection duties, provided they are otherwise following plant radiation protection procedures for entry into high radi ion areas.

GRAND GULF-UNIT 1 6-21 Amendment No. lli j

EO *';0.T" ADMINISTRATfVE CONTROLS 6.12 HIGH RADIATION AREA Continued)

LA2 Exces of 1000 m m*thatarelocatedwitlEinlargeareas,suchasthecontainment, where enclosure xists for p rposes of lockin,g, and no enclosure can be

- reasonab construct around th individual areas then that area'shall be roped o f, consp uously po ed and a fla inglightshallsbeactivatedas'awarning s

j de ce.

In ieu of the tay time spe fication of th'e RWP, continuous ' surveillance f dire or rem te, such as se of close ircuit TV cameias, may be made tiyspersonne quali ed in r diation prot ction procedures to provide positive exposure control (overtheactivitieswithint area.,

6%L PROCESS CONTROL PROGRAM (PCP)

Changes he PCP:

a.

Shall cumented and records of reviews per retained as require pecification 6.10.2n.

Th cumentation shall contain:

gj 1)

Sufficient info

_ tion t port the change together with the appropriate analys evaluations justifying the change (s) and 2)

A de nation that the ch e will maintain the overall nformance of the solidified w e product to existing requirements of Federal, State, or her applicable regulations, b.

Shall become effective after review and accept e by the PSRC and the approval of the General Manager, Plant Operations.

5.5.1 6.14 0FFSITE DOSE CALCULATION MANUA_ (0DCM)

^

p q Tn % d C75 l.RM Changes to the ODCM:

C a.

Shall be documented and records of reviews performed shall be retained g "n" M d by Specific: tina 6.10r29.

This documentation shall contain:

L 1)

Sufficient information to support the change together with the appropriate analyses or evaluations justifying the change (s) and 2)

A determination that the change will maintain the level of radioactive effluent control required by 10 CFR 20.1302, 40 CFR l

Part 190, 10 CFR 50.36a, Appendix I to 10 CFR Part 50 and not adversely impact the accuracy or reli_ab it of effl nt, dose or setpoint calculations.

{g, ag Shall become effective after review and acceptaRe616 ^3E e[he b.

approval of theLGeneral Manaoer. P_lant Ooeratinn { _

c.

Shall be submitted to the Commission in the form of a complete, legible copy of the entire ODCM as a part of or concurrent with the Annual ji Radioactive Effluent Release Report for the period of the report in which any change to the ODCM was made.

Each change shall be identified by markings in the margin of the affected pages, clearly indicating the area of the page that was changed, and shall indicate the date (e.g.,

month / year) the change was implemented.

Qeasurementmadeat30cmfromsourceofradioactivity.

l GRAND GULF-UNIT 1 6-22 Amendment No. Bh111 to GNRO 94/00060 Chapter 5 Page 51 DEFINITIONS PRIMARY CONTAINMENT INTEGRITY 1.31 PRIMARY CONTAINMENT INTEGRITY shall exist when:

a.

All containment penetrations required to

.g accident conditions are either:

1.

Capable of being closed by r

.inment automatic isolation system, or

\\

2.

Closed by at least or aI blind flange, or deactivated automatic valve se'

. sed position, except for valves that are opened (f.

ative control as permitted by Specification b.

The containmer ch is closed and sealed, s

c.

Each cont is in compliance with the requirements of Specif' d.

TF eakage rates are within the limits of Specification

_ssion pool is in compliance with the requirements of cation 3.6.3.1.

sealing mechanism associated with each primary containment

,e penetration; e.g., welds, bellows or 0-rings, is OPERABLE.

PROCESS CONTROL PROGRAM (PCP) b1 1.32 The PROCESS CONTROL PROGRAM (PCP) shall contain the current formulas, sampling, analyses, test, and determinations to be made to ensure that proces-,

sing and packaging of solid radioactive wastes based on demonstrated processin of actual or simulated wet solid wastes will be accomplished in such a way as to assure compliance with 10 CFR Parts 20, 61, and 71, State regulations, burial ground requirements, and other requirements governing the disposal of solid radioactive waste.

~

PURGE - PURGING 1.33 PURGE or PURGING is the controlled a-confinement to maintain temperat"-

1

.,,,e r operating condition, in - '

CC~

,, required to purify the conf

C'ba h#

n.-

6p C

. anall be a total reactor core heat transfer rate to the

,,. J833 MWT.

Insk_h GRAND GULF-UNIT 1

$Y C O Amendment No. 87,102

E U 8 fs ' N."$ f "

DEFINITIONS MINIMUM CRITICAL

/)M eg-S5d In 1.25 The MINIMUi

('.Q hr

.I the smallest CPR which exists in the e g, g, ) 0FFSITE DOSE CALCULATION MANUAL (00CM) 1.26 The OFFSITE DOSE CALCULATION MANUAL (00CM) shall contain the methodology and parameters used in the calculation of offsite doses resulting from radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring Alarm / Trip Setpoints, and in the conduct of the Environmental Radiological Monitoring Program. The ODCM shall also contain l) the Radioact ve Effluent Controls pd aicicicFcat-hvirc=_^nta}A to gZbri d= Progra t p y:1 b Sectier 6.q_J Mnd (2) descriptions of the information that shoulc >be included in the Annual Radiological Environmental O@]tspecifications6.9.1.0,OpTratfng and TnnuaT Ra'dioactive Effluent Release Repo l

6.9.1./, b.v.1.o and 0.v g 0PERABLE - OPERABILITY 1.27 A system, subsystem, train, component or device shall '

have OPERABILITY when it is capable of performing its specifir

.nd when all necessary attendant instrumentation, controle

ser, cooling or seal water, lubrication or other auxilia"

.c are required for the system, subsystem, train, compor-

.e perform its

.et function (s).

function (s) are also capable of performing thei OPERATIONAL CONDITION - CONDITION 1.28 An OPERATIONAL CONDITION, i.e 4

.all be any one inclusive combination of mode switch posit' reactor coolant temperature as specified in Table 1.2.

V f

PHYSICS TESTS 1.29 PHYSICS TESTS v

'f ests performed to measure the fundamental nuclear character'

. actor core and related instrumentation and

1) described ir one FSAR, 2) authorized under the provisions of 10 CFR 50.59

,e approved by the Commission.

&g PRESSURr 1.'

.JARY LEAKAGE shall be leakage through a non-isolable fault-Aant system component body, pipe wall or vessel wall.

GRAND GULF-UNIT 1

.1-5 Amendment No. Bh 111

,aeA J >-

n sa y

a M

ATTACHMENT 1B CTS - PSTS COMPARISION DOCUMENT DISCUSSION OF CHANGES l

. _ ~

... ~.

~. -

Enc 6sure 2 to GNRO-94/00056 '

Chapter 5 Page 54 i

DISCUSSION OF CHANGES CTS: 6.1 - RESPONSIBILITY ADMINISTRATIVE A.1 Where possible, plant specific management position titles in the current Technical Specifications are replaced with generic titles as provided in ANSI /ANS 3.1.

Personnel who fulfill these positions are required to meet specific qualifications as detailed in proposed Specification 5.3, and compliance details relating to the plant specific management position titles are identified in licensee controlled documents (such as the UFSAR).

The two major specific replacements are the generic " plant manager" for the manager level individual responsible for the overall safe operation of the plant and the generic descriptive use of "the corporate executive responsible for overall plant nuclear safety" in place of the Vice President position.

The plant specific titles fulfilling the duties of these generic positions will continue to be defined, established, documented and updated in a plant controlled document with specific regulatory review requirements for changes, such as the UFSAR or OQAM.

This approach is consistent with the intent of Generic Letter 88-06 which recommended, as a line item improvement, relocation of the corporate and unit organization charts to licensee controlled documents.

The intent of the Generic Letter, and of this proposed change, is to reduce the unnecessary burden on intC and licensee resources being used to process changes due solely to personnel titles changes during reorganizations.

Since this change does not eliminate any of the qualifications, responsibilities or requirements for these personnel or the positions, the change is considered to be a change in presentation only and is therefore' administrative.

A.2 Lines of authority are also required to be defined and maintained in the UFSAR by TS 6.2.la.

Repeating the organizational responsibilities via an internal management directive only increases the administrative burden on the facility with no resulting benefit.

Since the actual lines of authority are not affected by this change and since the requirement to define these lines of authority in the UFSAR will remain in affect this is considered an administrative change.

RELOCATED SPECIFICATIONS None in this section.

TECHNICAL CRANGES - MORE RESTRICTIVE None in this section.

Endosure 2 to GNRO 9400050 Chaplet 5 Page 65 I

DISCUSSION OF CHANGES CTS: 6.1 RESPONSIBILITY TECHNICAL CHANGES - LESS RESTRICTIVE None in this section, j

t

)

Encbsure 2 to CNRO-94205E.

Chapler 5 Page 56 DISCUSSION OF CHANGES CTS: 6.2 - ORGANIZATION ADMINISTRATIVE A.1 The requirement to update the UFSAR is contained in Title 10 of the Code of Federal Regulations.

Repeating the requirements of the regulations is redundant and unnecessary, and creates an unnecessary burden to revise the Technical Specifications when the regulations change.

Therefore, this requirement is not repeated in the proposed Technical Specifications.

Since the requirements remain, this change only affects the presentation method and is considered administrative.

A.2 This change provides clarifying examples that constitute the accepted shift period and " foreseen" circumstances during which substantial overtime is acceptable.

Since these are only examples and do not change the actual requirements, this change is considered administrative.

A.3 This comment number is not used for this station.

A.4 Where possible, plant specific management position titles in the current Technical Specifications are replaced with generic titles as provided in ANSI /ANS 3.1.

Personnel who fulfill these positions are required to meet specific qualifications as detailed in proposed Specification 5.3, and compliance details relating to the plant specific management position titles are identified in licensee controlled documents (such as the UFSAR).

The two major specific replacements are the generic " plant manager" for j

the manager level individual responsible for the overall 1

safe operation of the plant and the generic descriptive-use of "the corporate executive responsible for overall plant nuclear safety" in place of the Vice President position.

The plant specific titles fulfilling the duties of.these generic positions will continue to be defined, established, documented and updated in a plant controlled document with specific regulatory review requirements for changes, _such as the UFSAR or OQAM.

This approach is consistent with the intent of Generic Letter 88-06 which recommended, as a line item improvement, relocation of the corporate and unit organization charts to licensee controlled documents.

The intent of the Generic Letter, and of this proposed change, is to reduce the unnecessaJy burden on NRC and licensee resources being used to process changes due solely to j

personnel' titles changes during reorganizations.

Since chis change does not eliminate any of the qualifications, responsibilities or requirements for these personnel or the positions, the change is considered to-be a change.in presentation only and is therefore administrative.

Encbsure 2 to GNRO-942006 Chapter 5 Page 57 j

DISCUSSION OF CHANGES CTS: 6.2 - ORGANIZATION ADMINISTRATIVE (continued)

A.5 The details of the methods for implementing this requirement are contained in the Commission's Policy Statement.

Repeating these details is redundant and unnecessary, and the omitted information provides no additional restriction or requirement.

Therefore, this requirement is not repeated in the proposed Technical Specifications.

Since the requirements remain the same, this change only affects the presentation method and is considered administrative.

A.6 This comment number is not used for this station.

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

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

A.9 This comment number is not used for this station.

A.10 The requirement for an SRO to be present during fuel handling and core alterations is contained in 10 CFR 50.54.

Therefore, there is no need to repeat these requirements in the Technical Specifications.

Since the requirements remain the same, this change only affects the presentation method and is considered administrative.

RELOCATED SPECIFICATIONS None in this section.

TECHNICAL CHANGE - MORE RESTRICTIVE M.1 This comment number is not used for this station.

M.2 This comment number is not used for this station.

M.3 This comment number is not used for this station.

TECHNICAL CHANGE - LESS RESTRICTIVE

" Generic" LA.1 Details of the Operator License requirements for these i

specific positions are relocated to the UFSAR and procedures although a requirement that the Operations Superintent or the Manager Plant Operations have an SRO License is retained in 5.2.2.f.

Additionally, changes to the procedures and the UFSAR are controlled in accordance with 10 CFR 50.59.

I I

- to GNRO-94/0005G Chapler 5 Page 58 DISCUSSION OF CHANGES CTS: 6.2 - ORGANIZATION LA.2 This comment number is not used for this station.

LA.3 The staffing requirements of Table 6.2.2-1 are relocated to the UFSAR and procedures. The requirements of Table 6.2.2-1 are removed for the Technical Specifications and will be controlled by the licensee's administrative controls. 10 CFR 50.54 provides the requirements for shift complement regarding licensed operators.

Additionally, the Technical Specifications will continue to specify when a licensed operator must be in the control. room.

The Table 6.2.2-1 requirements associated with the auxiliary operators are retained as 5.2.2.a with the associated allowance for unexpected absences retained in 5.2.2.c.

Changes to the procedures and the UFSAR are controlled in accordance with 10 CFR 50.59.

LA.4 This comment number is not used for this station.

LA.5 The requirements relating to the Independent Safety Engineering Group (ISEG) is relocated to the UFSAR and procedures.

Changes to the procedures and the UFSAR are controlled in accordance with 10 CFR 50.59.

LA.6 This comment number is not used for this station.

" Specific" L.1 This comment number is not used for this station.

L.2 This comment number is not used for this station.

- - _ _. - to GNRO-04/00056 Chapter 5 Page 50 DISCUSSION OF CHANGES CTS: 6.3 - UNIT STAFF QUALIFICATIONS ADMINISTRATIVE A.1 This requirement is contained in 10 CFR 50.

Therefore, this requirement is not repeated in the proposed Technical Specifications.

Since the requirements remain the same, this change only affects the presentation method and is i

considered administrative.

RELOCATED SPECIFICATIONS None in this section.

TECHNICAL CHANGES - MORE RESTRICTIVE None in this section.

LESS RESTRICTIVE

]

TECHNICAL CHANGES l

" Generic" LA.1 The requirement that the training program be accredited by INPO is relocated to the UFSAR and procedures.

Changes to the procedures and.the UFSAR are controlled.in accordance with 10 CFR 50.59.

LA.2 The training requirements relating to the Independent Safety Engineering Group (ISEG) are relocated to the UFSAR and procedures.

Changes to the procedures and the UFSAR are controlled in accordance with 10 CFR 50.59.

q

" Specific a L.1 This comment number is not used for this station.

. _ _. _.= _ -_.

m Ecciosure 2 to G4RO 94/00056 Chapter 5 Page 60 DISCUSSION OF CHANGES CTS: 6.4 - TRAINING ADMINISTRATIVE A.1 This comment number is not used for this station.

A.2 This comment number is not used for this station.

A.3 This comment number is not used for this station.

RELOCATED SPECIFICATIONS None in this section.

TECHNICAL CHANGES - MORE RESTRICTIVE None in this section.

TECHNICAL CHANGES - LESS RESTRICTIVE

" GENERIC" l

LA.1 The requirements relating to the retraining and replacement program are relocated to the UFSAR and procedures.

Changes to the procedures and the UFSAR are controlled in accordance with 10 CFR 50.59.

" SPECIFIC" i

None in this section.

1 1

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Enc 6sure 2 to GNRO-9400056 Chapter 6 Page 61 DISCUSSION OF CHANGES CTS: 6.5 - REVIEW AND AUDIT ADMINISTRATIVE A.1 This comment number is not used for this station.

A.2 This comment number is not used for this station.

A.3 This. comment number is not used for this station.

A.4 This comment number is not used for this station.

I A.5 This comment number is not used for this station.

A.6 This comment number is not used for this station.

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

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

A.9 This comment number is not used for this station.

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.

RELOCATED SPECIFICATIONS None in this'section.

TECHNICAL CHANGE - MORE RESTRICTIVE M.1 This comment number is not used for this station.

M.2 This comment number is not used for this station.

M.3 This comment number is not used for this station.

M.4 This comment number is not used for this station.

M.5 This comment number is not used for this station.

TECHNICAL CHANGE - LESS RESTRICTIVE

" Generic" LA.1 This comment number is not used for this station.

~..

Enc 6sure 2 to CNRO-94/00056 Chapter S Page 62 DISCUSSION OF CHANGES CTS: 6.5 - REVIEW AND AUDIT TECHNICAL CHANGE - LESS RESTRICTIVE (continued)

LA.2 This comment number is not used for this station.

LA.3 This comment number is not used for this station.

LA.4 Review and Audit requirements are relocated to the.UFSAR and procedures.

Changes to the procedures and the UFSAR are controlled in accordance with 10 CFR 50.59.

" Specific" L.1 This comment number is not used for this station.

L.2 This comment number is not used for this station.

L.3 This comment number is not used for this station.

L.4 This comment number is not used for this station.

,a

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.. - -. - - - - _ _ _ _ _ to CNRO-94/00056 Chapter 5 Page 63 DISCUSSION.0F CHANGES CTS: 6.6 - REPORTABLE EVENT ACTION ADMINISTRATIVE A.1 This requirement is contained in Title 10 of the Code of Federal Regulations.

