Text

Proposed Tech Specs,Incorporating New Heatup & Cooldown P/T Limit Curves W/Associated Requirements & New Power Operated Valve Setpoint Curves W/Associated Requirements
	ML20217R299
Person / Time
Site:	Millstone
Issue date:	08/29/1997
From:	; NORTHEAST NUCLEAR ENERGY CO.
To:	;
Shared Package
ML20217R297	List: B16636, Application for Amend to License NPF-49,incorporating New Heatup & Cooldown Pressure/Temp Limit Curves W/Associated Requirements & New Power Operated Relief Valve Setpoints Curves W/Associated Requirements; ML20217R299;
References
	NUDOCS 9709050017
	Download: ML20217R299 (159)
	v • d • e

- Docket No. 50-423 B16636 Millstone Nuclear Power Station Unit No, 3 Proposed Revision to Technical Specification Reduction in the Cold Overoressure Protection System Enablina Temperature (PTSCR 3-21-97)

NNECO's Commitments f

T h

August 1997 9709050017 970829 PDR ADOCK 05000423 PDR s

p

- "~

U.S. Nucicar Regulatory Commission B16636%ttachment 1\\Page 1 Enclosure List of Regulatory Commitments The following table identifies those actions committed to by NNECO in this document.

Please notify the Manager - Nuclear Licensing at the Millstone Nuclear Power Station Unit No. 3 of any questions regarding this document or any associated regulatory commitments.

Commitment Committed Date or Outage Procedures will be revised to eliminate a xenon credit Prior to Startup currently allowed when calculating shutdown margin for Modes 4 or 5.

Procedures will be revised to address loss of a RCP in Prior to Startup MODE 5 with no RHR pump in operation (i.e. a new entry condition for Loss of Shutdown Cooling)

i.

1 4

Docket No. 50-423

~

B16636 4

5 a

p i

i 1-I Millstone Nuclear Power Station Unit No. 3 -

Proposed Revision to Technical Specification Reduction in the Cold Overoressure Protection System Enablina Temoerature (PTSCR 3-21-97) "

Marked Up Paaes August 1997

l l

l U.S. Nuclear Regul: tory Commission B16636\\ Attachment 2\\Page 1 l

MARKUP OF PROPOSED REVISION Refer to the attached markup of the proposed revision to the Technical Specifications.

The attached markup reflects the currently issued version of the Technical Specifications listed below. Pending Technical Specification revisions or Technical Specification revisions issued subsequent to this submittal are not reflected in the enclosed markup.

Applicability revised to add MODE 4. Added new Limiting Condition for Operation (LCO) Action for MODE 4 and modified existing Action to apply to MODES 5 or 6.

3.1.2.1 Deleted footnote applying to boron injection flow path operability.

Modified Applicability to delete MODE 4.

3.1.2.2 l

l Revised Applicability to add MODE 4.

Added Action for MODE 4 and revised existing Action to apply to MODE 5 and 6. Deleted 31 day surveillance to verify charging pump motor circuit breaker position when less than or equal to 350 degrees.

3/4.1.2.3 Deleted footnote. Revised Applicability to delete MODE 4. Revised Action MODE change to HOT SHUTDOWN and completion time requirements to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . Deleted 31 day surveillance to verify charging pump motor circuit breaker position when less than or equal to 350 degrees.

3/4.1.2.4 Replaced LCO statement to be consistent with new restrictions on Reactor Coolant Pump (RCP) starts. Revised second footnote to reflect new RCP start criteria.

Added a third footnote to address Reacter Coolant System injection flowpath restrictions.

3.4.1.3 Revised first LCO Action to modify RHR loop Operability requirements. Reworked first footnote to address RHR removal from service.

Revised third footnote to provide more stringent requirements for starting and operating an RCP in MODE 5.

3.4.1.4.1

.r i

[

U.S. Nucl::ar Regulttory Commission-B16636\\ Attachment 2\\Page 2

. Added third footnote that applies to APPLICABILITY and provides more_ stringent requirements for starting and operating an RCP in MODE 5.~

3 E

3.4.1.4.2 i

e L Added new LCO to require all RCPs to be de-energized prior to opening and RCS loop stop valves. Re-identified existing LCOs.

-3.4.1.6 Revise heatup and cooldown limits and add RCP operational restrictions. Added 72 i.

i hour completion time limit to LCO Action.

Clarified Surveillance for leak and i

. hydrostatic testing.

3/4.4.9.1 i

Curves revised to incorporate instrument uncertainties, updated chemistry data, and I

- Reactor Coolant Pump, Charging pump, and Safety. Injection Pump Operational restrictions. Also revised the allowable heatup and cooldown rates to compensate j

for the ~ increased instrumentation uncertainties and system hydraulic

[

considerations, i

Figure 3.4-2 j

Figure 3.4-3 i

Table clarified to state " approximate" withdrawal time.

j l

Table 4.4-5 i

Added 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months completion time requirement for engineering evaluation when i-e temperature limits are exceeded.

3.4.9.2

. = Revised to add pump injection limits and reword pressure relief capabilities. The enabling temperature for-Cold Overpressure protection has been reduced from 350 F to 275 F.

Added footnotes. Consolidated and clarified existing Surveillance

. requirements and added three new sections.

3/4.4.9.3-Replaced Curves which contain increased margins to compensate for instrument

'i uncertainties and system hydraulic losses.-.

i i

Figure 3.4-4a i

5 l

U.S. Nucl:ar Regul:: tory Commission B16636%ttachment 2\\Page 3 Figure 3.4-4b -

Deleted footnote. Clarified OPERABLE ECCS flow path wording. Revised first surveillance requirement to allow for different valve alignment. Deleted second surveillance performing 31 day motor circuit breaker position verifications.

3/4.5.3 v.

The Bases are reworded to reflect the changes B 3/4.1.2 l

B 3/4.4.1 l

B 3/4.4.9 B 3/4.5.2 and 3/4.5.3 1

l l

l

' March al'. 1991 REACTIVITY CONTROL SYSTEMS 3/4;1.2-B0 RATION SYSTEMS-FLOW PATH --SHUTDOWN-

' LIMITING = CONDITION FOR OPERATION 3.1.2.1 As a minimum; one of the following boron injection flow paths shall lbe OPERABLE and capable of being powered from an OPERABLE emergency power.