Repeating the requirements of the regulations is redundant and unnecessary, and creates an j

unnecessary burden to revise the Technical Specifications 1

when the regulations change.

Therefore, this requirement is not repeated in the proposed Technical Specifications.

Since the requirements remain the same, this change only affects the presentation method and is considered administrative.

l A.2 This comment number is not used for this station.

RELOCATED SPECIFICATIONS None in this section.

TECHNIfAL CHANGES - MORE RESTRICTIVE None in this section.

LESS RESTRICTIVE TECHNICAL CHANGES aGeneric" LA.1 Review and Audit requirements are' relocated to the UFSAR and procedures.

Changes to the procedures and the UFSAR are controlled in accordance with 10 CFR 50.59.

" Specific" L.1 This comment number is not used for this station.

-. ~ -.--

Enc 6sure 2 to CNRO-94/00056 Chapter 5 Page 64 DISCUSSION OF CHANGES SAFETY LIMIT VIOLATIONS CTS: 6.7 ADMINISTRATIVE A.1 The technical content of this requirement is being moved to another chapter of the proposed Technical Specifications in accordance with the format of the BWR Standard Technical Specifications, NUREG-1434.

Any technical changes to this requirement will be addressed with the content of the-proposed chapter location.

RELOCATED SPECIFICATIONS None in this section.

I TECHNTCAL CHANGES - MORE RESTRICTIVE None in this section.

a TECHNICAL CHANGES - LESS RESTRICTIVE

" Generic a LA.1 Review and Audit requirements are relocated to the UFSAR and procedures.

Changes to the procedures and the UFSAR are controlled in accordance with 10 CFR 50.59.

" Specific" None in this section.

-w..-

r 4

.m+

3

Encbsure 2 to GNRO-W00056 Chapter 5 Page 65 DISCUSSION OF CHANGES PROCEDURES AND PROGRAMS CTS: 6.8 ADMINISTRATIVE A.1 These types of procedures are required by the item immediately preceding which references Regulatory Guide 1.33.

Therefore, it is not necessary to specifically identify each type of procedure.

Since the requirements remain, this is considered to be a change in the method of presentation only, and therefore, is considered an administrative change.

A.2 Procedures to implement the Emergency Plan and the Security Plan are required by 10 CFR 50, Appendix E and 10 CFR 50.54 (p).

Since conformance with 10 CFR Chapter I is a license condition and the Emergency Plan and Security Plan are required to be implemented by 10 CFR Chapter I, specific identification of these plans is unnecessary duplication.

This is a change in the presentation of the requirements only, and therefore, is considered an administrative change.

A.3 The specific program is covered by a more generic item which requires this activity for all Programs and Manuals.

Therefore, it is not necessary to specifically identify each program.

Since the requirements remain, this is considered to be a. change in the method of presentation.only, and therefore, is considered an administrative change.

A.4 The technical content of several requirements are being j

moved from another chapter of the current Technical Specifications and are proposed to be identified as Programs in accordance with the format of the BWR Standard Technical Specifications, NUREG-1434.

Other Programs currently identified in the Administrative Controls section are consolidated into this section.

Any technical changes to the requirements are identified in their respective markups and addressed as indicated.

These Programs include:

PSTS CIS 5.5.1 6.14 & 1.26 Offsite Dose Calculation Manual 5.5.2 6.8.3.a Primary Coolant Sources Outside Cont.

5.5.3 6.8.3.c Post Accident Sampling 5.5.4 6.8.3.d Radioactive-Effluent-Controls Program 5.5.5 5.7.1 Component Cyclic or Transient Limit 5.5.6 4.6.6.3

& 4.7.2 Ventilation Filter Testing Program 5.5.7 3/4.11.1.4 Explosive Gas and Storage Tank

& 3/4.11.2.6 Radioactive Monitoring Program 5.5.8 4.8.1.1.2.c Diesel Fuel Oil Testing Program

.~

Enctosure 2 to GNRO-9400056 Chapter $ Page NI DISCUSSION OF CHANGES CTS: 6.8 PROCEDURES AND PROGRAMS ADMINISTRATIVE (continued)-

A.5 This comment number is not used for this station.

A.6 This comment number is not used for this station.

A.7 An statement of applicability of SR 3.0.2 or SR 3.0.3 is needed to maintain the current allowances for surveillance frequency extensions since these SRs are not normally applied to frequencies identified in the Administrative Controls section of the Technical Specifications.

Since this change maintains current requirements, it is considered a change of presentation method only.

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

A.9 This comment number is not used for this station.

RELOCATED SPECIFICATIONS None in this section.

TECHNICAL CHANGES - MORE RESTRICTIVE M.1 A requirement is included to maintain Emergency Operating Procedures (EOPs) as implemented in response to NUREG-0737.

Although EOPs are included as a necessary procedure type in Regulatory Guide 1.33, the additional procedures and changes made in response to the guidance provide in NUREG-0737 and Supplement 1 are not currently included.

This change assures these commitments, as made in response to Generic Letter 82-33, are maintained, and that the guidance and commitments are appropriately considered for any changes to these procedures.

M.2 Two new programs are included in the proposed Technical Specifications.

These programs include:

5.5.9 Safety Function Determination' Program S.5.10 Technical Specification Bases Control The Safety Function Determination Program is included to support implementation of the support system operability-characteristics of the Technical Specifications.

The Bases Control program is provide to specifically delineate the appropriate methods and reviews necessary for a change to the Technical Specification Bases.

. _... _ to GNRO-94/00056 Chapter 5 Page 67 DISCUSSION OF CHANGES CTS: 6.8 PROCEDURES AND PROGRAMS TECHNICAL CRANGE - MORE RESTRICTIVE (continued)

M.3 This comment number is not used for this station.

TECHNICAL CHANGES - LESS RESTRICTIVE

" Generic" LA.1 Details of the methods for implementing this specification are relocated to the UFSAR and procedures.

The guidance documents which dictate the methods are also identified in the UFSAR.

Additionally, changes to the procedures and the UFSAR are controlled in accordance with 10 CFR 50.59.

LA.2 Details of the methods for implementing this specification are relocated to the UFSAR, procedures and the proposed Administrative Controls section of the Technical Specifications as a Program.

The guidance documents which dictate the methods are also identified in the UFSAR.

Additionally, changes to the procedures and the UFSAR are controlled in accordance with 10 CFR 50.59, and the Program requirements are controlled as a proposed Technical Specification.

LA.3 This comment number is not used for this station.

LA.4 Review and Audit requirements are relocated to the UFSAR and procedures.

Changes to the procedures and the UFSAR are controlled in accordance with 10 CFR 50.59.

LA.5 This comment number is not used for this station.

LA.6 Requirements relating to the Process Control Program are relocated to the UFSAR and procedures.

Changes to the procedures and the UFSAR are controlled in accordance with 10 CFR 50.59.

LA.7 Requirements relating to the In-Plant Radiation Monitoring Program are relocated to the UFSAR and procedures.

Changes to the procedures and the UFSAR are controlled in accordance with 10 CFR 50.59.

LA.8 Requirements relating to the Radiological Environmental Monitoring Program are relocated to the UFSAR-and procedures.

Changes to the procedures and the UFSAR are contiolled in accordance with 10 CFR 50.59.

LA.9 Requirements relating to the inservice inspection and

.._m Enc 6sure 2 to GNRO-94100056 Chapter 6 Page 68 DISCUSSION OF CHANGES CTS: 6.8 - PROCEDURES AND PROGRAMS testing are relocated to the UFSAR and procedures.

Changes to the procedures and the UFSAR are controlled in accordance with 10 CFR 50.59.

3

" Specific" L.1 This comment number is not used for this station.

L.2 This comment number is not used for.this station.

1 L.3 This comment number is not used for this station.

q l

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., a v

_ = _.

Enc 6sure 2 to GNRO-04/CD56 I

Chapter 5 Page 69 DISCUSSION OF CHANGES CTS: 6.9 - REPORTING REQUIREMENTS ADMINISTRATIVE A.1 This comment number is not used for this station.

A.2 This reporting requirement is unnecessary since it generally included in the LER requirements to report fuel cladding failures that exceed expected values or that are caused by unexpected factors, i.e.,

being seriously degraded.

Since the criteria identified in 10 CFR 50.73 have been identified as the criteria in the area of degraded boundaries that necessitates reporting, any minor differences are negligible with regard to safety.

Therefore, the current reporting requirement is a duplication of the 10 CFR 50.73 reporting requirement and can be deleted.

A3 This note is deleted since the second unit at Grand Gulf has been cancelled.

Therefore, the note would sorve no useful purpose.

A.4 This change provides additional time to obtain calendar year based analyses results which are needed.for submittal of this report.

Since the report frequency is unchanged from annually, this change is considered administrative.

t A.5 This comment number is not used for this station.

A.6 This comment number is not used for this station.

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

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

A.9 The initial report requirements for the Annual Report and the annual Radiological Environmental Operating Report are being deleted.

These initial reports have been submitted, i

and deleting the discussion surrounding these initial reports are no longer necessary.

Therefore, the deletion is purely administrative in nature.

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

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

j A.12 The reporting of challenges to safety and relief valves is revised from an Annual report to a monthly report.

Since no change'in the details of the reporting are required, this is a change in'the timing only and is considered an administrative change.

l

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Encbsure 2 to GNRO-94/C0056 Chapter 5 Page 70 DISCUSSION OF CHANGES CTS: 6.9 - REPORTING REQUIREMENTS ADMINISTRATIVE (continued)

A.13 The applicable LCOs reference the COLR without these additional cross references.

Therefore, the references to the LCOs serve no functional purpose, and their removal is purely an administrative difference in presentation.

RELOCATED SPECIFICATIONS None in this section.

TECENICAL CHANGES - MORE RESTRICTIVE M.1 This change details the information to be included in the report.

These details are necessary to assure the reports are provided with similar content and format for comparison with other plants and with prior reports.

M.2 A new report is required in conjunction with the changes described in Section 3.4 for the reactor coolant system i

pressure and temperature limits.

In addition, requiremento J

are included for methods used to determine such limits and for submitting the report to the NRC.

TECHNICAL CHANGES - LESS RESTRICTl,YE

" Generic" LA.1 This comment number is not used for this station.

LA.2 Requirements relating.to the Process Control Program are relocated to the UFSAR and procedures.

Changes to the procedures and the UFSAR are controlled in accordance with 10 CFR 50.59.

j i

LA.3 Requirements for Special Reports and Startup Reports are relocated to the UFSAR and procedures.

Changes to the procedures and the UFSAR are controlled in accordance with 10 CFR 50.59.

" Specific" L.1 This comment number is not used for this. station.

L.2 This comment number is not used for this station.

...~n.

. to GNRO-94/00056 Chapter 5 Page 71 DISCUSSION OF CHANGES CTS: 6.10 - RECORD RETENTION ADMINISTRATIVE r.

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

A.2 This comment number is not used for this station.

A.3 This comment number is not used for this station.

A.4 This comment number is not used for this station.

A.5 This comment number is not used for this station.

A.6 This comment number is not used for this station.

RELOCATED SPECIFICATIONS None in this section.

TECHNICAL CHANGES - MORE RESTRICTIVE M.1 This comment number is not used for this station.

M.2 This comment number is not used for this station.

TECHNICAL CHANGES - LESS RESTRICTIVE

" GENERIC" LA.1 Record Retention requirements are relocated to the UFSAR and procedures.

Changes to the procedures and the UFSAR are controlled in accordance with 10 CFR 50.59.

" SPECIFIC" None in this section.

4

Encbsure 2 to GNRO.W00056 i

Chapter 6 Page 72 DISCUSSION OF CHANGES CTS: 6.10 - RECORD RETENTION ADMINISTRATIVE i

None in this section.

RELOCATED SPECIFICATIONS None in this section.

TECHNICAL CHANGES - MORE RESTRLCTIVE None in this section.

TECHNICAL CHANGES - LESS RESTRICTIVE

" GENERIC" LA.1 Radiation Protection Program requirements are relocated to the UFSAR and procedures.

Changes to the procedures and the UFSAR are controlled in accordance with 10 CFR 50.59.

" SPECIFIC" None in this section.

i l

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Encbsure 2 to CNRO 04X)0056 Chapter 6 Page 73 DISCUSSION OF CHANGES i

CTS: 6.12 - HIGH RADIATION AREA 1

ADMINISTRATIVE i

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

A.2 This comment number is not used for this station.

RELOCATED SPECIFICATIONS None in this section.

TECHNICAL CHANGES - MORE RESTRICTIVE None in this section.

TECHNICAL CHANGES - LESS RESTRICTIVE

" Generic" LA.1 This comment number is not used for this station.

LA.2 High Radiation Area control requirements are relocated to the UFSAR and procedures.

Changes to these controls can 1

only be made with pre approval of the NRC consistent with paragraph 20.203 of 10 CFR 20.

" Specific" L.1 This comment number la not used for this station, j

7 r.

-r y

.-p

.y

.. to CNRO-94/00056 Chapter 5 Page 74 DISCUSSION OF CHANGES CTS: 6.13 - PROCESS CONTROL PROGRAM ADMINISTRATIVE A.1 This comment number is not used for this station.

A.2 This comment number is not used for this station.

P A.3 This comment number is not used for this station.

A.4 This comment number is not used for this station.

RELOCATED SPECIFICATIONS None in this section.

TECHNICAL CHANGES - MORE RESTRICTIVE None in this section.

i a

TECHNICAL CHANGES - LESS RESTRICTIVE

" GENERIC" LA.1 Process Control Program requirements are relocated to the UFSAR and procedures.

Changes to the procedures and the UFSAR are controlled in accordance with 10 CFR 50.59.

" SPECIFIC" None in this section.

_ to GNRO-9N00056 Chapter 5 Page 75 DISCUSSION OF CHANGES CTS: 6.14 - OFFSITE DOSE CALCULATION MANUAL l

ADMINISTRATIVE A.1 The applicable Specifications provide the requirements without these additional cross references.

Therefore, the references to the Specifications serve no functional purpose, and their removal is purely an administrative difference in presentation.

A.2 Where possible, plant specific management position titles in the current Technical Specifications are replaced with generic titles as provided in ANSI /ANS 3.1.

Personnel who fulfill these positions are required to meet specific qualifications as detailed in proposed Specification 5.3, and compliance details relating to the plant specific management position titles are identified in licensee controlled doctiments (such as the UFSAR).

The two major-specific replacements are the generic " plant manager" for the manager level individual responsible for the overall safe operation of the plant and the generic descriptive use of "the corporate executive responsible for overall plant nuclear safety" in place of the Vice President position.

The plant specific titles fulfilling the duties of these generic positions will continue to be defined, established, documented and updated in a plant controlled document with specific regulatory review requirements for changes, such as the UFSAR or OQAM.

This approach is consistent with the intent of Generic Letter 88-06 which recommended, as a line item improvement, relocation of the corporate and unit organization charts to licensee controlled documents.

The intent of the Generic Letter, and of this proposed change, is to reduce the unnecessary burden on NRC and licensee resources being used to process changes due solely to-personnel titles changes during reorganizations..Since this change does not eliminate any of the qualifications, responsibilities or requirements for these personnel or the positions, the change is considered to be a change in presentation only and is therefore administrative.

A.3 This comment number is not used for this station.

A.4 This comment number is not used for this station.

A.5 This comment number is not used for this station.

RELOCATED SPECIFICATIONS None in this section.

l l

Encbsure 2 to GNRO-94/00J56 Chapter 5 Page 76 DISCUSSION OF CHANGES I

OFFSITE DOSE CALCULATION MANUAL CTS: 6.14 TECHNICAL CHANGES - MORE RESTRICTIVE i

None in this section.

TECHNICAL CHANGES - LESS RESTRICTIVE

" Generic" LA.1 Review and Audit requirements are relocated to the UFSAR and procedures.

Changes to the procedures and the UFSAR are controlled in accordance with 10 CFR 50.59.

LA.2 Requirements relating to the Radiological Environmental Monitoring Program and the Process Control Program are relocated to the UFSAR and procedures.

Changes to the procedures and the UFSAR are controlled in accordance with 10 CFR 50.59.

LA.3 Record Retention requirements are relocated to the UFSAR and procedures.

Changes to the procedures and the UFSAR are controlled in accordance with 10 CFR 50.59.

l

" Specific" None in this section.

1 i

l

=;;";r-F ATTACHMENT 1C CTS - PSTS COMPARISION DOCUMENT NO SIGNIFICANT HAZARDS CONSIDERATIONS i

Enc 6suf e 2 to CNRO.94/00056 Chapter 5 Page 78 NO SIGNIFICANT HAZARDS CONSIDERATIONS CTS: 6 - ADMISTRATIVE CONTROLS There are no

" SPECIFIC" No Significant Hazards Considerations evaluations required for the chapter.

'v:."??~~

ATTACHMENT 2 ITS - PSTS COMPARISION DOCUMENT 2A: MARKUP OF ITS 2B: DISCUSSION OF CHANGES

c"hae Pg 0 ATTACHMENT 2A ITS - PSTS COMPARISION DOCUMENT MARKUP OF ITS i

l l

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Encbsura 2 to GNRO-94/00056 Chapter $ Page 81 s

Responsibility 5.1 5 '. 0 ADMINISTRATIVE CONTROLS 5.1 Responsibility f

fnCMcapfS

@.5.1.1 Th'liMlant Ssperinun&nt}Ishall be responsible for overall unit operation and shall delegate in writing the succession to this responsibility during his absence.

Thklantd:YrYr$r.t]','or his desi nee, (;c::rth: i$h A 3rr- = __

.N :t---wa cre::=resa sha 1 approve, prior en implementation, each pr sed te experimenf and pr. ;;;D OC2 "E = :: " modi ficatio t

tems or equipment t1at affect nuclear safety.

5opdededsch h

5.1.2 ThhShiftAp;r.;5er(SS)[shallberesponsibleforthecontrol room command function.

-ma ement ective o this' ffect, CR3

.)

s n ny e Lniggst le 1 of rporat r sit naa Ouring any ' an',

j C,44 -

ha ic<

d annuMiy t 11 st ion per

.nel absenceoftheMSSFfromthecontrol,roonwhiletheunitisin MODE 1, 2, or 3, an individual with : : M Senior Reactor Y C-Y l

Operator (SRO) license shall be designated to assume the control-

i cM room command function.

During any absence of the95SP-from the control room while the unit is in MODE 4 or 5, an individual with an gin e velid SRO license or Reactor Operator. license shall be designated to assume the control room command function.

G % & G.,lf M5hn Mo, A CLT,/5 STS 5.0-1

" r..

O, 05/2S/92-

Encbsure 2 to GNRO-94C056 Chapter 5 Page 82 Organization 5.2 5.0 ADMINISTRATIVE CONTROLS 5.2 Organization 5.2.1 Onsite and Offsite Oroanizations Onsite and offsite organizations shall be established for unit operation and corporate management, respectively.

The onsite and offsite organizations shall include the positions for activities affecting safety of the nuclear power plant, a.

Lines of authority, responsibility, and comunication shall be defined and established throughout highest management levels, intermediate levels, and all operating organization positions.

These relationships shall be documented and updated, as appropriate, in organization charts, functional descriptiens of departmental responsibilities and relationships, and job descriptions for key personnel positions, or in equivalent forms of documentation.

These requirements shall be documented in thMFSARf;'

m a nc,p c D

Th7ft,lant Superin.:nd:ntfshall be responsible for overall b.

safe operation of the plant and shall have control over those onsite activities necessary for safe operation and maintenance of the plant; c.

specified corporate executive position shall have corporate responsibility for overall plant nuclear safety and shall take any measures needed to ensure acceptable performance of the staff in operating, maintaining, and providing tech isal support +c-+he_niant to ensur norlaar OP16 safety and

^ *- S P GLAMMA d h A ca L

d <>

'.f or #*-

r

, % ofsAR q

b_<.

3 d.

The individuals who train the operating staff,' carry out health physics, or perfom quality assurance functions may report to the appropriate onsite manager; however, these individuals shall have sufficient organizational freedom to ensure their independence from operating pressures.

7.

g b g goo'ing 5.2.2 Unit Staff Theunitstafforganizationshallk?:: 1011:= :

~

a.

Och en tty sh4ft 05:11 be-::gered cf :t 1:nt the si =

Qift gew cmaperition showa in T:ble 5.2.2-1.

2

-1 e

(continued) 1 I

BWR/6 STS 5.0-2 Rev.

O, 09/28/92 i

- to GNRO-04/00056 Chapter 5 Page 83 INSERT 2A A non-licensed operator shall be on site when fuel is in the reactor and an additional non-licensed operator shall be on site while the unit is in MODE 1, 2,

or 3.

i

.l INSERT 5.0-2

.. ~ -....

~-

~,. -

Encbsure 2 to CNRO 94/03056 -

Chapter 5 Page 84 Organization; 5.2 5.2 Organization 5.2.2 Unit Staff (continued) b.

At le'ast one licensed Reactor Operator (RO) shall be present Q4 in the control room when fuel is in the reactor.

In-l addition, while the unit is in MODE 1,.2, or 3,~at least one gg l licensed Senior Reactor Operator (SRO) shall be present in the control room.

3 d X.

ealthlhysicsfechnician shall be'on site when fuel is in he reactor.

The position may be vacant for not more than 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , in order to provide for unexpected absence, provided innediate action is taken to fill the required-position.

Ither aMicensed SR(or licensel RO limited t fuel d.

n h dling wh has no co urrent resp sibilities ing this ope tion sha be prese during fue handling an shall direc supery se all COR ALTERATIONS.

Administrative procedures shall be developed and implemented e.

to limit the working hours of unit staff who. perform safety related functions (e.g., licensed SR0s, licensed R0s, health physicists, auxiliary operators, and key maintenance

_J:L personnel).

h Adequate shift coverage shall be maintained with'out routine A

heavy use of. overtime.

The-objeetfr " ' " ^

M ara +4a; aere-1 ::rk n [" :r 12] S:;r dei, neada:1 N heer eek, h'le the mit is Op:r:tingt However, in the event that. unforeseen problems require substantial amounts of overtime to be used, or during extended periods of shutdown for refueling, major maintenance, or major plant modification, on a temporary basis the following guidelines shall be followed:

l 1.

An' individual should not'be permitted to work more than 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months straight, excluding shift turnover time;'

2.

An individual should not'be permitted to work more than.

16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months in any' 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months period, nor more than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months in any 48 hour5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months period, nor more than 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months in any 7 day period, all ' excluding shift turnover time; (continued)

BWR/6 STS 5.0-3 Rev.

O,J09/28/92 to GNRO 94/00050 Chapter 5 Pags 85 INSERT 3A Shif t crew composition may be one less than the minimum requirement of 10CFR50.54 (m) (2) (1) and 5.2.2.a for a period of time not to exceed 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months in order to accommodate unexpected absence of on-duty shif t crew members provided immediate action is taken to restore the shift crew composition to within the minimum requirements.

INSERT 5.0-3

Enc 6sure 2 to GNRO-94100056

- Chapter 5 Page 80 4

Organization 5.2 5.2 Organization 5.2.2 Unit Staff (continued) 3.

A break of at least 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months should be allowed between work periods, including shift turnover time; 4.

Except during extended shutdown periods, the use of overtime should be considered on an individual basis and not for the entire staff on.a shift.

4

, mo n %e e Any deviation from the above guidelines shall be authorized O6I y

""~sby thQlant S perint;nd;nt?or his designee, in a

accordance with approved administrative procedures, or by higher levels of management, in accordance with established procedures and with documentation of the basis for granting the deviation.

Controls shall be included in the procedures such that individual overtime shall be reviewed monthly by thi%(lant l

menese P Superintendent}Por his designee to ensure that excessive hours have not been assigned.

Routine deviation from the above guidelines is not authorized.

M

~

OW ihe m at of overtime worked by unit staff

)

performing safety reiaud functi--- -

" be its ted and controlled in acco e NRC Po ic nt on worki eneric Letter 82-12).

A Y

h f.

Th peratio or A;;ist=t Operations "ri :r' O shal h

n SR cense.

g g.

The Sh echnical Advisor (STA) shall provide advisory -

g technical support to the Shift kpervitr-(SS) in the areas of thermal hydraulics, reactor engineering, and plant analysiswithregardtothesafeoperationoftheunit.1 r

a Tnse.c4 4A (C-7 G BWR/6 STS 5.0-4 Rev.

O, 09/28/92

.-. _ _. _ _ _ _ - _ _ - _ _ - _ _ - - - _ _ _ - _ _ - - - _ _ _ _ _ to GNRO-94/00056 Chapter 5 Page 87 j

i j

INSERT 4A In addition, the STA shall meet the qualifications specified by the Commission Policy Statement on Engineering Expertise on Shift.

l 4

I l

l l

i I

j INSERT l

5.0-4 t

I i

.1

.- l I

i

Encbsure 2 to CNRO-94/00056 Chapter 5 Page 88 Organizat Table 5.2.2-1 (page 1 of 1) e Minimum Shift Crew Composition (a)

[ Single Unit Facility]

P0 N(D)

MINIMUM CREW NUMBER

~

d UNIT IN MODE 1, 2, OR 3 UNIT IN/ DE 4 OR 5 SS 1

1

\\

SRO 1

None R0 2

1 A0 (C) 2 1

STA 1

None N

(a)

The shift crew composition may b one ss than the minimum requirements of Table 5.2.2-1 for not more t n 2 ho to accomodate unexpected absences of on-duty shift cre members pro ded immediate action is taken to restore the shift ew composition within the minimum requirements of Table 5.2

-1.

This provision does not permit any shift crew position to be unma ed upon shift change e to an oncoming shift creunan being late or sent.

(b)

Table Notation:

[ Shift pervisor] with a Senior Reactor Operator icense; SS SRO - Indiv ual with a Senior Reactor Operator license; IndiidualwithaReactorOperatorlicense; RO A0

- Aufiliary Operator; STA - Shift Technical Advisor.

(c)

The TA position may be filled by an on-shift SS or SRO provided t in vidual meets the Commission Policy Statement on Engineering pertise on Shift.

BWR/6 STS 5.0-5 Rev.

O, 09/28/92

Encbsura 2 to GNRO-94/CCoS6 Chapter 5 Page 89 CM Organization 5.2 Table 5.2.21 (page 1 of 1)

Minimum Shif t Crew Composition (a)

(Two Units With a Conmon Control Room)

(Totals for Both Units)

/

POSITION (b MINIMUM CREW NUMBER

\\

ONE UNIT IN MODE

/

1,2,OR3,ANDONF/

CH UNIT IN MODE UNIT IN MODE 4 0 EACH UNIT IN MODE 1, 2, OR 3 OR DEFUELED 4 OR 5 CR DEFUELED i

1 i

SRO 1

1 None RO 3

3 2

I 3

3 3

)

AO (C)

STA 1

1 None N

J

-l (a)

The shift crew composition may be one ess an the minimum requirements of Table 5.2.2 1 for not more than 2 ho s to ac ate unexpected absence of on-duty shif t crew members, provided immed to action 1 taken to restore the shift crew composition to within the minimum equirements o Table 5.2.2-1.

This provision does not permit any shift crew posit n to be unmanned shift change due to an oncoming shift crewman being late or a ent.

(b)

Table Notation:

$$ -(ShiftSupervisor) ith a Senior Reactor Operator 1 ense for each unit whose reactor contains uel.

SR0 - Individual with Senior Reactor Operator license for ch unit whose reactor contains fuel. Otherwise, provide an individual for eac unit who holds a Senior React Operator license for the unit assigned. O ing CORE ALTERATIONS-on either u t at least one licensed SR0 or licensed SR01 ited to fuel-

handling, has no other concurrent responsibilities, must e present.

1 R0 - Individu with a Reactor Operator license or a Senior Reacto Operator license 1

for uni assigned. At least one R0 shall be assigned to each u it whose reactor.

j contai s fuel and one R0 shall be assigned as relief operator fo unit (s) in MODE, 2, or 3.

Individuals acting as relief operators shall ho a license j

for oth units. Otherwise, for each unit, provide a relief operat who holds a 11 nse for the unit assigned.

AO - A least one auxiliary operator shall be assigned to each unit whose eactor tains fuel.

STA - ahift Technical Advisor.

(c)

The STA position may be filled by an on-shif t $$ or SR0 provided th's individual meets th Comunission Policy Statemerit on Engineering Expertise on Shif t.

BWR/6 STS 5.0-6 Rev.

O, 09/28/92

,,..,,., ~,,

,.--n.

s.

Enc 6sure 2 to GMRO-94/0005G Chapler 5 Page 20 C

Organizati n Table 5.2.2-1 (page 1 of 1)

/'

Minimum Shift Crew Composition (a)

[Two Units With Two Control Rooms]

/

(Numbers for Each Unit)

/

POSITIO MINIMUM CREW NUMBER h

UNIT I ODE UNIT IN MODE 1, 2 OR 3; UNIT IN MODE

, 2, OR 3; UNIT IN MODE OT R UNIT 4 OR 5; OTHER OT R UNIT IN 4 OR 5; OTHER IN MODE UNIT IN MODE DE UNIT IN MODE 4 OR 5 OR 4 OR 5 OR 1, 2, OR 3 1, 2, OR 3 DEFUELED DEFUELED j

SS 1(d) 1 1(d) 1(d)

SRO 1

e 1

None R0 2

1 2

1 A0 (c) 2(d) 1 2

2(8)

STA 1

None 1

None

/

\\

(a)

The shift crew compo tion may be one ss than the minimum requirements of Table 5.2.2-1 f not more than 2 hou in order to accomodate unexpected absenc of on-duty shift crew mbers provided imediate action is taken o restore the shift crew c position to within the minimum requir nts of Table 5.2.2-1.

This rovision does not permit any shift cre position to be unmanned upon sh ft change due to an oncoming shi t crewman being late or absent.

(b)

Table Nota ion:

SS

-[

ift Supervisor] with a Senior Reactor Operat r license; SRO -

dividual with a Senior Reactor Operator licens -

R0 ndividual with a Reactor Operator license; AO

- Auxiliary Operator; STA Shift Technical Advisor.

(c)

The STA position may be filled by an on-shift SS or SRO provi ed the in ividual meets the Comission Policy Statement on Engineerin E pertise on Shift.

(d) ndividual may fill the same position on the other unit if licens both.

for (e

One of the two required individuals may fill the same position on the other unit.

/

BWR/6 STS 5.0-7 Rev.

O, 09/28/92 to GNRO-94/00056 Chapter 5 Page O!

Unit Staff Qualifications 5.3 5.0 ADMINISTRATIVE CONTROLS 5.3 Unit Staff Qualifications

_- g-Dr'rNeteMQualifications for members of di s i.a f f M ~

(

be specified by use of an overaHg i fi ement referencing an ANSI 6)

Standard acceptable to the NR r Dps 'f ing individual position qualifications.

, the first method is pre de;however,the second t

is adaptable to those unit staffs requiring 7s en

}

tion statements because of unique organizational structure.

/

hsc-t T A

-J 5.3.1 Each member of t gunit staff shall meet or exceed the minimum qualifications o tRegulatory Guide 1.8, Revision 2, 1987, e. u.or+s2 eeent-r:visicas, or f951 Stenderd e :

h --

+The :t:f' ot ce cred by [Reguiatory Gu;J; 1.0] :hallEpteble te the-NAC-sta meet er

-exc::d the minimum qualifications of [Reguletions,- Regulatory _

l -Guide;

r ^95! Stenderd5 eccepteble to NRC-staff],,f dditionh (the hift Tech ~ al Advisor shall meet t qualifications gg 1peci ' d by th ission Poli h Stateme on Engineeri (Exhqtis on Shift.

BWR/6 STS 5.0-8 Rev.

O, 09/28/92

Encbsure 2 to GNRO-94/C3056 -

Chapter $ Page 92

~

INSERT 8A ANSI N18.1-1971 for comparable positions, except for the Superintendent, Radiation Control and the STA, who shall' meet or i

exceed the education and experience requirements of ANSI /ANS 3.1-1981 as endorsed by

-1 1

l

)

INSERT 5.0-8

=,.

.. -. ~

. ~.. -.

Encbsure 2 to CNRO-944X:356 Chapter 5 Page 93 W

+ 4-5.0 ADMINISTRATIVE CONTROLS 4 Trainin D 6[

^

's x

j\\

5.4.1 A retrainin replacement training r or the unit staff k

shall be maintaine r the d n of the [ position title]

and shall meet or e uirements and recommendations of Section[

an ANSI Standar eptable to the NRC staff] and 10 CF

, and, for appropriate design positions, shall ude familiarization with relevant indus erational experience.

\\

i 1

BWR/6 STS 5.0-9 Rev.

O, 09/28/92

Enebsure 2 to GNRO 94Q056 Chapter 5 Page 04 RUVIcW) and Adht$

.5 5.0 ADMINISTRATIVE CONTROLS 5.5 hvi = and-Audits o o eviewer's Note:

The licensee shall describe the method (s) established co duct independent reviews and audits.

The methods may take a range forms acce able to the NRC.

These methods may include creating an organ' ational unit o standing or ad hoc committee, or assigning individuals pable of conductin these reviews and audits.

When an individual perfo a review function, a oss disciplinary review detemination is necess y.

If deemed necessary, suc eviews shall be perfonned by the review per onnel of the appropriate disci ine.

Individual reviewers shall not r iew their own work.

Regardless of the od used, the licensee shall specif the functions, organizational arrange nt, responsibilities, appropri te ANSI /ANS 3.1-1981 qualifications, and repo ng requirements of each f ctional element or unit that contributes to these p cesses.

Reviews and audits of activities ffecting pla safety have two distinct elements.

The first element is th reviews rformed by plant staff personnel to ensure that day to day activities ec ducted in a safe manner.

These reviews are described in Section 5.5.1.

e second element, described in Section 5.5.2, is the [offsite] reviews audits of unit activities and programs affecting nuclear safety tha are rformed independent of the plant staff.