. source:

a.

- A flow path.from the beric acid storage system via~ either a boric acid transfer pump or a gravity feed connection and a charging pump to the Reactor Coolant System if the boric acid storage system in Specification 3.1.2.5a. is OPERABLE, or b.-

The flow path from the refueling water storage tank via a charging jg pump to the Reactor Coolant System if tha refueling water. storage g,, g.

tank in Specificatien 3.1.2.5b. is OPED h.E.

-APPLICABILITY: MODES and 6.

E112i:

--+

beros o.neckon b With-none of the abovejflow paths OPERABLE or capable of. being powered from

(=

an OPERABLE-emergency power-sourc suspend all _ operations involving CORE ALTERATIONS or positive reactivit changes.

is hcDCS T cv G SURVEILLANCE REOUIREMENTS 4.1.2.1 At least one of the above require'd flow paths shall be demonstrated OPERABLE:

a.-

At least once per 7 days by verifying that the Boric Acid Transfer Pump Room temperature and the boric-acid storage tank solution temperature are greater than or equal-to 67eF when a flow path from the boric-acid tanks'is used, and b.-

At least once per 31 days by verifying that each valve (manual, power-operated, or_ automatic) in the flow path that-is not locked.

sealed, or otherwise' secured in position, is in its correct position.

m C

. MILLSTONE - UNIT 3 3/4 1-13 Amendment No. 60.,

eeer

Insert 3.1.2.1

a. With none of the above boron injection flow paths OPERABLE or capable of being powered from an OPERABLE emergency power source in MODE 4, provide an -

OPERABLE flow path capable of being powered from an OPERABLE emergency power source within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or be in COLD SHUTDOWN within the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

l l

\\

i I

[

H ACTIVITY CONTROL SYSTEMS D3cember 28,1995 FLOW PATHS - OPERATING LIN! TING CONDITION-FOR OPERATION-3.1.2.2 At least twodf the following three boron injection flow paths

.shall be OPERABLE:-

4 a.

The flow path from the beric acid storage ' system via a boric acid transfer pump and a charging pump to the Reactor Coolant System (RCS),and j

~

b.

Two flow path's from the refueling water storage tank via charging pumps to the RCS.gnd3, MODES 1, 2,p,.no C APPLICABILITY:

Ellati:

l With only one of the above required boron injection flow paths to the RCS

^

OPERABLE, restore at least two boron injection flow paths to the RCS to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY and borated to a SHUTDOWN MARGIN equivalent to at least-the limits as shown in Figure 3.1-4 at 200*F within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; restore at least two flow paths to OPERABLE status within the next 7 days or be in40h0x5HUTDOWN within the next.atr hours.

H oT SURVEILLANCE REQUIREMENTS 4.1.2.2 At least two of the above required flow paths shall be' demonstrated OPERABLE-At least once per 7 days by verifying that the Boric Acid Transfer a.-

Pump Room temperature and the boric acid storage _ tank solution temperature are greater than or equal to 67'F-.when it is a

-required water source; b.

At least once per 31 days by verifying that each-valve (manual, power-operated, or automatic) in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct

position, At least once each REFUELING INTERVAL by verifying that each l

c.

automatic valve in the flow path actuates to its correct position on a Safety Injection test signal; and d.-

At least once each REFUELING INTERVAL by verifying that the flow l

path required by Specification-3.1.2.2a.-delivers at least 33 gpm to the RCS.

i

'Oni; ene b;;en irde:tha fh; p:th h 7:;;ir:d t: h CPEPf3LE'uh: :v:r the t=;;r:tur; cf ::: ;r ;;r: Of th: PI! ::ld h;; h h:: th:n ;r :;2:1 10

-tWF MILLSTONE - UNIT 3 3/4 1-14 Amendment No. 79 77,122 om

March Aa, 1993 REACTIVITY CO'ITROL SYSTEMS

(,

CHARGING PUMP --SHUTDOWN LIMITING CONDITION FOR OPERATION 3.1.2.3 One charging pump in the boron injection flow path required by Specification 3.1.2.1 shall be OPERABLE and capable of being powered from an OPERABLE emergency power source.

g r APPLICABILITY: MODESA5 and 6.

3 3, l.O ACTION:

ggfgQg 4.With no charging pump' OPERA 3LE or capable of being owered from an OPERABLE l

emergency power source, suspend all operations invo ving CORE ALTERATIONS or I

positive reactivity changes.

SURVEILLANCE REOUIREMENTS l

4.1.2.3.1' The above required charging pump shall be demonstrated OPERABLE by verifying, on recirculation flow, that a differential pressure across the pump of greater than or equal to 2411 psid is developed when tested pursuant to Specification 4.0.5.

4.1.2.3.2 All :h:rgin; purp, excluding th: b:';; r ;;ir:d ^^E"ASLE pump,

(

hell b; d: ::n:tr:t:d in:per:bl: :t 10::t :n:: per 31 d:y:, except wherA he-7 ;.;t:r v:;;;l h::d i; r:::v:d, by v:rifying th:t the m:ter-e4Muit br::hcrs-

r: :::;r:d in th: : p : r p er i ti er..

(-

MILLSTONE - UNIT 3 3/4 1-15 AmendmentNo.Eg,60 0007

. Insert 3.1.2.3

a. With no charging pump OPERABLE or capable of being powered from an

- OPERABLE emergency power source in MODE 4, provide an OPERABLE charging -

pump capable of being powered from an OPERABLE emergency power source within

- I hour or be in CGLD SHUTDOWN within the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

G I

i-i:

I

March 11, 1991 REACTIVITY CONTROL SYSTEMS ka (B&gGING PUMPS - OPERATING i

LIMITING CONDITION FOR OPERATION At least twoEfchargi g pumps shall be OPERABt E.

3.1.2.4 an 3.

. APPLICABILITY: MODES 1, 2 A, ;.d t.

ACTION:

With only one charging pumo OPERABLE, restore at least two charging pumps to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY and borated to a SHUTDOWN MARGIN equivalent to at least the limit as shown in Figure 3.1-4 at 200*F within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; restore at least two charging pumps to L

OPERABLE status within the next 7 days or be in 40b9xSHUTDDWN within the next Je hours, go y-SURVEILLANCE REOUIREMEt!TS 4.1.2.4.1 At least two charging pumps shall be demonstrated OPERABLE by verifying, on recirculation flow, that a differential pressure across each pump of greater than or equal to 2411 psid is developed when tested pursuant to Specification 4.0.5.

(

0.2.2.i.2 All :h:rgin; pt ;:, ::::;t th: 05::: :ll:'.::d OPEP.^3LE ;" ;, :h:ll I

be d:- n:tr:ted ineper:bl: :t 1:::t ::: p:r 31 d:y: wh:::. r the I

t; +;r;;;r; :f ::: Or :: : Of th:'P:::t:r C::12 t Syrter ("CS) reld le;: i 1;;; th: Or :;::1 t: 250*F by :r*fying th:t th: ::ter etreeft hr:the : tre

r:d in th: :p;r ;;;it'.:n.

' :: 4

- Of ::: :::trife;21 :h:rgin; per; :F-ll 5 CPEP"3LE ah:::v:r th:-

t: ;:r ter ef ::: er ::r: ef the P.CS cold le;: i: 12:: th: er :;r:1 t:

350*f.

i MILLSTONE - UNIT 3 3/4 1-16 Amendment No. 60,,

0007

RE ACTOF COOL Af?T SYSTE?.'.

HOTSHUTDOWN

- Lit.'.lTING CONDITION FOR OPER ATION

'3 t.l.3 At least two of. the reactor coolan: loops, stec below sh 11 be 1

N Y'p-OPERA E, with a: -1 s two reactor colant loops operation w n the n-

/]h React Trip System b eakers are closed At least two i the loops list d below g,h3 shal:

OPERABLE a e at least one of. est loops sha be in opera:!o with the Kea.or Tr!; System retkers opent*

a.

React. Coolan Loop and 1:s ass. lated steam g nerator a..

/

Q reac r coolant pump,*

b. -

E actor Coolant op 2 and its ssoc!sted stet. generator and cactor coolant pu..p,**

{

Reactor Coola.: Loo ts associated

.etm 'generr. r and c.

reactor coola. pump,"p _ 3 and l

Reactor C olan: Loep 4 nd 1:s associat d steam gen rator and reactor c' lant purr.),**

e.

RHR L.; 1, and

(

f.

E H R uoop 2.

APPLIC AE!*. TY: MODE 4.

ACTION:

With less than the above recuired loops OPkRAELE, immeclately a.

in!: late corrective ac:lon to return the required loops'to OPERABLE s:stus as soon as possible; if the remaining OPERABLE loop is an RSR loop, be in COLD SHUTDOTN within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . -

b.

Wl:h no loop in operation, suspend all operations involving a reduction in boren concentration of the _ Reactor Coolant System and immediately initiate corree lve at:lon to return the required loop to operation.

- Qg7-*,*

All reactor coolant pumps and RHR pumps may be deenergized for up to 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months provided (1) no operations are permitted that would cause dilution of

' b t).3. 4 she Reactor Coolant System boron concentration, and (2) core ou:let temperature is maln:alned at least IDOF below saturation temperature.

L eactor coo d't pump s.ll not be star d with one o more of t Reactor L g

oolant Sy m cold I temperatur ess than or al to.33

. unless the _

seconda., water t. perature of ach steam

.nerator i ess than 500F

[

above ach of thVdeactor Coo t System col jeg tempe tures.

(.

MILLSTONE - UNIT 3 3/44-3 1.mendment tio. 7 y

Insert 3.4.1.3 3.4.1.3 Either:',* *

a. With Reactor Trip System Breakers closed, at least two RCS loops shall be OPERABLE and in operation, or

b. ' With Reactor Trip System Breakers open, at least two loops consisting of any combination of RCS loops and residual heat removal (RllR) loops *" shall be OPERABLE, and at least one of these loops shall be in operation. For RCS loop (s) to be OPERABLE, at least one reactor coolant pump (RCP) shall be in operation, insert 3.4.1.3.a

A reactor coolant pump (RCP) shall not be started unless one of the following l

conditions ir met:

At least one RCP is operating.

a.

b. The secondary side water temperature of each steam generator, not isolated from f

the RCS, is less than or equal to the lowest RCS wide range cold leg temperature of the unisolated RCS loops.

c. With a maximum of one RCS loop isolated and with the RilR relief valves isolated from the RCS, the secondary side water temperature of each steam generator, not isolated from the RCS, is less than or equal to 250 F.

d. All RCS wide range cold leg temperatures >275'F and no cold overpressure protection relief valves are in service as follows:

1) COPPS is blocked or tha PORV block valves are closed, and

2) RHR relief valves are isolated from the RCS (3RHS*MV8701C or 3RHS*MV8701 A is closed and 3RHS*MV8702B or 3RilS*MV8702C is clov 9.

" Prior to opening 3RHS*MV8701C and 3RHS*MV8701 A, or 3RHS*MV8702B and 3RilS*MV8702C, all safety injection pumps and all but one centrifugal charging pump shall be incapable ofinjecting into the RCS, Surveillance Requirements 4.4.9.3.4 and 4.4.9.3.5 apply whenever any RHR relief valve is unisolated from the RCS.

REACTOR COOLANT SYSTEM

,ig 3j jj.

COLD SHUTDOWN - LOOPS FILLED LIMITING CONDITION FOR OPERATION 3.4.1.4.1 At least or.e residual heat removal (RHR) loop shall be OPERA *LE and in operation *, and either:

a.

One additional RHR loop shall be OPERABLE **, or

-C b.

The secondary side water level of at least two steam generators shall be greater than 17%.

AoPLICABILITY:

MODE 5 with at least two reactor coolant loops filled ***.

ACTION:

\\

le u 1 % k ee v i d GNMOLL With r.. M thdRHR loo;4)n.m. it'.Jor with less tnan the required a.

steam generator water level, immediately initiate corrective action to return the inoperable RHR loop to OPEPABLE status or restore the required steam generater water level as soon as possible, b.