The [offsite] reviews and a its shou provide integration of the reviews and audits into a cohesiv program that rovides senior level utility management with an assessment o facility operati and reconmends actions to improve nuclear safety and p1 t reliability.

It s uld include an assessment

_of the effectiveness of revi s conducted according t Section 5.5.1.

5.5.1 Plant Revie Reviewe /s Note:

The licensee shall describe prov ions for plant revie (organization, reporting, records) and the a ropriate

_ ANSI NS Standard for personnel qualification.

5.5.1.1 nctions The [ plant review method specified in Specification 5.5.1]

hall, as a minimum, incorporate functions that:

a.

Advise the [ Plant Superintendent] on all matters related nuclear safety; (continued)

BWR/6 STS 5.0-10 Rev.

O, 09/28/92 to GNRO-94/00050 Chapter 5 Pagn 95 Rui4xs--ae-Attwa-

- - b 5-Q Reviews and Audits p

5.5.1.1 Functions (continued)

Recomend to the [ Plant Superintendent] appr al or disapproval of items considered under Spe ' ications

.5.1.2.a through 5.5.1.2.e prior to th 'r implementation, e cept as provided in Specification 5.

1.3; c.

Dete ine whether each item consid ed under Specifications 5.5.1.

through 5.5.1.2.d cons 'tutes an unreviewed safety question s defined in 10 CFR 5.59; and CN d.

Notify the ce President-uclear Operations] of any safety signifi nt disagre nt between the [ review organization or ndividu specified in Specification 5.5.1]

and the [ Plant Su rint dent] within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

However, the [ Plant Superint nt] shall have responsibility for resolution of such i greements pursuant to Specification 5.1..

5.5.1.2 Responsibilities The [ plant revie method specifie in Specification 5.5.1] shall be used to cond ct, as a minimum, r iews of the following:

All pro osed procedures required b Specification 5.7.1.1 a.

and c nges thereto; b.

All reposed programs required by Spec 1 ication 5.7.2 and ch ges thereto; I proposed changes and modifications to u "t systems or c.

l quipment that affect nuclear safety; d.

All proposed tests and experiments that affect clear safety; and All proposed changes to these Technical Specificati ns (TS),

e.

their Bases, and the Operating License.

4 (continued)

BWR/6 STS 5.0-11 Rev.

O, 09/28/92

Encbsur3 21o GNRO4m50 Chapter 5 Pags 96 Rev4ewsc-end-Atrditt W

kReviewsandAudits (continued) 5.5.2 f0ffsitel Review and Audit eviewer's Note:

The licensee shall describe iews and audits independent of the plant' provisions for r

staff (organization, ing, and records) and the appropria ANSI /ANS Standards for rep perso el qualifications.

These indivi or offs (e provided organizational in als may be located onsite pendence from plant staff is maintal ed.

The [ technical] rev' w responsibilities, Specificatt 5.5.2.4, shall inci onsite.

e several individuals located 5.5.2.1 Functions The [offsite review Specification 5.5.2] nd a dit provisions specified in following functions tht 1, as a minimum, incorporate the a.

Advise the [Vi Pres ent-Nuclear Operations] on all matters relat to nucl r safety; b.

Advise the nagement of t audited organization, and [its Corporate anagement and Vic President-Nuclear Operations),

of the a it results as they r late to nuclear safety; c.

Reco d to the management of th audited organization, and its nagement, any corrective act safe y and plant operation; and n to improve nuclear d.

N ify the [Vice President-Nuclear Ope tions gnificant disagreement between the [rev*ew o]rganization oro ndividual specified in Specification 5.5.

and the

[ organization or function being reviewed] wi hin 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

5.5.2.2

[ fsite] Review Responsibilities he [ review method specified in Specification 5.5.2] s all be responsible for the review of:

The safety evaluations for changes to procedures, eq 'pment, a.

or systems, and tests or experiments completed under t e provisions of 10 CFR 50.59, to verify that such actions do not constitute an unreviewed safety question as defined 'n 10 CFR 50.59; (continued)

BWR/6 STS 5.0-12 Rev.

O, 09/28/92

I Enciosurs 2 to GNRO-94/00056 i

Chaptre 5 Paga 97

'j Mi= =d Acis C.C h{ Reviews 'and Audits 7

5.5.2.

[0ffsite] Review Responsibilities (continued) b.

Proposed changes to procedures, equipment, or ystems that involve an unreviewed safety question as de 'ned_in 10 CFR 50.59;

c.. Proposed tests or experiments that in Ive an unreviewed afety question as defined in 10_CF 50.59; d.

Pr osed changes to TS and the 0 rating License; e.

Viola ons of codes, regulat ns, orders, license requir nts, and internal rocedures or instructions having nuclear fety significan e; gg f.

All License Event Rep rts required by 10 CFR 50.73;-

g.

Plant staff pe o nce; h.

Indications of nticipated deficiencies in any aspect of design or oper io of structures, systems, or components that could af ect n lear safety; i.

Significan accidental unplanned, or uncontrolled

'l radioact e releases, i ciuding corrective action to prevent recurre ce; j.

Sign *ficant operating abno lities or deviations from no 1 and expected performa e of equipment that affect n lear safety; and k.

The performance of the correctiv action system.

R orts or records of these reviews shal be forwarded to the ice President-Nuclear Operations] with 30 days following completion of the review.

5.5.2.3 Audit Responsibilities The audit responsibilities shall encompass:

a.

The conformance of unit operation to provision contained j

within the TS and applicable license conditions, i

b.

The training and qualifications of the unit staff; (coninued)

BWR/6 STS 5.0-13 Rev.

O, 09/28/

e Encbsura 2 to GNRO-94/00056 Chapttr5 Pzga 98 Re.in3 erc A xits

-+5

% Reviews and Audits

/

5.5.2.

Audit Responsibilities (continued) j i

c.

The implementation of all programs required by Specification 5.7.2; d.

ctions taken to correct deficiencies oc rring in e

ipment, structures, systems, compone s, or method of ope tion that affect nuclear safety; nd Og^

e.

Other a ivities and documents as equested by the [Vice President Nuclear Operations].

Reports or record of these audits hall be forwarded to the [Vice President-Nuclear erations] wi in 30 days following completion of the review.

5.5.2.4

[ Technical] Review Resp sib' ities The [ technical] review re nsibilities shall encompass:

Plant operating ch racte istics, NRC issuances, industry a.

advisories, Lice ee Even Reports, and other sources that I

may indicate ar s for imp ving plant safety; b.

Plant operati ns, modificatio s maintenance, and surveillanc to verify indepen ently that these activities are perfo d safely and correc ly and that human errors are reduced a much as practical; c.

Intern and external operational e erience information that y indicate areas for improvin plant safety; and d.

Mak.g detailed recommendations throug the [Vice Pr sident-Nuclear Operations] for revis ng procedures, uipment modifications, or other means o improving nuclear afety and plant reliability.

5.5.3 R cords Written records of reviews and audits shall be maint ined.

As a minimum these records shall include:

a.

Results of the activities conducted under the prov ions of Section 5.5; (con 'nued)

R/6 STS 5.0-14 Rev.

O, 09/28/92

.. _ -. -.~ -

Enclosura 2 to GNRO 94/00056.

' i Chapter 5 Pags 99 5CviCW3 ahd Augyt$

+-

ws and Audits 5.5.3 Records tinued)

OC2$

b.

Recommendation the management o e organization.being audited; c.

An assessment of the sa y

' nificance of the review or audit findings; 1

d.

Recomended a oval or disapproval o tems considered under Spe ications 5.5.1.2.a through 5.

.2.e; and e.

De ination whether each item considered unde ecifications 5.5.1.2.a through 5.5.1.2.d consti es an unreviewed safety question as defined in 10 CFR 50.

BWR/6.STS 5.0-15 Rev.

O, 09/28/92

.R i

w

~'

~

- to CNRO 94/00050 Chapter 5 Page 100

-( ((by e -to 6,6, /h IS Coasa Cwnb 04 da55 h Ecoef-sor J (

3]

5.0 ADMINIS CONTROLS j

5.5 hical Specifications (TS) Bases _ Control feoph

~

Tnd i, =r, t o W IloA M Q @ Changes to the Bases of the TS shall be made under appropriate administrative _c_ontrols qrc: erec :cerdi".&A C2) h ecificati:r 5. 3 i

m-s

'3M b Licensees may make changes to Bases without prior NRC approval provided the changes do not involve either of the following:

l I'

h.

A change in the TS inc orated in the license; or

'L h.

A change to the (ifEIEibufSAR or Bases that involves an

+

unreviewed safety question as defined in 10 CFR 50.59.

Cp S Q C-The Bases Control Program shall contain provisionsA ensure that the Bases are maintained consistent with theuFSAR.E9 h~

- $ h Proposed changes that meet the criteria of 6) er LNbove shall be reviewed and approved by the NRC prior to implementation.

Changes to the Bases implemented without prior NRC approval.shall be provided to the NRC on a frequency consistent with 10 CFR 50.71.

t

@k BWR/6 STS 5.0-16 Rev.

O, 09/28/92

Enc 6sure 2 to GNRO-%C350 Chaptor 6 Page tot INSERT 16A This' program provides a means for processing changes to the Bases of these Technical Specifications.

\\

1 INSERT 5.0-16

Encbsure 2 to CNRO 94/CD0$0 Chapter 5 Page 102 Procedures f?.earem;, and urMA 53 4 4

5.0 ADMINISTRATIVE CONTROLS 5.T'/ Proceduresgrey ami, end ".:nua4P bl

?,+e#tD' b L

5.X.1@7- @

l 4

\\ ritten procedures shall be established, implemented, and W

maintained covering the following activities:

The applicable procedures recommended in Regulatory a.

Guide 1.33, Revision 2, Appendix A, February 1978; b.

The emergency operating procedures required to implement the requirements of NUREG-0737 and UREG-0737, Supplement Q PW

' ' 9 2: :tatcd in [Gneric better-42-33}+ '

c-Mrurfty plan implementationT]

yd cy M mentatign;

) Quality assurance for effluent and environmental monitoring;

/ c't Q Fire Protection Program implementation; and, g

.} All programs specified in Specification 5..M 5 4 1.2 hviewandApproval

-M E(ch procedure of Specification 5.7.1.1, and chan get ereto, i

shaTK by the{e reviewed in accordance with Specifica n 5.5.1, approved

[Mant Superintendent] or his desipe in accordance with approved adiihustrative procedures prWto implementation and reviewed perioditaQy as set fort T administrative procedures, y

s 5.7.1.3 Temporary Changes N

,N Temporary changes procedures'of Specification 5.7.1 may be made provided:

\\ N

)

a.

T intent of the existing procedure t1ylot altered;

/

y x

/

\\

The change is approved by two members of thi plant

,/

\\<

management staff, at least one of whom holds a'Sqnior Reactor Operator license on the unit affected; an(#

Y c

cong ued)

Y BWR/6 STS

.0-17 Rev.

O, 09/28/92 to GNRO-94/00056 Chapter S Page_1 7

___._. / ' Procedures. Programe an& #anea1s,

/

~~5--?

LI Erncedures; Progragn_.and Manua4+

5.7.1.3 Temg Q anges (continued) x x

y OC c.

The change is document & and_rev.

- n accordance with I

d Specification 5.5.

oved W [ Plant Superin or his designee in accord 3nt with approved ad ' istrative oracedures within 14 days of imp

. ntation.

5 Ap m v o.0 AM Q j SIM)

Proorams and Manuals 95 The following programs shall be established, implemented, and maintained.

(S'.72.t-

-RadQtion Protection Prog?

O Procedures for peDn' net-radiatjon prottr. tion-4%^repared consistent with the re ui atPtif TD CFR 20 and shall be

approved,

'nt

, and adhered to for'aMwp ations involving radiation exposure.

I..

Process ontrol Program (PCP) (

5 j

The shall contain the current formulas, sampling,

yses, i

packaging %nninations to be made to ensure tha rocessing and '

tests, an of soM(radioactive wastes will be emplished to

(

l

\\

O',.y regulations; burial grouiidqquirements-nd other requirements

{

ensure compliance with40 CFR 20,10 CFR 61 nd 10 CFR 71; state governing the disposal of solld4adi ctive waste.

{

Licensee initiated changes to e PC.

a.

Shall be documente nd records of rev s performed shall be retained.

T documentation shall con in:

1.

suffi ent information to support the chan (s) and ap opriate analyses or evaluations justifyi the ange(s),and l

2 a detennination that the change (s) maintain the ov all i

conformance of the solidified waste product to the existing requirements of Federal, State, or other applicable regulations.

j (continued)

BWR/6 STS 5.0-18 Rev.

O, 09/28/92 to GNRO-94/00056 Chapter 5 Page 104 Programs, and Manuals u

55 g

5.5 g

5.7 krocedurcs ProgramgandManuals y

9.2.2

~ Process trol Program (PCP)

(conti 7

b.

Shall b 1ve aTQ r review and acceptance by the

\\

[r w method of Specification and the approval of

,A e [ Plant Superintendent].

5. L 2+

Offsite Dose Calculation Manual (00CM) r 5,1

~

a.

The 00CM shall contain the methodology and parameters used in the calculation of offsite doses resulting from radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring alarm and trip setpoints, and in the conduct of the ' Radiological l

(nvironmentalynitoring1rogram; and a

b.

TheODCMshallalsocontaintheladioactive4ffluent

%ontrols and'4adiological TJivironmental Konitoring programs (v{3

- Qequircrbr!betificme 5M and descriptions of the S

information that should be included in the Annual

__q Radiological Environmental Operatings and'Radioa tive Effluent Releasef reportsfr p;J.1.y pea nset+en.

rca

-- Q,.31-and-Spe<4++ eat +er, J.

4f Licensee initiated changes to the ODCM:

a.

Shall be documented and records of reviews performed shall be retained.

This documentation shall contain:

1.

sufficient infomation to support the change (s) and

,, I Macecther with thB appropriate analyses or evalua ns justifying the change (s), and 2.

a determination that the change (s) maintain the levels of radioactive effluent control required by

Jo 7 '~

10 CFR 2 40 CFR 190, 10 CFR 50.36a, and 10 CFR 50, ppendix I, and not adversely impact the v

accuracy or reliability of effluent, dose, or setpoint calculations; b.

Shall become effective after review and acceptanc N h P C9) J[rcview mthed vi Spnif kai.ivn 5.5.1] cnd the appfoval of the$lantSuperintendentf;'and

^ C\\ A,O S a ge -

(continued)

BWR/6 STS 5.0-19 Rev.

O, 09/28/92

Encbsurs 2 to GNRO-9400056 Chapter 5 Pegs 105 Cl (Pr;;;dir Q Programs and Manuals i

Mus tec;d Programy and Manuals

--5.7.2.3 Offsite Dose Calculation Manual (00CM)

(continued) f: 6. J c.

Shall be submitted to the NRC in the fonn of a complete, legible copy of the entire ODCM as a part of, or concurrent with, the Radioactive Effluent Release Report for the period of the report in which any change in the ODCM was made.

Each change shall be identified by markings in the margin of the affected pages, clearly indicating the area'of the page that was changed, and shall indicate the date (i.e., month and year) the change was implemented.

5.7.2.+

Primary Coolant Sources Outside Containment iT. S'.1 This program provides controls to minimize leakage from those portions of systems outside containment that could contain highly radioactive fluids during a serious transient or accident to levels as low as practicable.

The systems includWthenew a-h yt Pru suic Cer: Sprey. utg3 ore::27: cg73 3p,,,,

g,3;;,3; ; ;agg,,,

6I R:.=:1, o.neter c;rc nul.i. ion coo;;c.;;, hydrer ~c~.Mur,9 2g J pivuess sempiing, enu 3iontui Ce; - t :c.th gThe program shall include the following:

a.

Preventive maintenance and periodic visual inspection requirements; and b.

Integrated leak test requirements for each system at refueling cycle intervals or less.

5.7.2.5._

In Plant Radiation Monitoring

~

x.

This' prog qm provides controls to ensure the capab

' y to i

accurately determine the airborne iodine co ration in vital OC_.3.3 ' ' ~ ~

areas under accilfent. onditions.

This ram shall include the following:

a.

Training of person b.

Procedures r monitoring; and c.

Pr sions for maintenance of sampling and a sis quipment.

k (continued)

BWR/6 STS 5.0-20 Rev.

O, 09/28/92

Enc 60ure 2 to CNRO-94/C0056 Chapter 5 Page 103 i

INSERT 20A a.

LPCS System; b.

HPCS System; c.-

RHR System outside containment containing steam or water, except the line to the liquid radwaste system and headers that

'f are isolated by manual valves; d.

RCIC System outside containment containing steam or water, except the drain line to the main condenser; e.

Hydrogen analyzers of the Combustible Gas Control System; f.

Post Accident Sampling System;-

g.

Feedwater Leakage Control System; and h.