With no RHR loop in operation, suspend all operations involving a reduction in boren concentration of the Reactor Coolant System and immediately initiate corrective action to return the required RHR j

loop to operation.

l

(,

SURVEltl.ANCE REQUIREMENTS 4.4.1.4.1.1 The secondary side water level of at least two steam generators when required shall be determined to be within limits at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

4.4. M.1.2 At least one RHR loop shall be detemined to be in operation and circulating reactor coolant at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

No q

JhStd khe RHR pump may be deenergized for up to 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months provided:

(1) no rpcrations hi,j,tIA are permitted that would cause dilution of the Reactor Coolant System boron

?-

concentration, and (2) core outlet tem;,erature is maintained at least 10*F

' _s below saturation temperatore.

i

- 0ne RHR loop may be inoperable for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for surveillance testing e

provided the other RHR loop is OPERABLE and in operation.

- - A react coolant u'mp shall eratureslej[t'dwithon et be star r more of e Reactor fWrf Coola. System d leg t than or 1 to 350*. unless 3'g 9.,),1 the econdary ater tem rature of e ch steam erator-is ss than 50*F a ve each the Rea.or Coolant stem col eg terpera) W es.

a

/

MILLSTONE - UNIT 3 3/4 4-5

~

J

insert 3.4.1.4.1.a

b. All RilR loops may be removed from operation during a planned heatup to MODE 4

)

when at least one RCS loop is OPERABLE and in operation and when two additional steam generators are OPERABLE as required by LCO 3.4.1.4.1.b.

Insert 3.4.1.4.1.b

*

a. No reactor coolant pumps (RCP) may be in operation below 160'F unless COPPS is blocked or unless the PORV block valves are closed,

b. An RCP shall not be started unless one of the following conditions is met:

1.

At least one RCP is operating and the lowest RCS wide range cold leg temperature of the unisolated RCS loops is >l60'F.

2.

With two or more Reactor Coolant System (RCS) loops isolated, the first RCP shall not be started unless the secondary side water temperature of each steam generator not isolated from the RCS is less than or equal to the lowest RCS wide range cold leg temperature of the unisolated RCS loops, 3.

With a maximum of one RCS loop isolated, with the RilR relief valves isolated from the RCS, and with the PORVs prov! ding cold overpressure protection, the first RCP shall not be started until the secondary side water temperature of each steam generator not isolated from the RCS is less than or equal to 50 F above the lowest RCS wide range cold leg temperature of the unisolated RCS loeps.

4.

With a maximum of one RCS loop isolated and with any RHR relief valve unisolated from the RCS, the first RCP shall not be started until the secondary side water temperature of each steam generator not isolated from the RCS is less than or equal to 200 F and less than or equal to 50*F above the lowest RCS wide range cold leg temperature of the unisolated RCS loops.

Dec';mb:r 29,1994

-h REACTOR COOLANT SYSTEM COLD SHtJTDOWN - LOOPS NOT FILLED LINITING CONDITION FOR OPERATION 3.4.1.4.2 Two residual heat removal (RHR) loops shall be OPEPABLE* and at.

least one RHR loop shall be in operation.**

l APPLICABILITY: MODE 5 with less than two reactor coolant loops filled.

ACTION:

a.

With less than the above required RHR loops OPERABLE, immediately initiate corrective action to return the required RHR loops to OPERABLE status as soon as possible.

b.

With no RHR loop in operation, suspend all operations involving a l

reduction in boron concentration of the Reactor Coolant System and immediately initiate corrective action to return the required RHR loop to operation.

SURVEILLANCE REQUIREMENTS l

(

4.4.1.4.2.1 The required RHR loops shall be demonstrated OPERABLE pursuant to Specification 4.0.5.

4.4.1.4.2.2 At least one RHR loop shall be determined to be in operation and circulating reactor coolant at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

Insert 3, #, I.i; 1 t

1

0ne RHR loop may be inoperable for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for surveillance testing provided the other RHR loop is OPERABLE and in operation.

- The RHR pump may be deenergized for up to I hour provided:

(1) no opera-g tions are permitted that would cause dilution of the Reactor Ccolant System boron concentration, and (2) core outlet temperature is maintained at least 10*F below saturation temperature.

MILLST0Hi - UNIT 3 3/4 4-6 Amendment No. 77 99 0215

__a

Insert 3.4.1.4.2

a. No reactor coolant pumps (RCP) may be in operation below 160 F unless COPPS is blocked or unless the PORV block valves are closed.

b. An RCP shall not be started unless one of the following conditions is met:

1.

At least one RCP is operating and the lowest RCS wide range cold leg temperature of the unisolated RCS loops is >l60 F.

2.

With two or more Reactor Coolant System (RCS) loops isolated, the first 1

RCP shall not be started unless the secondary side water temperature of each steam generator not isolated from the RCS is less than or equal to the lowest RCS wide range cold leg temperature of the unisolated RCS loops.

3.

With a maximum of one RCS loop isolated, with the RHR relief valves isolated from the RCS, and with the PORVs providing cold overpressure protection, the first RCP shall not be started until the secondary side water temperature of each steam generator not isolated from the RCS is less than or equal to 50'F above the lowest RCS wide range cold leg temperature of the Lnisolated RCS loops.

4.

With a maximum of one RCS loop isolated and with any RHR relief valve unisolated from the RCS, the first RCP shall not be started until the secondary side water temperature of each steam generator not isolated from the RCS is less than or equal to 200'F and less than or equal to 50'F above the lowest RCS wide range cold leg temperature of the unisolated RCS loops.

~~

tdc t.

11, nM BLAC. TOR - COOL ANT SYSTEM ISOLATED LOOP STARTUP-

-LIMITING CONDITION FOR OPERATION

-3.4.1.6 A reactor coolant loop shall remain isolated with power-removed from the associated RCS loop stop valve _ operators until:-

a.

The temperature at the cold leg of the isolated loop is within 5

-20*F of-the highest cold leg temperature of the operating loops, b.

The boron concentration of the isolated loop is greater than or equal to the boron concentration of the operating loops,- or greater 3

than 2600 ppm whichever is lessle-energittel e

c,-

All resch>- culsJ mys cs re c c) t.

The isolated portion of the loop has been drained and is refilled, and e f.

The reactor is subcritical by at least the value required by Specifications 3.1.1.1.2 or 3.1.1.2 for Mode 5 or Specification 3.9.1.1 for Mode 6.

APPLICABILITY: MODES 5 and 6.

ACTION:

With the requirements-of -the above specification not satisfied, do

- a.

not open the isolated loop stop valves.

SURVEILLANCE RE0VIREMENTS 4.4.1.6.1 The isolated loop cold leg temperature shall be determined to be within 20*F of the highest cold leg temperature of the operating loops within 30 minutes prior to opening the cold leg stop valve.

4.4.1.6.2 The reactor shall be determined - to be suberitical by at least

- the value required by Specifications 3.1.1.1.2 or 3.1.1.2 for Mode 5 or Specification 3.9.1.1 for Mode 6 within 30 minutes prior to opening the cold leg stop valve.

4.4.1.6.3 Within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months prior to opening the loop stop valves, the isolated loop shall be determined to:

a.

Be drained and refilled, and b.

Have a boron concentration greater than or equal to the boron concentration of the operating loops, or greater than 2600 ppm whichever is less.

MILI. STONE - UNIT-3 3/4 4-8 Amendment No. JE, E7,60 0020

REACTOR COOLANT SYSTEM

iI 066 3/4.4.9 PRESSURE / TEMPERATURE LIMITS LIMITING CONDITION FOR OPERATION

-&r4r9rl The-Reactee-C : lent Sy: tem (except the -preswelaer) t: peeature-and-

-peessuee-she-11 be 1 Sited in ::cerdan:: with the limR-44nes shown on Figures-

-&-4-2-and 3.4-3-during-heatup;-e:01d wn, crit 4caMtyr-and-inserv4ce-4eak-and-hydrettst k tee + M v4+h-

[n JCeY A-mantmum-heatup-of--100*F ' n-any 1-heue-peefod,.

380'I'd

=L.

A maximum ;ccidown of 100"i in any-1-heer period, and-b c.

A-madmum tempereturi change of less then er egel to 5 i in-eny

--l-hour period-due+ng-+nservice-hydrostat4c--end-4eek-test 4cg-operat4onss

-ebov the hectup :nd :::1down limit curven APPLICABILITY:

At all times.

ACTION:

With any of the above limits exceeded, restore the temperature and/or pressure to within the limit within 30 minutes; perform an engineering evaluation to determine the effects of the out-of-limit condition on the structural integrity of the Reactor Coolant System; determine that the Reactor Coolant System remains acceptable for continued operation or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and reduce the RCS Tavg _and pressure to less than 200*F and 500 psia, respectively, within the following '30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

' w.tkin 71 Aw.s SURVEILLANCE REOUIREMENTS 4.4.9.1.1 TheReactorCoolantSystemtemperatureandpressureshallbe determined to be within the limits at least once per 30 minutes during system heatupjgcooldown,, and inservice leak and hydrostatic testing operations, ad efersGN du re sy th che-l,ove p e re'*J price to wl duws9 4.4.9.1.2 The reactor vessel material irradiation surveillance specimens shall be removed and examined, to determine changes in material properties, as required by 10 CFR Part 50, Appendix H, in accordance with the schedule in Table 4.4-5.

The results of these examinations shall be used to update Figures 3.4-2 and 3.4-3.

lcss veguired MILLSTONE - UNIT 3 3/4 4-33

Insert 3.4.9.1.a 3.4.9.1 The reactor coolant system (except the pressurizer) temperature and pressure shall be limited as follows:

During an RCS heatup, the heatup limits of Figure 3.4 2 apply with the additional a.

restriction that only one reactor coolant pump can be operating when the lowest unisolated RCS loop wide range cold leg temperature is $160 F.

b. During an RCS cooldown, the limits of Figure 3.4-3 apply with the additional restriction that only one reactor coolant pump can be operating when the lowest unisolated RCS loop wide range cold leg temperature is $160 F and no reactor coolant pump may be operated when the lowest unisolated RCS loop wide range cold leg temperature is $120 F.

l I

During steady state conditions, when the maximum temperature increase or decrease c.

l in any one hour period is <10 F and when the plant is not changing temperatures in I

accordance with a heatup or cooldown procedure, only one reactor coolant pump can be operating when the lowest unisolated RCS loop wide range cold leg temperature is

$160 F. The limits of Figure 3.4-2 and 3.4-3 do not apply during steady state 3

conditions.

5

d. During RCS inservice leak and hydrostatic testing operations, the Hydrostatic and Leak Test limit of Figure 3.4-2 apply with the additional restrictions that within a one-hour period prior to exceeding the heatup curve, and during each one-hour period above the heatup curve, a maximum temperature increase or decrease of 5 F in any one-hour period is allowed, o

MATERIAL PROPERTY BASH 3 -2li}

C ROLLING MATERIAL

PLATE METAL CD ER CONTENT

CONSERVATIVELY ASSUMED TO BE 0.10 T%

PHOS HORUS CONTENT

0.010 n7 %

l RTNDT NITIAL 60 7

.RTHDT TER 10 EFPY

1/4T,1227 3/4T,301T CURVE APPL ABLE FOR HEATUP RATES UP TO 607/HR FO THE SERVICE -

4 PERIOD UP T 10 EFPY AND CONTAINS MARGINS OF 307 PC 60 PSIG

- FOR POSSIBLE STRUMENT ERRORS 3000.0 f

i LEAK 7

TEST I

LIMIT \\M

/

~

mg 2000.0

>b

!$m

)-

g

!=

~

U 1000.0 HEA1UP CURVE

/

C TICALITY LIMIT

/

~BA DN INSERV3CE j

HYDR STA'T]C' TEST TEMPE TURE GBS T)

FOR TH:. SERVICE PERIOD U TO 10 EFPY

\\

I 0.0 0.0 200.0 200.0 300.0 400.0 500.0 INDICATES TEMPERATURE (DEG. F) i FIGURE 3.4-2

]

(RUCTOR CCO'.it:T ;YSTE!/ " AT' Fu:/ TATION; - AFFUL4ui J

UF TO 10 EFF

.LSTONE - UNIT 3 3/4 4-34 Amencment No. 60

_ _ ---- u

Millstone 3 Reactor Coolant System i

Heatup Limitations for up to 10 EFPY 2500 i

/

........4..........t......t.....t.....i.....