Suppression pool level detection portion of the Suppression Pool Makeup System.

i 6

6 i

INSERT 5.0-20 l

r

_, _ _.l to CMRO-94/00356 Chapter 5 Page 107 DcedureR) Programs, and Manuals

{

4r;--

ES 5.I h ;cet re Q Programp and Manuals Q.7.2 Frogram> and Hanuai>

(;;ntiued 4.7.2.5-Post Accident Sampling

5. 5. 3 This program provides controls that ensure the capability to obtain and analyze reactor coolant, radioactive gases, and particulates in plant gaseous effluents and containment atmosphere samples under accident conditions.

The program shall include the following: a. Training of personnel; b. Procedures for sampling and analysis; and c. Provisions for maintenance of sampling and analysis equipment.

5. 7.2. 7 -

Radioactive Effluent Controls Program Es.H This program confoms to 10 CFR 50.36a for the control of radioactive effluents and for maintaining the doses to members of the public from radioactive effluents as low as reasonably achievable. The program shall be contained in the 00CM, shall be R implemented by procedures, and shall include remedial actions to be taken whenever the program limits are exceeded. The program shall include the following elements: a. Limitations on the functional capability of radioactive liquid and gaseous monitoring instrumentation including surveillance tests and setpoint determination in accordance OPd with the methodology in the ODCM; b. Limitations on the concentrations of radioactive material g* released in liquid effluents to unrestricted areas, h c m.' # \\ conforming to4 10 CFR 20, Appendix B, Table olumn 2; n rece .g Monitoring, sampling, and analysis of rad oactive liquid and @lAs in c. gaseous effluents in accordance with 10 CFR 20. nd with the methodology and parameters in the ODCM; g d. Limitations on the annual and quarterly doses or dose commitment to a member of the public from radioactive materials in liquid effluents released from nit to unrestricted areas, confoming to 10 CFR 50, Appendix 1; ig (continued) BWR/6 STS 5.0-21 Rev. O, 09/28/92

Encicsurs 2 to GNRO-94/00056 Chaptu 5 Pzgs 108 N Precedur M Programs and Manuals y 7" -t.+ E5 ( ecedu Y Prograag and Manuals 5.7.2.7-Radioactive Effluent Controls Program (continued) 5%, y e. Determination of cumulative and projected dose contributions from radioactive effluents for the current calendar quarter and current calendar year in accordance with the methodology and parameters in the 00CM at least every 31 days; f. Limitations on the functional capability and use of the liquid and gaseous effluent treatment systems to ensure that appropriate portions of these systems are used to reduce releases of radioactivity when the projected doses in a period of 31 days would exceed 2% of the guidelines for the annual dose or dose consnitment, conforming to 10 CFR 50, Appendix 1; g. Limitations on the dose rate resulting from radioactive material relglAenfmToinii te the d;;; a::::hted witQ ed in gaseous effluents to areas beyond the site boundarjt p 10 CFR 20, ".;pendh B, T:ble II, Cch= '1R Tose d ? l Limitations on the annual and quarterly air doses res Q2A h. from noble gases released in gaseous effluents from unit to areas beyond the site boundary, conforming to 9" I 10 CFR 50, Appendix I; 1 i. Limitations on the annual and quarterly doses to a member of the public from iodine-131, iodine-133, tritium, and all radionuclides in particulate fann with m:lf lives >-8 daysj in gaseous effluents released from(;@hunit to areas beyond-the site boundary, conforming to 10 CFR 50, Appendix I; and J. Limitations on the annual dose or dose commitment to any { member of the public due to releases of radioactivity and to radiation from uranium fuel cycle sources, conforming to 40 CFR 190. ( ~ M ic al Environmental Monitoring Program N This program is for mon r4rtg'the r ion and radionuclides in t cp _ _ the environs of the plant. T ogram shall provide-i representative measure of radioactivitydn the highest potential ex o athways and verification of~the accuracy of the ef monitoring program and modeling of envh ental j L ure pathways. The program shall be contained in the CM, f k (continued) BWR/6 STS 5.0-22 Rev. O, 09/28/92

i to CNRO-94/00066 I Chapter S Pags 109 INSERT 22A 1. For noble gases: Less than or equal to 500 mrem /yr to the total body and less than or equal to 3000 mrem /yr to the skin, and 2. For iodine-131, for iodine-133, for tritium, and for all radionuclides in particulate form with half-lives greater than eight days: less than or equal to 1500 mrem /yr to any organ; INSERT 5.0-22

hcedure) Programp and Manuals 3. i-j T.S. N Encbsure 2 to CNRO-94M0056 ++ uCtr:Q-Programs and Manuals Ch*P S Page 110 4

5. h 2. 8 Radiological Environmental Monitoring Program (conti dl' x

skaR conform to the guidance of 10 CFR 50, A dix I, and shall include t following: a. Monitoring, ling, analysi and reporting of radiation and radionuclides the methodology and par (a ironment in accordance with the s in the 00CH; b. A Land Use Cen to ensu changes in the use of areas at and beyo the site boundary ar dentified and that-modifica ' ns to the monitoring progr are made if required by th esults of this census; and c. rticipation in an Interlaboratory Comparison ogram to ensure that independent checks on the precision an accuracy of the measurements of radioactive materials in environmental sample matrices are performed as part of he

5. s'. 5 quality assurance program for environmental monitoring. j

5. '.2. F Component Cyclic or Transient Limit 8I J

This program provides controls to track thekintien cyclic and transient occurrences to ensure tha maintained within the design limits. {%% g,gy g PT 5.7.2.10 Pre-Stressed Concrete Containment Tendon Surveillanc rtFgram s This'progra rovides controls for mon ng any tendon a degradation n pre-stressed con ett containments, including effectiveness of its cor protection. medium, to ensure containment struct ntegri The program shall include baseline nts prior to initia gerations' The Tendon Sur ance Program, inspection ~frequencie, nd acceptance iteria shall be in accordance with [ Regulatory de 1.35, Revision 3,1989]. The provisions of SR 3.0.2 and SR 3.0.3 are applicable to'the f Tendon Surveillance Program inspection frequencies. x g -5Wn2Qce. Inspection Program This program provides con rvice i 1on of ASME Code Class 1, 2, and 3 components, i able supports. The program sh'all includs--th following: 7 w (continuea) BWR/6 STS 5.0-23 Rev. O, 09/28/92 to GNRO 94/00356 Chapter 5 Page 111 f f " Q ::de-q Programs, and Manuals 7 t.4-s'. 5 i' CErecedu--Q Programs and Manuals g 5}.M1 Inservice Inspection Program (continued) sions that inservice inspection of A Code Class 1, 2, an components shall be perfonne n accordance with ASME Boile Q Pressure Vessel C and Addenda, Section XI, as r W red by 10 50.55a; b. The provisions of 5 .2 applicable to the frequencies for performing ' rvice inspec ' activities; OC34~~ c. An inserv3'c inspection program for pipin 'dentified in NRC GeneriMetter 88-01 in accordance with the staff po 'tfons on schedule, methods, personnel, and sa e ansion included in Generic Letter 88-01, or in acc ance ith alternate measures approved by the NRC staff; and / 'd. 7 Nothing in the ASME Boiler and Pressure Vessel Code shall be construed to supersede the requirements of any TS. 5.7.2712 Inservice Testing Program This pro a provides controls for inservice testing of ASME Code Class 1, 2, d 3 components including applicable supports. The C34 p[ogram shall i lude the following: a. Provisions that service testing of AS Code Class 1, S d 3 pumps, valves, and snubbers shall be erformed in ac rdance with Sectio I of the ASME Boile nd Pressure / Vess 1 Code and applicab ddenda as required 10 CF 50.55a; b. Testing f quencies specified in ion XI of the ASM Boiler and ressure Vessel Code and applicable Addenda as s follows: 4 (continued) BWR/6 STS 5.0-24 Rev. O, 09/28/92

Enclosura 2 to GNRO.94/00056 Chapter 5 Pags 112 C,1 hydires) Programs'andManuals v 5.5 5'.5 ( c a d r @ Programs E+ and Manuals g 7.2,.12 Inservics Testing Program (qontinued) x s \\ ASME oiler and Pressure \\ s Vessel ode and N K g \\equiredFrequencies applicab Addenda \\ terminolog for inservice t ting performing inservice activities tes ino activities l W ly At lea once per 7 days Mont y At least once per 31 days Quarte ly or every 3 mon s At least on per 92 days \\ Semiannua ly or j every 6 .nths At least once. r 184 days Every 9 months At least once pe 276 days Yearly or annu lly t least once per 66 days f { Biennially or e ry f 2 years At east once per 73 days c. The rovisions of SR 0.2 are applic le to the above f ) requi ed Frequencies fo performing ins vice testing activi 'es; d. The provis'ons of SR 3.0.3 a applicable to 'nservice testing act ities; and e. Nothing in the SME Boiler and Pressure Vessel Code shall be construed to sup sede the requirements of any TS. 4.J,1. 9 Ventilation Filter Testing Program (VFTP) s.s. 6 A program shall be established to implement the following required testing of Engineered Safety Feature (ESF) filter ventilation systems at the frequencies specified in CReWe:ry Sid: U b AG-d-6 rcordenea-with3 Regulatory Guide 1.52, Revision 2,0%

J" $59: 2M wr 1].

(continued) BWR/6 STS 5.0-25 Rev. O, 09/28/92

h e h r a) Programs a

Ma uals 0I k +:+- ss C^reair regramp and Manuals -5,it 2d Ventilation Filter Testing Program (VFTP) (continued) r T3. (, 7 Demonstrate for each of the ESF systems that an inplace test a. of the high efficiency particulatz{ air LHEPA) filters shows a penetration and system bypass < 0.05D when tested in accordance with7 Regulatory Guide 1.52, Revision 2, and @ AN3 N510-19 pat the system flowrate specified belowf2 10% ESF Ventilation System Flowrate 5GT5 4eeo cQ C. R F A j <jooo eb h' b. Demonstrate for each of the ESF systems that an inplace test of the h coal adsorber shows a penetration and system o,oS bypass < %whentestedinaccordancewith9 Regulatory Guide 1.52, Revision 2, and N510-19 at the system ( flowrate specified below} 10% ,g 1 Aust ESF Ventilation System Flowrate [5GTS 4.ao d m CRFA go..c h c. Dehnstrate for each of the ESF systems that a laboratory test of a sample of the charcoal adsorber, when obtained as described in9 Regulatory Guide 1.52, Revision 2}'r shows the methyl iodide penetration less than the value specifie_d 'below when tested in accordance with qr.N Mee m -u > r -- - m. - ..je'n and greater than or equal to the u, relative humidity specified below: \\ ESF Ventilation System Penetration RH j 2 5GT5 .9--

o. n 5 '/.

90 /. CRFA c.n.5'/o \\ 7o% 2 i ~ ~ \\Qgg gy Q l.%, W Y ) g jdor q fos l Mon Cab 9 (continued) BWR/6 STS 5.0-26 Rev. O, 09/28/92

r Enclosurs 2 to GNRO 94/C0056 Chaptst 5 Pest 114 Qr;;; der;$;j r c/ Programs and Manuals 3.7 s,6 -+.+ k::duM> Programs and Manuals 4 -5 '.2.13 Ventilation Filter Testing Program (VFTP) (continued) 5,5. (o F-h Note: Allowable penetration = [100% - iodide efficiency or c 'ted in staff evaluation]/ (safety factor). Safet W 5' for systems with heaters. e 7' for systems without heaters. = d. Demonstrate for each of the ESF systems that the pressure drop across the combined HEPA filters, the prefilters, and the charcoal adsorbers is less than the value specified belowwhentestedinaccordancewitWRegulatoryGuide1.52, Revision 2, and 0-19 ] at the system flowrate specified below^[2 10%)'?

s ESF Ventilation Sys m Delta P Flowrate

~ S GT5 ,W'G p4 y% 4o.uvrn C,6FA 7 0 LG 4, o csm Demonstrate that the heaters for each of the ESF systems e. dissipate the value specified below$-10"d*wjlefLtested in S accordance witH*T y N510-19 M u.pt 4.< y a p a e ( ~ wsr % l-w cr< n e a y,., g ESF Ventilation System Wattait L e, 4 SGTS %3) .sz, A_q u g,a w A .a C. R 7 b .Q4.y2.i ku The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the VFTP test frequencies. 5.7.2.14 Explosive Gas and Storage Tank Radioactivity Monitoring Program S. 5'. 7 This program provides controls for potentially explosive gas mixtures contained in the*[iiaste Ce5,;eldupjy;t:4 [th; que.tity y Of r:di:::ti.ity cent;in:d in g : :tcrege ter.a's ;r f:d Ste thg offgas treatment system, and the quantity of radioactivit c.nknW contained in unprotected outdoor liquid storage tanks 3 g33;;;; -. 4,u,. + n, m, .. 45:it;,;,,;; _ ;____ n,a -a1iese C1f (continued) BWR/6 STS 5.0 27 Rev. O, 09/28/92

c?$.0"100%Y" Cl + N ' " =d h Programs and Manuals k t:+ U 5',5 +.-7he4M Programs, and Manuals +:h2.1+ Explosive Gas and Storage Tank Radioactivity Monitoring Prggram 5'. 5.7 (continued) (m6agy in [ Branch Technical Position (BTP) ETSS 11Ds h"Pqstula Radt etive Rele e due to ste Gas Sysi Leak or O7, - a(cor nce wi [Stan-rdReview lan, Sect'on 15.7.3, ostulated Fail re"). he li d radwast quantitie shall be d ermined in RaMoac 've Rei se due p Tank Fai reg The program shall include: The limits for concentrations of hydrogen D d = vi d in the a. rv1S n cabnr phte ;n "cidw syd and a surveillance program to e,qvys w,J ensure the limits are maintained. Such limits shall be g g,m appropriate to the system's design criteria (i.e., whether or not the sysy g is d ed to withstand a hydrogen explosion); QW.. Ps, b. A sury illance program to en ure that the uantity of dioac vity con ned in [e gas storag tank and fed i the fgas trea nt syste is less tha the amount ha ould uit in a ole body osure of a .5 rem to i ividual n an unre ricted are in the e t of [an un ntr led rele e of the nks' cont ts);and 1 bh. A surveillance program to ensure that the quantity of radioactivity contained in all outdoor liquid radwaste tanks that are not surrounded by liners, dikes, or walls, capable of holding the tanks' contents and that do not have tank overflow and surr_ounding area drains connected to the l OB1 4{Jiquid adwaste 9:2;;;;;;3t jystemfis less than the amount t.1at wou d result in concentrations less than the limits of 10 CFR 20, Appendix B. Table Column 2, at the nearest potable water supply and the nearest surface water supply in OC9 an unrestricted area, in the vent of an uncontrolled release of the tanks' contents. g The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Explosive Gas and Storage Tank Radioactivity Monitoring Program surveillance frequencies. (continued) BWR/6 STS 5.0-28 Rev. O, 09/28/92

... ~ _. I Encbsurs 2 to GNRO-94/000f4 Chapler 5 Pags 116

h n d.rc h Programs, and Manuals i

7 r:::tr@ Programy and Manuals Q..: rra...,,a a.,~.n - careg t 5.7.2.15 Diesel Fuel Oil Testing-Program 5.5 % A diesel fuel oil testing program to implement required testing of both new fuel oil and stored fuel oil shall be established. The program shall include sampling.and testing requirements, and acceptance criteria, all in accordance with applicable ASTM Standards. The purpose of the oronram is to ettahlich following:

a n A e c c.vbbW%t e# d o ce.

O c. ) i) <.v u n 9 :t de a. Acceptability of-new fuel oil for. use prior to addition to i storage tanks by detemining that the fuel oil has: A o A c-

A 5<.)' ^ ^' 1.

Cr! -=cHe r -+'-M 3 "' cr2"f G within t c"y 1 limits, Q ~ 2. kinematic viscosity within limits for- { e cieer ::d bri3 f elp;;r;r.:: af th ;;r;;;r ::':[ jg [g g b. =:r x:: W@ ASTM 20 fuel a[U ace within limitt+ within days %.:u nc :- ~ m -^ addition to storage M " h

pjq tanks; an

[, g c. Total particulate concentration of t fuel oil is s S mg/l g ;I when. tested every 31 days in accordance with A y 22'b _- w :: a er - 2. 1 gwg 5.7. (16 Fire rotectio rogra ? I 's pro a provid contr s to ensure at appropri e fire. 3 pro etion asures a inta ed to prote the. plant om fire and t ensure he capab to ieve and tain safe utdown in the vent o fire is intaine l ..] u BWR/6 STS' 5.0-29 Rev. O, 09/28/92' ,,J_. .--n.. n r to GNRO.94/C0056 Chapter 5 Pag:e 117 CSFOP ?W es Pwg-s e J r% Ig Proyam5 o~3 r%ods J m ~ 5.0 ADMINISTRATIVE CONTROLS E'5 j 5,D -^ M Safety Function Determination Program (SFDPj @ c.-r k b [c o '3 Q "W This program ensures loss of safety functinn ie afected ana 4-e i ir,c re LCM appropriate actions taken. Upon u iiuic OO2 tu..- <-u an evaluation shall be made to determine if loss of safety function exists. Additionally, other appropriate limitations and remedial or compensatory actions may be identified to be taken as a result of the support system inoperability and corresponding exception to entering supported system Condition and Required Actions. This program implements the requirements of LCO 3.0.6. The SFDP shall contain the following: a. Provisions for cross division checks to ensure a loss of the capability to perform the safety function assumed in the accident analysis does not go undetected; b. Provisions for ensuring the plant is maintained in a safe condition if a loss of function condition exists; c. Provisions to ensure that an inoperable supported system's Completion Time is not inappropriately extended as a result of multiple support system-inoperabilities; and d. Other appropriate limitations and remedial or compensatory actions. A loss of safety function exists when, assunting no concurrent single failure, a safety function assumed in the accident analysis-cannot be perfomed. For the purpose of this program, a loss of safety function may exist when a support system is inoperable, and: a. A required system redundant to system (s) supported b the inoperable support system is also inoperable 'Cn:

or b.

A required system redundant to system (s) in turn supported $ nc $ ; perable supported system is also inoperable by the ino I or (continued) BWR/6 STS 5.0-30 Rev. O, 09/28/92

Encbsure 2 to GNRO-9100056 i Chapter 5 Pags 118 CI ?ftf \\ ,~ f?:S m:r w A man ho Com5 w& ds (5.5~.N_$V00 N (continued) J A required system redundant to support system (s) for the c. d systems (a) and (b) above is also inoperable o ^ Gene 'c M e: N Division B i System i System i --Ca s e C \\ 4 4 I System ii port System System ii ~ 4 Inoperable) 4 m iii Sys iii +-Case A 4 4 System iv System iv se B wQ The SFDP identifies where a loss of safety function exists. If a i l loss of safety function is detemined to exist by this program, the appropriate Conditions and Required Actions of the LCO in which the loss of safety function exists are required to be entered. 1 BWR/6 STS 5.0-31 Rev. O, 09/28/92

$NlrNPSg".$ i Reporting RQquirQmonts 51 4 5.0 ADMINISTRATIVE CONTROLS

s. co

-64 Reporting Requirements ) 1 Routh e4 e The following reports shall be submitted in accordance with 10 CFR 50.4. 5 A 1.1 Startup Rep A sumary report of plant startup and power escalation testing hall be submitted following: /

a. N eceipt of an Operating License; l

b. Amendment to the license involving a pla ed increase in power to 1; c. Installatio fuel that has a di erent design or has been manufactured different fuel upplier; and d. Modifications thathy have s nificantly altered the nuclear, thermal, or'h drau c performance of the unit. The initial Startup Report sh address esch of the startup tests identified in FSAR, Chapter 14], and shall include a description of the measured values of e oper ing conditions or characteristics obtained uring the st program and a comparison of these values with d ign prediction and specifications. Any corrective actions t t were required to btain satisfactory operation shall als be described. Any ad tional specific details required license conditions base n other commitments shall be include in this report. Subsequent tartup Reports shall address s artup tests that are necessary demonstrate the acceptability f changes and modifications. Startup Re rts shall be submitted within 90 days fo owing I completio of the Startup Test Program; 90 days follow g resumpti or commencement of commercial power operatio or 9 month following initial criticality, whichever is earl st. If 1 j the St tup Report does not cover all three events (i.e., itial I criti lity, completion of Startup Test Program, and resumpt'on or cement of commercial operation), supplementary reports hall i co be s bmitted at least every 3 months until all three events ha e bee completed. j N (continued) BWR/6 STS 5.0-32 Rev. O, 09/28/92 1

Encbsure 2 to GNRO-t4/0C056 Chapter 6 Page 120 C. I 11P' Reporting Requiromonts g er. G Gh ++ Reporting Requirements AF- . DoutinsAs0Ft5 -(tontinue b pl.J.1.2- & a e l ";p r_^S ( 5X]o,1 i- -- N E---------- ) -A---- ..-------N 61 A h qgle s mittal may e made for multiple it station. he \\ ubm1 al sh Id combine \\ctions co e on to all its at the s tion. 9.................N.................................... C e /, Anntre P acper4Wrovering the activities of the uitittac-d : crib 1 f-I h4eFfor the previous endar year shall be submitted by rg Qarch I,LaLeach year 'The 4'iti:1 report tha!! b,e submitted-b4 .. N _u.. a a u. fnin nwinn, 4 n 4. +. 4.,1

4. +. 4.,, i. n..

s nu r wi s .e 4 we wiew - i wy. J v. P 4 % f's -Reports r:quir:d on en 6nnuel bes; E(PO P g' W-' % abe mon b lne i ~ gq c,n p Occupational Radiation Exposure RepofD J A tabulation on an annual basis of the number of stat 1on, i C4 utility, and other personnel (including contractors) receiving exposures > 100 mres/yr and their assoc ated man rem exposure according to work and job function, . g.,- C.':t t k reactor operations and surveillance, inseric_e 1 spection, maintenance}}' waste processing, and refueling). routine main Thi 6 tabulation supplements the requirements of 10 CFR 20. The dose assignments to various duty functions may be uc,s estimated based on pocket dosimeter, thermoluminescent dosimeter (TLD), or film badge measurements. Small exposures totalling < 20% of the individual total dose need not be accounted for. In the aggregate, at least 80% of the -j total whole body dose received from external sources sh uld,, be assigned to specific major work functionsgy, c.a h -[b. Any dhen unit vrdque g rts-+sq" ired an tr=1 basi:.]4 2n (continued) BWR/6 STS 5.0-33 Rev. O, 09/28/92

Encbsure 2 to GNRO-94K10056 Chapter 5 Page 121 Reporting Requiroments C\\ f -5 4

5. (o 5.Ca Reporting Requirements

-(renttmreh_ X. I out4ne-4ese rt 5 a y. Sj.1.3-Annual Radiological Environmental Operating Report e.G.2 ...............................N TE-----.-------------- N 'bmA s/ngle sittal may T made fo a multip unit statio. The -9 tal sho d combine ctions c on to a l units at t bI st tio The Annual Radiological Environmental Operating Report covering the operation of the unit during the previous calendar year shall be submitted by May 15 of each year. The report shall include i sumaries, interpretations, and analyses of trends of the results of the Radiological Environmental Monitoring Program for the reporting period. The material provided shall be consistent with the objectives outlined in the Offsite Dose Calculation Manual l (ODCH), and in 10 CFR 50, Appendix I, Sections IV.B.2, IV.B.3, and IV.C. The Annual Radiological Environmental Operating Report shall include the results of analyses of all radiological environmental sawles and of all environmental radiation measurements taken i during the period pursuant to the locations specified in the table and figures in the 00CM, as well as summarized and tabulated j results of these analyses and measurements #{in the format of the 1 table in the Radiological Assessment Branch Technical Position, _j Revision 1, November 1979}NThe report shall identify the TLD i results that represent collocated dosimeters in relation to the NRC TLD program and the exposure period associated with each result.}* In the event that some individual results are not available for inclusion with the report, the report shall be submitted noting and explaining the reasons for the missing results. The missing data shall be submitted in a supplementary report as soon as possible. 1 (continued) BWR/6 STS 5.0-34 Rev. O, 09/28/92

Encbsure 2 to CNRO-94/00056 Chapter 5 Page 122 Reporting RQQuiromonts i

f. c.

5 (o L 9-Reporting Requirements .1 Ruuune Revuris (centinucu O' 559.1.Q Radioactive Effluent Release Report 5.G.3 ) ................~............... NOTE----------------------..--.-..-3 A single submittal' be made for multiple unit station. The cubmitt Q should com sections co n to all units at the 's ation; R0 wever, for u s with separa radwaste systems, th di su ittal sha specify t releases of ra active material fra each unit. ~ J OMy L The RadLoactive Effluent Release Report covering the operag of the uniiShd ' M ::t::: ec r accor6-^ -QC CIR -50...y. The 3G@, report shall include a summary of the quantities of radioactive ~ liquid and gaseous effluents and solid waste released from the ,LD unit. The material provided shall be consistent with the objectives outlined in the 00CM and Process Control Program and in confonnance with 10 CFR 50.36a and 10 CFR 50, Appendix 1, Section IV.B.I. 5.0.1.5 ; Monthly Operating Reports Routine reports of operatin!11 challenges to the sa ety/ relief statistics and shutdown xperienc including documentation of ._,,7 valves,}& hall be submitted on a monthly basis no later than the 15th of each month following the calendar month covered by the report. CORE OPERATING LIMITS REPORT (COLR) t s.0 C Core operating limits shall be established prior to each a. reload cycle, or prior to any remaining portion of a reload A llowing-cycle,andshallbedocumentedintheCOL{.forthw A The individual pccificatitms-that-address uv e vpersting ' 93 ly-44mi-t+-must be referenced 4ers ^ b. The analytical methods used to detennine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents h (continued) BWR/6 STS 5.0-35 Rev. O, 09/28/92

Chapter 5 Page 123 ' i Enclocure 2 to GNRO-94/00056 ) INSERT 35A .i during the previous calendar year shall be submitted by May 1 of each year 1 l 'l 1 INSERT 5.0-35

Enclosurs 2 o GNRO-94/00056 Chapter 5 Paga 124 Reporting Requir:ments 5.9 i 5.9 Reporting Requirements - 5.0.1.5-CORE OPERATING LIMITS REPORT (COLR) (continued) OC3'? 'Ti G.5 ~_ [ Ident Hy the Top' cal Report ( by nu , title, d e, and g RC staf approva document, identify the staff 5 ety E luation eport fo a plant s cific me odology by RC %.g d \\ let r and te. 3M The core operating (e.mits shall be detemined such that all -li c. applicable limits g., fuel themal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as 50M, transient analysis limits, and accident analysis limits) of the safety analysis are met, d. The COLR, including any midcycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC. Reactor Coolant System (RCS) PRESSURE AND TEMPERATURE LIMITS -15'.0.Cof REPORT (PTLR) Q ] M The RCS pressure and temperature limits, including heatup N L 3, criticality, and hydrostatic and leak test limits, ' M' s:t;..w re: hall be established and documented in_the PTLR. f[The incividu @ / @4 i 3p::tfican=: =:t ederesi th: r-ector vn::1 prn: r =d

t.,,eretere li-it&_ind Uie heetes ad ;eeldea. --te: re 3 jf;r::::d.]j$The analytical methods used to detemine the

~ pressure ana temperature limits in:hding th? 'catr; =d = !d:=( ~ l @ - Arates-shall be those previously reviewed and approved by the NRC in {Tepi n! Deper.(:), ~ ';r, iiilu, dete,.r. ""C

t:ff :pprovat @C ie n e. fer : ;!!nt ::=i#!c d::

=t, Or :t:f' = fety : =leei;vn ethedelv3j ty ""C 1:tter =d dete]. 4Thf reactor vessel pressure and t r Mc! din; da== fer 'tetup a d-n eide= C-Q - { % emperature limits shall be determined so that all applicable limits fers.J-- i) -%eatep li=it:, eteld:= limits, end ~ inse~in !=k and hydrnt:t4e i i revisionsorsupplementsthereto,shallbeprovideduponissuance$ % sting liefts) of the analysis are met.t Ine PTLR, including 7 for each reactor vessel fluency period j _ _Q ~ c-Rm @c,7q Gm'S t ).% Puidan 1 oms 10 C5~ h " ~,\\ p p p24 x s G==d H. / u-.x (continued) BWR/6 STS 5.0-36 Rev. O, 09/28/92

...~ -. Enck>sure 2 to CNRO 94/00050 Chapler $ Pros 125 INSERT 36A 1) XN-NF-79-71(P), " Exxon Nuclear Plant Transient Methodology for Boiling Water Reactors," Exxon Nuclear Company, Inc., Richland, WA. 2) XN-NF-8 0-19 ( P) ( A), Volume 1, " Exxon Nuclear Methodology for Boiling Water Reactors - Neutronic Methods for Design and Analysis," Exxon Nuclear Company, Inc., Richland, WA. 3) XN-NF 19 ( P) ( A), Volume 1, " Advanced Nuclear Fuels Methodology for Boiling Water Reactors: Benchmark Results for the 'CASMO-3G/MICROBURN-B Calculation-Methodology," Advanced Nuclear Fuels Corporation, Richland, WA. 4) XN-NF 19 ( P) ( A), Volume 3, " Exxon Nuclear Methodology for Boiling Water Reactors THERMEX: Thermal Limits Methodology Summary Description," Exxon Nuclear Company, Inc., Richland, WA. 5) ANF-913 ( P) ( A), Volume 1, "COTRANSA2: A Computer Program for Boiling Water Reactor Transient Analysis," Advanced Nuclear Fuels Corporation, Richland, WA. 6) ANF-112 5 ( P) ( A), "ANFB Critical Power Correlation," Advanced Nuclear Fuels Corporation, Richland, WA. 7) XN-NF-84 -105 (P) ( A), Volume 1, "XCOBRA-T: A Computer Code for BWR Transient Thermal Hydraulic Core Analysis," Exxon Nuclear Company, Inc., Richland, WA. 8) XN-NF-573 (P), "RAMPEX Pellet-Clad Interaction Evaluation Code for Power Ramps," Exxon Nuclear Company, Inc., Richland, WA. 9) XN-NF 5 8 (P) ( A), "RODEX2: Fuel ^ Rod Thermal-Mechanical Response Evaluation Model," Exxon Nuclear Company, Inc., Richland, WA. 10) XN-NF 74 (P) ( A), "RODEX2A (BWR) : Fuel Rod Thermal-Mechanical Response Evaluation.Model, "' Exxon Nuclear Company, Inc., Richland, WA. 11) XN-CC-33 (P) ( A), "HUXY: A Generalized Multirod Heatup Code with 10CFR50 Appendix K Heatup Option," Exxon 4 Nuclear Company, Inc., Richland, WA. INSERT 5.0-36 (1)

Encbsura 2 to GNRO 942056 Chaplsr 5 Page 126 12) XN-NF-825 (P) ( A), "BWR/6 Generic Rod Withdrawal Error

Analysis, MCPR for Plant Operation Within the E

Extended Operat.ing Domain," Exxon Nuclear Company, Inc., Richland, WA. 13) XN-NF-81-51(P) ( A), "LOCA-Seismic Structural Response of an Exxon Nuclear Company BWR Jet' Pump Fuel Assembly," Exxon Nuclear. Company, Inc., Richland, WA. 14) XN-NF-84-97 (P) (A), "LOCA-Seismic Structural Response' of an ENC 9x9 BWR Jet Pump Fuel Assembly," Advanced ~ Nuclear Fuels Corporation, Richland, WA. 15) XN-NF-86-37(P), " Generic LOCA Break Spectrum Analysis for BWR/6 Plants," Exxon Nuclear Company, Inc., Richland, WA. 16) XN-NF 07 (P) ( A), " Exxon Nuclear Company ECCS Cladding Swelling and Rupture Model," Exxon Nuclear Company, Inc., Richland, WA. 17) XN-NF-80-19(A), Volumes 2, 2A, 2B, & 2C, " Exxon Nuclear Methodology for Boiling Water Reactors EXEM BWR ECCS Evaluation Model," Exxon Nuclear Company, Inc., Richland, WA. 18) XN-NF 59 (P) ( A), " Methodology for Calculation for Pressure Drop in BWR Fuel Assemblies, " Exxon Nuclear Company, Inc., Richland, WA. INSERT 5.0-36 (2)

$$?.",?10?.%?" Reporting Requirsments -9:i-GG -Se Reporting Requirements (continued) i ],5.9.2 $Decial Reports '-~~-- x \\ Special R orts mayT requiredcovlinginspection, est, and inti; nance ctivitie These specia reports are det ruined on OC3'l a individual asis for ch unit, and eir preparatio and su ittal are d ignated i he Technica Specifications. 3 \\ ports shal ecial be submi d in accor ce with 10 CF 50.4 \\hinth time period ecified or each rep t. Thefoqowing ecial Repo s shall submitted. (s. In the event M CS s actuated and injects wate into the RCS in MODE, 2, or a Special port shall be repared ~~ \\and spbmitte within 90 ays describ g the circum ances of he a tion d the tot accumulate actuation c les to e. curre value o the usage f tor for each af cted fety i ction no le shall b provided in this Qpec al Report when er its va e exceeds .70 CN-IfWinddl emer ) bsNexperfences four or%gency dies 1 generator ( f re valid ilures in the 1 st x CI \\ deman'd these fail res and an nonvalid fallu s \\ e erience y that EDG n that ti period shall bA 39 rep ed wit (n 30 days, eports on failures sh& 1 ( inclu the infqrmation re nded in 1(egulatory Gui 1.9, ( evisic 3, Regulatory Posit n C.5, or hxisting Regulat ry s_ ide 1.108 reporting requirement. w _iiN S ecial iisport is requiredsby Conditio B or o ((. Whe N LCO 3. 1], "Posb4ccident Monita ing (PAN) estrumen ion,'are)s shall be s mitted wi in the 0",,g fo'1 ing 14 s. The r ort shall ou ine the p eplanned a'lte te metho f monitor , the cause f the in rab ity, and e plans a schedule restori the ins nt tion chann s of the nction to ERABLE s atus. (continued) BWR/6 STS 5.0-37 Rev. O, 09/28/92