.....t.... 4... 4.... 4.. I...

i i

... 4......:

..........f....

H,drostatic and...,

3

.....!.......:......... ::I.....

1 i

i Leak, Test Limit...

i i

......l......

....a.....j..........{......

.....L.....:.....

.....a.........a.....

....1....

1 l

....... i :......

}

8 i

2000 i

i i

Heatup At a Maximum of 40'F in Any i.

i

- : -i - -)f.

1-Hour Period Up To 160*F, and - i --

~ ~ t. --

At a Maximum of 80'F in Any 1-Hour

....i..

...................i.....

Period above 160'F.

i

i. fi

........../. :....:..

.......i.....

ns

/:

1500 S

i i

!/ !

....'. Heatu*n Limit i

i as

......../.... :.....:...

l '

3'-

i i

l

/

m g

,r r

QJ 6.

g i

Unacceptable

.../.>.....<..........<.....

i f;.

g t

1000 O erat. ion i

o i

P i

/.i i

i

.c:

... 4......:.........i.....

/

.........:..:...4....

i Criticality Limit i

! ll i

i i ~+ -

-:---. M:

--~ -i- -:--i--

-- - t--: - s -

/

500 g.

i:

..........t.....

.v...>.....<.....)n..

.................g......

.....>..o.4....,...4......

I >:

...n

?

........:....a.....i....

..i...........3....

... 1 o....:....

..4...

..4.....{......:....4.....

..'....?.....:

...a.......4....4.....

....J....4..........6.....

.4....

...4.....J...3....4....

3 1:

0

?

' i:'

i 0

100 200 300 400 indicated Cold Leg Temperature (*F)

FIGURE 3.4-2 MILLSTONE - UNIT 3 0525 3/4 4-34 Amendment No. 60,

TERIAL PROPERTY BASIS -

CD THOLLING MATERIAL PLATE METAL F

CDP R CONTENT

CONSERVATIVELY ASSUME 0 TD BE

.10 WT %

s PHOSP RUS-CONTENT

0.010 WT %

)

RTNDT ITIAL

60 7 RTNDT AF R 10 EFPY

1/4T,1227 3/4T.101*F CURVE ' APPLIC LE FOR COOLDOWN RATES UP TO 1007 R FOR THE SERVICE PER100 UP TO 1 EFPY AND CONTAINS MARG)NS OF

F AND 60 PSIC

-FOR POSSIBLE INS MENT ERRORS

3000.0

}

eV l

i 2000.0 Ab

',) -

i m.

E

\\

t 5

/

1000.0 COOLDOWN' RATES OF/HR)

O Y

t 20 -

300 0

O.O 100.0 200.0 300.0 400.0

.O IfGI M TE3 TEMFE M ii. IRE OEC. P)

FIGURE 3.4-3

)

ME-ACTO' OOOLA?!' SYST !! OOOL;0'.i ' L;M; TAT ON; - AFF,;cA E j TC iG EFF Millstone - Unit 3 3/4 4-35 Amendment No. 60

Millstone 3 Reactor Coolant System Cooldown Limitations for up to 10 EFPY 2500

..........l....................!.....

....l......

.....2...........

1

....,I.....

.....}......

I l

............i.....

....1....

......I.....

......l......i.....

......'......!......I.....

......l...

l 2000 I

l i

l l

i i

l-i

......l>......

.....j!

I I

..............t.....;....

-i i

......I.................

...........1.................

t i

....1......,

I

... CosMown Limits

...g.....

i 8

....}.....}...........,.....

.....g.....

i.....

f f

l I

l l

1300 l

3 1

1 1

1 I

.....i.....,....!....1....

.....1..........1....

.....1.....

....1.......

.....1.....

t I

i I

1 l

....}......

.....g.....

i

...i....

Unacce table

.....:p.....q.....

i y

....q.....

(

i Operat10n

..t.... :...!..........!..........

.... 1......

...../....

)

i j

[

i i

l l

I I

I 1000

,3,,,,

l l

....7..........j......

1....

j i

l

.................1.....l......

....4.....

.....J......

1 1

'....4.....

3 I

t 1

8 1

1 t

1 f

1 t

i 1

i

..............4.....

i i

Cooldown At a Maximum of

.....g.....!......g.

80'F in Anv 1 Hour Pened to

.....g.

e

.>.....q.......

i i

160'F, Then At a Maximum of j

i 1 20'F in Any 1 Hour Period 2

500 A

i i

i i./

Below 160'F

.....g.....j......g.....<......

..... j...... g......g.....

....:....9.....

,.. 5.....

j Lp l

8

?

....3......

.... 5.... ;

, Cooldown At a Maximum of i

80'F in Any 1 Hour Period to

.....;.......q.....

.....{.....,....j..........

160'F, Then At a Maximum of x.

......i............i......

40'F in Any 1 Hour Period Below 160'F i

0 i

i i

0 100 200 300 400 Indicated Cold Leg Temperature (*F)

/

\\

Figure 3.4 3 gt. STONE - UINIT 3 3/4435 Amendment Ao. 60,

v.

March 6. Isso t

_TAntE 4.4-5

'-N'1 REACTOR VE55ft MTERIAL SIRVElll ANCE PROGRM - WilllDRANAl. SCHEDUL 6'M CAPSULE VESSEL IEAD Af/I(OXlD8IC PAMBER

[0 CATION FAC10R WITilDRAWAL TIME ((FPY)

U

'58.5*

3.98(a)

First Refueling (1.3 EFPY actual)

.Y 241'

.3.71 9

V 61*

3.74 16 W

121.5*

4.01 STAND 6Y I

238.5*

4.01 STA!ID6Y -

l Z

301.5*

4.01 STAND 6Y

)

{

a) Plant specific evaluation i

>=

-s n

3

.E i

i

'l j

m 9

wr-a++

-r 4-p f-+

g g

sn-y---

--eviw---wW-w y yr n+=www a---w+w

3-2.1N 7-REACTOR COOLANT $YSTEM PRES $URIZER

'1l D0b t

LIMITING CONDIT104 FOR OPERAT]ON 3.4.9.2 The pressurizer temperature shall be limited to:

a.