I Encbsure 2 to GNRO44/00056 Chapter 5 Page 128 Reporting Requirement \\ \\ 5.9 Reporting Requirements 5.9.2 Soecial Reoorts (continued) d. Any abiibrmal gradation of the contain structure detected dur th (tests require the Pre-Stressed Ogg i Concrete Conta nment Tend eillance Program shall be reported to the NR in 0%ys. The report shall include a d ption of the tendbn <ondition, the condition j of t crete (especially at tendon'arteharages), the i ' spection procedures, the tolerances on cr Ma ng, and the corrective action taken. y-i 8WR/6 STS 5.0-38 Rev. O, 09/28/92

Encbsura 2 to GNRO-9400056 ChapttrS Pags 129 Record Retention 5.10 5.0 ADMINISTRATIVE CONTROLS NO Record Retention 7 5.10.1 The following records shall be retained for at leas 3 years: All License Event Reports required by 10 C 50.73; b. ecords of changes made to the procedu s required by S cification 5.7.1.1; and c. Records of radioactive shipments. 5.10.2 The following re rds shall be r ained for at least 5 years: a. Records and lo of uni operation covering time intervals at each power e 1; Og) b. Records and logs o rincipal maintenance activities-inspections, rep r, d replacement of principal items of equipment relat d to nu lear safety; c. Records of s eillance ac vities, inspections, and calibratio required by the Technical Specifications (TS) [and he Fire Protection regram]; d. Record of sealed source and fis on detector leak tests and resul s; and e. Re rds of annual physical inventory f all sealed source erial of record. 5.10.3 Th following records shall be retained for the uration of the u t Operating License. i Records and drawing changes reflecting unit de ign modifications made to systems and equipment des ibed in the FSAR; b. Records of new and irradiated fuel inventory, fuel transfers, and assembly burnup histories; Records of radiation exposure for all individuals ente 'ng c. radiation control areas; (continued) BWR/6 STS 5.0-39 Rev. O, 09/28/92

Encbsura 2 to GNRO-94/00050 Chapter 5 - - Pags 130 - Record Retention 5 10 1 5.10 Record Retention / 5)10.3 (continued) d. Records of gaseous and liquid radioactive mate al released \\ te the environs; Records of transient or operational cycler for those unit components identified in [FSAR, Section ;; 1 'l f. cards of reactor tests and experi ts; j g. Rec ds of training and qualific ion for members of the ] unit aff; h. Records o inservice inspect'ons performed pursuant to the TS; '/ / i. Records of qua ity assur nce activities required by the Operational Qua 'ty As rance (QA) Manual [not listed in 1 Specification 5.1 1 nd which are classified as permanent records by applica regulations, codes,andstandards]; J. Records of revi per ormed for changes made to procedures, equipment, or r iews of tests and experiments pursuant to 10 CFR 50.59; ~' k. Records of t reviews and a its required by Specification' 5.5.1 and S ecification 5.5.2; i D 1. Records o -the service lives of a hydraulic and mechanical snubber required by [ document wher snubber requirements reloca d to], including the date at ich the service life coueer es, and associated installation nd maintenance reco is; m. Re rds of analyses required by the Radiolo ical E ironmental Monitoring Program that would rait aluation of the accuracy of the analysis at later date these records'should include procedures effect e at specified times and QA records showing that these rocedures j werefollowed); Records of reviews performed for changes made to the ffsite Dose Calculation Manual and the Process Control Progra and o. Records of pre-stressed concrete containment tendon surveillances. [ BWR/6 STS 5.0-40 Rev. O, 09/28/92

=. [High Radiation Area' 'l

to GNRO-94M0056 Chapter 5 P. age 131 5.0 ADMINISTRATIVE CONTROLS

[5K1 HighRadiationArea) [. 5.11.1 Pursuant to 10 CFR.20, paragraph 20.203(c)(5), in lieu f the requirements of 10 CFR 20.203(c), each high radiatio area, as defined in 10 CFR 20, in which the intensity of ra ation is 100 mres/hr but < 1000 mres/hr, shall be barri ded and co picuously posted as a high radiation area d entrance thereto sha be controlled by requiring issuance of Radiation Work Perni (RWP). Individuals qualified in ra 'ation protection procedu s (e.g., [ Health Physics Technic ns]) or personnel i continuou ly escorted by such individua may be exep t from the. RWP issuan requirement during the pe ormance of the assigned duties in hi radiation areas with posure rates. s 109.) mrem /hr, I y provided they e otherwise followi plant' radiation-protection' procedures for e try into such hig radiation areas. Any individual or g up of indi duals permitted to enter such areas shall be provi d with accompanied by.one or more of the following: a. A radiation monitori device that continuously indicates the radiation dose at in the area. b. A radiation mon oring dev ce that continuously' integrates the radiation' se rate in e area and alarms when a preset integrated d e is received. ntry into such areas with this monito ng device may be de after the dose rate levels in e area have been est lished and,sersonnel are aware of hem. c. An in vidual quali'fied in radiation otection procedures with a radiation dose rate monitoring vice, who is re onsible for providing' positive contr over the a ivities within the area and shall perfo periodic adiation surveillance at the frequency spe fied by the [ Radiation Protection Manager] in the RWP. 5.11.2 n addition to the requirements of Specification 5.11. areas with radiation levels a 1000 mres/hr shall be provided w'th locked-or continuously guarded doors to prevent unauthorized ent and' the keys shall be maintained under the administrative cont 1 of the Shift Foreman on duty or health physics supervision. Do s-shall remain locked except during periods of access.by person 1 (continued) BWR/6 STS 5.0-41 Rev. O, 09/28/92 t

~ ~- ._ to GNRO.94/00056 Chapter 5 Page 132 [High Radiation Area' j [5.115 HighRadiationArea) 5.11.2 (co inued). ~ under approved RWP that shall specify e dose rate levels in the imme ~ te work areas and the maxi allowable stay times for individuals those areas. In li of.the stay time specification o e RWP, dire or remote (such as closed circuit

2 TV cameras) continu surv ance may be made by personnel d

qualified in radiation ection procedures to provide positive exposure control over e ivities being perforTned within the area. -) 5.11.3 For individ high radiation areas h radiation levels of > 1000 mr /hr, accessible to personnel, hat are located within large eas such as reactor ccatainment, w e no enclosure exists-for urposes of locking, or that cannot be co 'nuously guarded, d where no enc?.osure can be reasonably constru d around the individual area, that individual area shall be barr ded and conspicuously posted, and a flashing light shall be ac 'vated as a warning device. / BWR/6 STS 5.0-42 Rev. O, 09/28/92

I

e.";;-'

ATTACHMENT 2B ITS - PSTS COMPARISION DOCUMENT DISCUSSION OF CHANGES -i

i Enc!asure 2 to GNRO-M/00056 Chapter 5 Page 134 DISCUSSION OF CHANGES TO NUREG-1434 CHAPTER 5 - ADMINISTRATIVE CONTROLS ] BRACKETED ADMINISTRATIVE CHOICE B.1 Brackets removed and optional wording preferences revised to reflect appropriate plant specific requirements. PLANT SPECIFIC DIFFERENCE P.1 This comment number is not used for this station. P.2 Plant specific staff qualifications have been previously ~ reviewed and approved as identified in NUREG-0831, as supplemented. P.3 This comment number is not used for this station. P.4 This comment number is not used for this station. P.5 The safety analysis report for this station is identified as the Updated Final Safety Analysis Report and'is correctly referred to as the UFSAR. P.6 This comment number is not used for this station. P.7 The references to "each" unit and "on the unit affected" has been revised to reflect that only one unit is located at this site. P.8 This requirement is only applicable to PWRs. P.9 The current plant specific requirement details for. diesel generator fuel oil testing are retained in the proposed program. P.10 Where possible, plant specific management position titles in the proposed Technical Specifications are replaced with generic titles as provided in ANSI /ANS 3.1. Personnel who fulfill these positions are required to meet specific qualifications as detailed in proposed Specification 5.3, and compliance details relating to the plant specific management position titles are identified in licensee controlled documents (such as the UFSAR). The two major specific replacements are the generic " plant manager" for the manager level individual responsible for the-overall safe operation of the plant and the generic descriptive use of "the corporate executive responsible for overall plant nuclear safety" in place of the Vice President position. The plant specific titles fulfilling the duties of these generic positions will continue to be defined, established, q l

.- _ +- ._. to GNRO-DO00056 Chapter 5 Page 135 DISCUSSION OF CHANGES TO NUREG-1434 CHAPTER 5 - ADMINISTRATIVE CONTROLS PLANT SPECIFIC DIFFERENCE (continued) documented and updated in a plant controlled document with specific regulatory review requirements for changes, such as the UFSAR or OQAM. This approach is consistent with the intent of Generic Letter 88-06 which recommended, as a line item improvement, relocation of the corporate and unit organization charts to licensee controlled documents. The intent of the Generic Letter, and of this proposed change, is to reduce the unnecessary burden on NRC and licensee resources being used to process changes due solely to-personnel titles changes during. reorganizations. Since this change does not eliminate any of the qualifications, responsibilities or requirements for these personnel or the positions, the change is considered to be a change in presentation only and is therefore administrative. The use of generic titles will decrease the administrative burden on both the utility and the NRC associated'with Technical Specification changes due to reorganizations and title changes which do not affect the functions of these positions. l P.11 This comment number is not used for this station. l P.12 This comment number is not used for this station. P.13 This comment number is not used for this station. P.14 This comment number is not used for this station. P.15 This comment number is not used for this station. P.16 This comment number is not used for this station. P.17 This comment number is not used for this station. P.18 The reference to Generic Letter (GL) 82-33 has been deleted since this GL does not cover all requirements of NUREG - 0737 as presently reflected in the current Technical Specifications. P.19 Appropriate clarification has been provided for sampling of diesel fuel oil. P.20 This comment number is not used for this station. P.21 The discussion has been modified to more accurately describe the safety / relief valves being discuscad. +w--- hea e s ---- --w-r

- Enctosure 2 to GNRO-94/C0056 Chapter 5 Page 136 DISCUSSION OF CHANGES TO NUREG-1434' CHAPTER S - ADMINISTRATIVE CONTROLS PLANT SPECIFIC DIFFERENCE (continued) P.22 This comment number is not used for this station. P.23 This comment number is not used for this station. t P.24 This change has been made to clarify that the limitations on the concentrations of radioactive material released in effluents to unrestricted areas, conform to ten times the concentration values in 10CFR 20, Appendix B. P.25 Plant specific criteria has been provided for limitations on the dose rate resulting from radioactive material. released in gaseous effluents to areas beyond the site boundary. P.26 This comment number is not used for this station. P.27 This comment number is not used for this station. P.28 This comment number is not used for this station. P.29 This comment number is not used for this station. P.30 This comment number is not used for this station. P.31 The plant specific requirements for the component cyclic or transient limit program are identified. P.32 The PTLR discussion is modified to reflect the retention of requirements in the associated LCO and editorial changes are made to reduce the repetition of information. As part of the retention of~ specific limits in ITS SRs 3.4.11.8 and 3.4.11.9 the acceptance criteria for these SRs were changed to include the limit (e.g., less than or eaual to 100*F/50*F). This was an administrative change in presentation only to provide consistency in the limits identified and resulted in no real change to the limits. 'I e w -w

2. m ~._ Enc 6sure 2 to GlMO-94/00056 Chapter 5 Page 137 1 DISCUSSION OF CHANGES TO NUREG-1434 CHAPTER 5 - ADMINISTRATIVE CONTROLS C_HANGE/ IMPROVEMENT TO NUREG STS C.1 The sections related to Technical Specification Bases Control, Procedures, Programs and Manuals, and the Safety Function Determination Program have been reformatted to include all programs in the Programs and Manuals section. Additionally, the Procedures requirements have been separated into an individual section apart from the Programs and Manuals. Appropriate renumbering of these sections.and the ones that follow them are also incorporated, along with revisions to references to these sections to reflect these changes and the relocation of requirements to licensee control conforming with the changes suggested by the NRC in the letter W. T. Russell (NRC) to the Chairpersons of the Owners' Groups Technical Specifications Committees dated-October 25, 1993. C.2 This reflects editorial changes to the wording of NUREG-1434. C.3 This comment number is not used for this station. C.4 Procedures to implement the Emergency Plan and the Security Plan are required by 10 CFR 50,_ Appendix E and 10 CFR 50.54 (p). Therefore, there is no need to repeat the requirements. This change conforms with the changes suggested'by the NRC in the letter W. T. Russell (NRC) to the Chairpersons of the Owners' Groups Technical Specifications Committees dated October 25, 1993. C.5 This comment number is not used for this station. C.6 This change is to provide consistency of requirements for the items specified as Manuals. This change prevents ? po*.ential confusion and misinterpretation. C.7 This change editorially _ rewords the requirement. The change avoids unnecessary confusion that may develop in attempting to determine the intent of the wording. C8 This comment number is not used for this station. C.9 10 CFR 20, Appendix B, has been revised such that the correct reference is Table 2 rather than Table II. C.10 The change avoids unnecessary confusion that may develop in attempting to determine the intent of different wording. C.11 This change is made to provide consistency with the generic surveillance frequencies.

Enck>sure 2 to CNRO-94/00056 Chapler 5 Page 138 DISCUSSION OF CHANGES TO NUREG-1434 CHAPTER 5 - ADMINISTRATIVE CONTROLS CHANGE / IMPROVEMENT TO NUREG STS -(continued) C.12 The description of the entry conditions into the SFDP are clarified and generalized to assure that they include all possible required entry conditions. C.13 This comment number is not used for this station. C.14 This comment number is not used for this station. C.15-This comment number is not used for this station. C.16 This comment number is not used for this station. C.17 This comment number is not used for this station. C.18 This comment number is not used for this station. C.19 These changes provide for consistency with the new 10 CFR 20. C.20 This comment number is not used for this station. C.21 This change removes a requirement that is only applicable to Pressurized Water Reactors (PWRs). I C.22 This comment number is not used for this station. i C.23 These changes reflect editorial corrections to the wording of NUREG-1434. These changes conform with the changes suggested by the NRC in the letter W. T. Russell (;NRC) to the Chairpersone of the Owners' Groups Technical Specifications Coemittees dated October 25, 1993. C.24 The requirements of Table 5.2.2-1 are removed for the Technical Specifications and will be controlled by the licensee's administrative controls. 10 CFR 50.54 provides the requirements for shift complement regarding licensed operators. Additionally, Section 5.2.2.b specifies when a licensed operator must be in the control room. The Table 5.2.2-1 requirements associated with the auxiliary (non-licensed) operators are retained as 5.2.2.a and 5.2.2.c retains the allowance for unexpected absences. This change conforms with the changes suggested by the NRC in the letter W. T. Russell (NRC) to the Chairpersons of the Owners' Groups Technical Specifications Committees dated October 25, 1993 with editorial rewording of 5.2.2.a to be consistent with the wording of 5.2.2.b, the addition of the

- to GNRO-94/00056 Chapter 5 Page 139 DISCUSSION OF CHANGES TO NUREG-1434 CHAPTER 5 - ADMINISTRATIVE CONTROLS CHANGE / IMPROVEMENT TO NUREG STS (continued) allowance for unexpected absences of an auxiliary operator which was allowed on Table 5.2.2-1, and the addition of the allowance for an on shift SS cv SRO to perform the STA function as was allowed by Table 5.2.2-1. C.25 The requirements for SRO presence during fuel handling and core alterations are contained in 10 CFR 50.54. Therefore, there is no need to repeat the requirements. This change conforms with the changes suggested by the NRC in the letter W. T. Russell (NRC) to the Chairpersons of the Owners' Groups Technical Specifications Committees dated October 25, 1993. C.26 The requirements for the STA to meet the qualifications specified by the Commission. Policy Statement on Engineering Expertise on Shift is moved from 5.3.1 to 5.2.2.g. This change conforms with the changes suggested by the NRC in the letter W. T. Russell (NRC) to the Chairpersons of the Owners' Groups Technical Specifications Committees dated October 25, 1993. C.27 The retraining and replacement training program requirements for the unit staff are relocated. Details of this program can be adequately controlled by the licensee's administrative controls. Technical Specification Section 5.3, Unit Staff Qualifications, provides adequate requirements to assure an acceptable, competent unit staff. In accordance with Technical Section 5.3 and GGNS commitments, members of the unit staff meet or exceed.the minimum qualifications of the specific Regulatory Guides or ANSI Standards for their positions. Additionally, Technical Specification Section 5.2, Organization, details unit ~ staff requirements. Sections 5.2.2.a and 5.2.2.b and 10 CFR 50.54 describe the minimum shift crew composition and delineates those positions which require an RO or SRO license. Training and qualification for these positions are specified in 10 CFR 55. This change conforms with the changes suggested by the NRC in the letter W. T. Russell (NRC) to the Chairpersons of the Owners' Groups Technical Specifications Committees dated October 25, 1993.

Enc 6sure 2 to GNRO-94A)0050 Chapter 5 Page 140 DISCUSSION OF CHANGES TO NUREG-1434 CHAPTER 5 - ADMINISTRATIVE CONTROLS CHANGE / IMPROVEMENT TO NUREG STS (continued) C.28 The review and audit requirements can be adequately controlled by the licensee's administrative controls. The control of the these requirements would rely-on the Operational Quality Assurance Manual implementing 10 CFR 50.54 and 10 CFR 50, Appendix B to control the requirements in conjunction with the UFSAR and procedures. Such an approach would result in an equivalent level of regulatory authority while providing for a more appropriate change control process. The net affect of the change is the level of safety of facility operation is unaffected and NRC and utility resources associated with processing license 1 amendments to this Administrative Control _are optimized. The onsite review function, composition, alternate membership, meeting frequency, quorum, responsibilities, authority and records are all covered in equivalent detail in ANSI N18.7-1976 and the current Technical Specification requirements will be contained in the UFSAR. The offsite review group is also addressed, although with less detail, in ANSI N18.7-1976. The UFSAR will include the requirements for the offsite review group. Audit requirements are specified in the QA Program to satisfy 10 CFR 50, Appendix B, Criterion XVIII. These audit requirements will be augmented by the inclusion in the UFSAR those audit requirements currently contained in the Technical Specifications. In addition, audits are also covered by ANSI N18.7, ANSI N45.2, 10 CFR 50.54 (t), 10 CFR 50.54(p), and 10 CFR 73. Therefore, duplication of the requirements contained in the above documents by the Administrative Controls Section of the Technical Specifications does not enhance the level of nuclear safety for the unit. Therefore, the provisions relating to audits are not necessary to assure safe operation of the facility. This change conforms with the changes suggested by the NRC in the letter W. T. Russell (NRC) to the Chairpersons of the Owners' Groups Technical Specifications Committees dated October 25, 1993. References to these controls are corresponding moved to the license's administrative controls.

Enc 6sure 2 to GNRO-94/C@S6 Chapler 5 Page 141 DISCUSSION OF CHANGES TO NUREG-1434 CHAPTER 5 - ADMINISTRATIVE CONTROLS CHANGE / IMPROVEMENT TO NUREG STS (continued) C.29 The details of review and approval requirements for procedures and programs can be adequately controlled by the licensee's administrative controls. These requirements will be included in the UFSAR. This proposal is based on the existence of the following requirements which are duplicative of 10 CFR 50.36 in these areas and which assure operation of the facility in a safe manner. The requirement for procedures is mandated by 10 CFR 50, Appendix B, Criterion II (Second sentence) and Criterion V. ANSI N18.7-1976, which is an NRC Staff-endorsed document used in the development of many licensee QA plans, also contains specific requirements related to procedures. ANSI N18.7-1976, Section 5.2.2 discusses Procedure Adherence. This section clearly states that procedures shall be followed, and the requirements for use of procedures shall be prescribed in writing. ANSI N18.7-1976 also discusses temporary changes to procedures, and requires review and approval of procedures to be defined. ANSI N18.7-1976, Section 5.2.15 describes the review, approval and control of procedures. The section describes the requirements for the Licensee's Quality Assurance program to provide measures to control and coordinate the approval and issuance of documents, including changes thereto, which prescribe all activities affecting quality. The Section further states that each procedure shall be reviewed and approved prior to initial use. The reviews required are also described. Licensees can continue to implement the requirements of 10 CFR 50, Appendix B, regarding procedures without duplicating the necessity of procedure requirements in the facility Technical Specifications. Safe operation of the plant will continued to be maintained, and therefore, the requirements for procedures and their control should not be re-addressed in Technical Specifications. Duplication of the provisions related to procedures is not necessary to assure safe operation of the facility. Procedure control requirements are already addressed in the Operational Quality Assurance Manual and those requirements will be supplemented by the addition of the current' Technical Specification requirements into the UFSAR. This change would result in an equivalent level of regulatory authority while providing for a more appropriate change control process. The net affect of the change is the level

Encbsure 2 to GNRO-94/C3056 Chapter 5 Page 142 DISCUSSION OF CHANGES TO NUREG-1434 CHAPTER 5 - ADMINISTRATIVE CONTROLS CHANGE / IMPROVEMENT TO NUREG STS (continued) of safety of facility operation is unaffected and NRC and utility resources associated with processing license amendments to this Administrative Control are optimized. This change conforms with the changes suggested by the NRC in the letter W. T. Russell (NRC) to the Chairpersons of the Owners' Groups Technical Specifications Committees dated October 25, 1993. C.30 The requirement for procedures to implement the personnel radiation protection requirements of 10 CFR 20 can be adequately controlled by the licensee's administrative controls. These procedures are developed to ensure nuclear plant personnel safety and have no impact on nuclear safety. Additionally, nuclear plant personnel are not " members of the public." Thus, the principal operative standard in Section 182a of the Atomic Energy Act; " health and safety of the public" does not apply. Based on these considerations, the Radiation Protection Program administrative control is not necessary to assure operation of the facility in a safe manner and can be deleted from Technical Specifications. The requirement to have procedures to implement Part 20 is also contained within 10 CFR 20.1101(b). Periodic review of these procedures is addressed under 10 CFR 20.1101(c). This change conforms with the changes suggested by the NRC in the letter W. T. Russell (NRC) to the Chairpersons of the Owners' Groups Technical Specifications Committees dated October 25, 1993. C.31 The requirement for a Process Control Program (PCP) and its review and approval requirements can be adequately controlled by the licensee's administrative controls. The PCP can be adequately described in another controlled documents, e.g., the ODCM and the UFSAR. Control of changes is preserved by 10 CFR 50.54a. The PCP implements the requirements of 10 CFR 20, 10 CFR 61, and 10 CFR 71. Relocating the description of the PCP does not affect the safe operation of the facility. This change conforms with the changes suggested by the NRC in the letter W. T. Russell (NRC) to the Chairpersons of the Owners' Groups Technical Specifications Committees dated October 25, 1993.

l Endosure 2 to GNRO-94/00056 Chapter S Page 143 DISCUSSION OF CHANGES TO NUREG-1434 CHAPTER 5 - ADMINISTRATIVE CONTROLS CHANGE / IMPROVEMENT TO NUREG STS (continued) C.32 The radiological environmental monitoring program is required to be part of the ODCM, the additional details of the program can be adequately controlled by the licensee's ) administrative controls. The ODCM requirements in conjunction with the UFSAR will continue to control this program. This change conforms with the changes suggested by the NRC in the letter W. T. Russell (NRC) to the Chairpersons of the Owners' Groups Technical Specifications Committees dated October 25, 1993. C.33 The requirement for an in plant radiation monitoring program and the associated details of the program can be adequately controlled by the licensee's administrative controls. This program provides controls to ensure the capability to accurately determine the airborne iodine concentration in-vital areas under accident conditions. This program was developed to minimize radiation exposure to plant personnel post-accident and has no impact on nuclear safety. Additionally, nuclear plant personnel are not " members of the public." Thus, the principal' operative standard in Section 182a of the Atomic Energy Act; " health and safety of the public" does not apply. Based on these corsiderations, the In Plant Radiation Monitoring Program administrative control is not necessary to assure operation of the facility in a safe manner and can be deleted from Technical Specifications. This program will continue to be controlled by the licensee's administrative controls. This change conforms with the changes suggested by the NRC in the letter W. T. Russell (NRC) to the Chairpersons of the Owners' Groups Technical Specifications Committees dated October 25, 1993. C.34 The requirement for an Inservice Inspection Program and the Inservice Testing Program can be adequately controlled by the licensee's administrative controls and the requirements of 10 CFR 50.55a. Removal of the Inservice Inspection Program and the details of the Inservice Testing Program conforms with the changes suggested by the NRC in the letter W. T. Russell (NRC) to the Chairpersons of the Owners' Groups Technical Specifications Committees dated October 25, 1993. to GNRO4t/00056 Chapter 5 Page 144 DISCUSSION OF CHANGES TO NUREG-1434 CHAPTER 5 - ADMINISTRATIVE CONTROLS CHANGE / IMPROVEMENT TO NUREG STS (continued) C.35 The requirement for a Fire Protection Program and the associated details of the program can be adequately controlled by the licensee'c administrative controls. The Fire Protection Program provides controls to ensure that appropriate fire protection measures are maintained to protect the plant from fire and to ensure the capability to achieve and maintain safe shutdown in the event of a fire. The administrative control provides assurance that the capability to provide for alternate / dedicated safe shutdown in accordance with 10 CFR 50, Appendix R. As such, it does not directly assure nuclear safety, but rather allows for the ability to place the unit in a more stable condition in the event of a fire. The deletion of this administrative control from Technical Specifications is also consistent with the guidance in NRC Generic Letter 86-10 " Implementation of Fire Protection Requirements." In that letter, the NRC concluded the provisions of 10 CFR 50.59 should apply directly to changes i the licensee desired to make in the fire protection program so long as those changes did not adversely affect the ability to achieve and maintain safe shutdown. The standard license condition, included within 86-10, stated that changes which adversely affected the ability to achieve and maintain safe shutdown in the event of a fire required prior approval of the Commission. Thus, the license condition established as part of the NRC Generic Letter 86-10 implementation also makes this administrative control unnecessary. Based on these considerations, the Fire Protection Program administrative control is not necessary to assure operation of the facility in a safe manner and can be deleted from Technical Specifications. This change conforms with the changes suggested by the NRC in the letter W. T. Russell (NRC) to the Chairpersons of the Owners' Groups Technical Specifications Committees dated October 25, 1993. C.36 The requirement for a Startup Report and the associated details of the report and when it should be submitted can be adequately controlled by the licensee's administrative controls. The report was a summary of plant startup and power escalation testing following receipt of the Operating License, increase in licensed power level, installation of nuclear fuel with a different design or manufacturer than-

Enc 6sure 2 to GNRO-94/CM56 Chapter 5 P ee 145 DISCUSSION OF CHANGES TO NUREG-1434 CHAPTER 5 - ADMINISTRATIVE CONTROLS CHANGE / IMPROVEMENT TO NUREG STS (continued) the current fuel, and modifications that may have significantly altered the nuclear, thermal, or hydraulic performance of the unit. The report provided a mechanism for NRC to review the appropriateness of licensee activities after-the-fact, but provided no regulatory authority once the report was submitted (i.e., no requirement for Commission approval). The approved 10 CFR 50, Appendix B, Quality Assurance Plan and Startup Test Program provide assurance the listed activities are adequately performed and that appropriate corrective actions, if required, are taken. Given that the report was required to be provided to the Commission no sooner than 90 days following completion of the respective milestone, it was clearly not necessary to assure operation of the facility in a safe manner for the interval between completion of the startup testing and submittal of the report. Additionally, given there is no requirement for the Commission to approve the report, then the Startup Report is not necessary to assure operation of the facility in a safe manner. Based on these considerations, the Start-up Report may be removed from TS and relocated to a licensee controlled document. This. change conforms with the changes suggested by the NRC in the letter W. T. Russell (NRC) to the Chairpersons of the Owners' Groups Technical Specifications Committees dated October 25, 1993. C.37 Since no report descriptions remain in the Special Report sections this section heading is removed, the corresponding Routine Report Section designation is removed, and the items are renumbered to reflect these changes. C.38 The requirement for a Special Report following an ECCS injection and the associated details of the report and when it should be submitted can be adequately controlled by'the licensee's administrative controls. Title 10, Part 50,. Section 73 already provides the requirement for the license-to submit a Licensee Event Report in the event of an ECCS actuation. The report is required to be submitted within 30 R days and will contain the same type of information as the-special report. Removing the duplicative requirement from

Enctosure 2 to GNRO-94X)0056 Chapter 5 Page 140 l DISCUSSION OF CHANGES TO NUREG-1434 CHAPTER 5 - ADMINISTRATIVE CONTROLS CHANGE / IMPROVEMENT TO NUREG STS (continued) TS has no impact on assuring safe operation of the facility since the requirement to submit a report to Commission still exists in 10 CFR 50.73. This change conforms with the changes suggested by the NRC in the letter W. T. Russell (NRC) to the Chairpersons of the 4 Owners' Groups Technical Specifications Committees dated October 25, 1993. l C.39 The requirement for a Special Report following EDG failures, I the associated details of the report, and when it should be submitted can be adequately controlled by the licensee's administrative controls. The report provided a mechanism for NRC to review the appropriateness of licensee activities after-the-fact, but provided no regulatory authority once the report was submitted (i.e., no requirement for Commission approval). Given that the report was required to o be provided to the Commission no sooner than 30 days following the EDG failure, it was clearly not necessary to assure operation of the facility in a safe manner for the interval between the EDG failure and submittal of the report. Additionally, given there is no requirement for the Commission to approve the report, then this Special Report is not necessary to assure operation of the facility in a safe manner. Based on these considerations, this Special Report may be removed from TS and relocated to a licensee controlled document. This change conforms with the changes suggested by the NRC in the letter W. T. Russell (NRC) to the Chairpersons of the Owners' Groups Technical Specifications Committees dated October 25, 1993. C.40 The requirement for a Special Report following extended Post Accident Monitoring instrumentation inoperability and the associated details of the report and when it should be submitted can be adequately controlled by the licensee's administrative controls. The report provided a mechanism for NRC to review the appropriateness of licensee activities after-the-fact, but provided no regulatory authority once the report was submitted (i.e., no requirement for Commission approval). Given that the report was required to be provided to the Commission no sooner than 30 days following the instrumentation failure, it was clearly not necessary to assure operation of the facility in a safe

\\ to GNRO-94/00056 Chapter 5 Page 147 j DISCUSSION OF CHANGES TO NUREG-1434 CHAPTER 5 - ADMINISTRATIVE CONTROLS CHANGE /IM_PJLQ_VJMENT TO NUREG STS (continued) manner for the interval between failure of the instrumentation and submittal of the report. Additionally, given there is no requirement for the Commission to approve the report, then this Special Report is not necessary to assure operation of the facility in a safe manner. Based on these considerations, this Special Report may be removed from TS and relocated to a licensee controlled document. This information has been added to the bases for the LCO Required Actions which required the Special Report to be written. This change conforms with the changes suggested by the NRC in the letter W. T. Russell (NRC) to the Chairpersons of the Owners' Groups Technical Specifications Committees dated October 25, 1993. C.41 Record retention requirements and the associated details of the report can be adequately controlled by the licensee's administrative controls. The requirements on record retention may be deleted from Technical Specifications on the basis that they are adequately addressed by the QA Plan (10 CFR 50, Appendix B, Criteria XVII), that the current Technical Specification specific requirements will be included in the UFSAR, and because provisions relating to record keeping do not assure operation of the facility in a safe manner. Facility operations are performed in accordance with approved written procedures. Areas include normal startup, operation and shutdown, abnormal conditions and emergencies, refueling, safety related maintenance, surveillance and testing, and radiation control. Facility records document appropriate station operations and activities. Retention of these records provides documentation retrievability for review of compliance with requirements and regulations. Post compliance review of records does not assure operation-of the facility in a safe manner as activities described in these documents have already been performed. Numerous other regulations such as 10 CFR 20, Subpart L, 10 CFR 50.71, etc. also require the retention of certain records related to operation of the nuclear plant. This change conforms with the changes suggested by the NRC in the letter W. T. Russell (NRC) to the Chairpersons of the Owners' Groups Technical Specifications Committees dated October 25, 1993.

r_ m Cnclosure 2 to GNROot/00056 - Chaplor 5 Page 148 j 'l DISCUSSION OF CHANGES TO NUREG-1434 CHAPTER 5 - ADMINISTRATIVE CONTROLS CHANGE / IMPROVEMENT TO NUREG STS (continued) C.42 The accepted alternate methods of complying with 10 CFR 20.203(c) for high radiation areas can only be changed with prior NRC approval and can be adequately controlled by the licensee's administrative controls. These controls are developed to ensure nuclear plant personnel safety and have no impact on nuclear safety. Additionally, nuclear plant personnel are not " members of the public." Thus, the principal operative standard in Section 182a. of the Atomic Energy Act; ' health and safety of the public' does not apply. Based on these considerations, the Radiation Protection Program administrative control is not necessary to assure operation of the facility in a safe manner and can be deleted from Technical Specifications. l This change conforms with the changes suggested by the NRC in the letter W. T. Russell (NRC) to the Chairpersons of the Owners' Groups Technical Specifications Committees dated October 25, 1993. C.43 In general the format of the NUREG-1434 does not include the 1 use of cross references. These unneeded cross references are deleted. C.44 The TS need not require an administrative letter be issued to station personnel on an annual basis describing responsibility to the Shift Superintendent. Organizational l responsibilities are adequately described by the stations internal administrative controls. Repeating the organizational responsibilities via an internal management directive only increases the administrative burden on the facility with no resulting benefit. Plant safety is not compromised by this proposed change. C.45 This comment number is not used for this station. C.46 The generic example is removed. _}}

ML20065L497

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