A saxtzum heatup of 100'F in any 1-hour period, b.

A maximum cooldown of 200'T in any 1 hout period, and c.

A saximum spray water temperature differential of 320'F.

APPLICABILITY: At all times.

ACTION:

g., ;g L,g With the pressurizer temperature limits in ex :ess of any of the above limits, 1

l restore the temperature to within the limits within 30 minutest perform an engineering evaluation to detemine the effee Ls of the out-of-limit condition on the structural integrity of the pressurf 2pri detemine that the pressurizer remains acceptable for continued cperation 6r be in at least HDT STANDBY within a

the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and reduce the pressuriter pressure to less than 500 psia within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

$URVElttANCE REQUIREMENTS 4.4.9.2 The pressurizer temperatures shall be determined to be within the limits at least once per 30 minutes during systee heatup or cooldown.

The spray water temperature differential shall be detemined to be within the limit at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months during auxiliary spray operation.

G MILLSTONE - UNIT 3 3/4 4-37

July 10,1997 i

REACTOR COOLANT SYSTEN OVERPRES$URE PROTECTION SYSTEMS LIMITING CONDITION FOR OPERATION CW 3.4.9.'3 11m40verpressure Protection Sy:t : shall be OPERABLE with eith:r : Or b b:us a nsti a o+ one et,re6 ul cm. s pony)* and no fdeby lire,rm f.v.,p t

Rese ts. nels,t Syste e t, a se/ o ne o f IAe /* Ils w n *,y p,-dJu n ca ya y a + tsje s t.~y as to ne Tw; relief v:h::, :: fol hu:: r e l

(P LI ' hJ :

Ins e r f I-Te W ir-DPerited Filiif VilVi; (IORYQ*ithlifteittinsi 3, y, t 3,9 ehic de

--t :=:d the limit =t:bif h d in Figur: 3.1 - S er a

?

Tiger: ?. '

'b,

:ppr:pri:te, er t

2.

Two residual heat removal (RHR) suction relief valves with setpoints 2 426.8 psig tnd i 453.2 psig, or l

a r..,, s l ca r wsle f 3.

One PORV with411ft settingt within the linit

p = ifieddin Figure 3.4-4a or Figure 3.4-4b/;; :pprepri te and one RHR suction relief valve with a setpoint 1426.8 psig and s 453.2 psigry o r-

)f. f. M:

A c3 hat:r C :hnt Sy; tem (RCS)4depressurized with an RCS vent of.h A

v.eeter th= ;r ;;uel 1: 5.4 square inches.

4 re ea r <c e.s 6 0 C*P'*"

APPLICABILITY:

MODE *7 when 'th: i;;pr:ture ;f any RCS cold (egAir hn ihn er 0;11 tc 250*r u d CDE '; MODE 5, and MODE 6 when the head is luce?

on the reactor vessel.

3,o i.3. 6 l

l

' ACTION:

ti,eespierj l

c t.

With one eMe required relief valvet inoperable in MODEJ-er 4, restore tw relief valvet to OPERABLE status within 7 days,or depressurize and vent the RCS through at least a 5.4 square inch vent within the next 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

(Aerewirel d If.

With one cf tuc required rel' ef valvet inoperable in MODE 5 OR 6, tither (!} restore h relief valvet to OPERABLE status within l

24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , or.(2) =mphte depressuriz! tin =d i:nting ;ef % hours.

f the RCS thr ugh t k n+ R,.4 square inch,ved within,; tete.

5 os) esisdLA an ec:s ven:yj/ th:frequired relief valves inoperableft es u e,cc s e 4.

With b:t' j

depressurizbi:r =d v= ting the RCS %r::;F :t -kut - 5.4 square 2

inchg e e within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , a.IestU.AonAc4ven[6 2

d.

utih th; nts vented per AC4

N:

, b cr c, verify the vent-pewer

t hut== per 31 d:y: uh= %e-pathway i: provided by :

g,t v h:(:), th:t i; h: Led, :=hd cr-;therwin :=ured ir th Op;r-3, f. 't. 3. C F3 0 0 YO E

I

?

a,,,/ y,1 4 n o d C.s <,,r a y,9 p,.e i.,, 4 s MILLSTONE - UNIT 3 3/4 4-38 Mendment No. JF 77, pp.143 04st

~

Insert 3.4.9.3.a

1. One power operated relief valve (PORV) with a nominal lift setting established in Figure 3.4-4a and one PORV with a nominal lifl setting established in Figure 3.4 4b, or Insert 3.4.9.3.b a With two or more centrifugal charging pumps capable ofinjecting into the RCS, immediately initiate action to establish that a maximum of one centrifugal charging i

pump is capable ofinjecting into the RCS.

l

. b. With any Safety injection pump capable ofinjecting into the RCS, immediately initiate action to establish that no Safety injection pumps are capable ofinjecting into the RCS.

Insert 3.4.9.3.c

.Two centrifugal charging pumps may be capable ofinjecting into the RCS for less than one hour, during pump swap operations. Ilowever, at no time will two charging pumps be simultaneously out of pull to lock during pump swap operations.

When an RilR suction relief valve or a PORV, which is armed for COPPS, is unisolated from the RCS, and when the RCS cold leg temperature is greater than 275'F, a maximum of one centrifugal charging pump and no safety injection pumps shall be capable ofinjecting into the RCS.

p h

1/19/97 b E

,A JTOR COSLAVT TY.etrw

- Et*FE55Utr PROTrtT10v sysTN LIMIT!E C0c!T20N FOR OPERATION ACTf 0W (Centinued) f,,ef In t event the PORYs, the FRR suction relitf valves, or the R;S tre used to citigate an RCS pressure transient, t Speciti ven-Reper shall be prepared and sub:itted to the Co=ission pursuant to Specification 6.5.2 within 30 dtys.

The report shall describe the circu: stances initiating the transient _1 e effect of the PORYs, the RHR suction relief valves, or RCS vent corrective action necessary to prevent rec;; rence.'n the transient, and any Entry into an OPEPAT10NAL MDDE is per:itted while subject to these ACTION requirements.

SURYE!LLAN E REQUIREMENTS h n3 r r e 7A., r ea cs,-ep 4.4.9.3.1 -Eich "DRV-sht11 bi de;.gre.rrd Nxv.3 trdtfA0PEFABLE by:

Perferr.tnce of an ARALO3 CHANNEL OPEFATIOKAL TEST on the PORY t.

tetuttien chtnnel, but excluding valve operation, within 31 days prior to entering a condition in which the PORY is required OPEFAELE and at least once per 31 days thereafter when the PORY is required OPEFABLE b.

Perfer:Ence of a CRANNEL CALIBPATION on the PORY actuation channel at least once each REFUELING INTERVAL; an cold l

WcK o Iile tcRV urtrawe restatta,, syster.,

Yerifying the PORYA4schtin nive is cpenAtt ietst once per (com),s ero.,ral c.

72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months when the PDF.Y is being used for overpressure protection,

w..s tnh L r caes res.<cJ 4.4.S.3.2 EtehtRHR suction relief valve sht4-be dimenstnted=0 PEP

-the ?= avet4:r, reiie f vthes-ere-being-esed-fee-eek-everpressvee ABLE then-e5 foiices O ofu ^ C4T. 4,y ;

prettet4cn-

%,.,f o., ac aolar;

. tves bet. e e a fle ACS a,.l each er p.<..,,/

q e e cpen Iw.gPER suction relief valve 25!*RYS70?'.. by-ver4fy4r,9/.at least t.

once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months that O'WS*MGM1A-and-BRHS*MV6MM-ace-opent und b.

Fer FM au:t4ct, reHei v:1v 3RH6*RV6 h.

v, u...,..m.t-4eeet-ence per-44L-hour 4-that-3RHDM&lL025-and-3RH&2MV6MM-ace--epeni-and-

6. 4:

Testing pursuant to Specification 4.0'.5.

InJ erI

't, i. D. 2 L.-y MILT. STONE - UNIT 3 1

ac 3/4 4-39 A:centsent No. 77, pp 797 133 1

I

1

)

i Insert 4.4.9.3.2 4.4.9.3.3 When complying with 3.4.9.3.4, verify that the RCS is vented through a vent pathway 25.4 square inches at least once per 31 days for a passive vent path and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months for unlocked open vent valves.

4.4.9.3.4 Verify that no Safety [njection pumps are capable ofinjecting into the RCS at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

4.4.9.3.5 Verify that a maximum of one centrifugal charging pump is capable ofinjecting into the RCS at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

+

_,.__3,,_.,.,,

_y_.-_._.

e krch 11,typ) a gdc6T 1H G.h '3t t' Psi 9T fot'<vC Neve 800

/

.~3

\\

i j

700 l

l

\\

f s

)

i f

800 I

t

/

)

,/

i

/

500

/

t i

s

[

's 50 100 200 300 s,,

400 RTD - ADOTDNEERED LDW MEASURI.D RCS TEMPERATURE F200RE 3.4-4o NDMINAL MAXIMUM ALLOWABLE PORY

/

SETPDIRT TDR THE COLD DYERPRESSURE SYSTEM

-s

_.,,,,_ (OUR LOOP OPERATION)

J

]

't 9

-g

.LSTONE - UNIT 3 3/444o Amendment No. 60

High Setpoint PORV Curve p te e m For the Cold OverpressureASystem

(

800 3

1 4

>.....j......t,.....,....<l.....<.

..........g....

....}......p.....p.....g..

.....<....4.....

....g....

..o I

1 t

t 8

4

....g.....g.... g....

....7....

..... p... o

....t"

- - - ?'***

- * ' " + ' +.

t t

t l

3 1

t

...4....

... 4.. o 4.... a.. o..'...

.. 4.... 4..... J., u u s........J.....

..o 4.

t g

l 1

1 1

.......... ~....

...4...

.....,...4.....,....e......

.......4....4....g....

...4.....

.....~.....a...3...

i 8

l i

i.

1

?

700 i

.........1.....i.....i,....

.....:,.....i,,....i......i....

i g

g

....y....<i....3......!.....

i 662 psia

....i.... i...

I

....g....

p p

.........1......i....

i i

i

.....i

..................o

..f

.u...

.....I......*..4....'......1 1

1 l

.g....4......t.....4......

..4......!...

i 3... 4...

4..

4

...4.....a...

.... 1.. o.1.... p..

.....I....!....::.... :....

I t

i p

p

.. <..... <..... <..... <......... q..... t.. o n o.o

.o.

oopnu

o. n.p. m. >.. o t

1 t

/

1 l

t 60.g w

.....t............!.o.!....

.........t.1.....

i.

t

.........1.....

................n n..

1 g

. u d... 4.. o.

..t....i....!...

.........4.....L.u.d....

..to...l.........1....

.o t

...b.......

4 1

?

...p.....p. u..b..4....

"4....

...<..o<.....O...(...

...b...b.n.p....b...

.l.....b....b.4..

e

>....p..

.u

...., o. c..

?

e o p..o!..

1

.o..

...p......u

... g.. o. g..... p... 9.. o

.g n o.g o. u. oo.p o n G

n.

500 m,

2

. 4... 4.. o e..

418 Ps1a

......s....i....

o. 4.. o. i... u.

..:no

..b...io...J..

2

...c.....<....g..

p...

()

...<.....g....4.......

...;......3......i.....,....

4...j o

,!... g:.,

.u..p.....g.....p..

s

.. g o n. p o..g o ogon

.un.

.g no. goo.t"'"

(.

..p.

u 7..

..........,g........

.... t. " " * ! ' " " : "

w 400 i

y

....$......b...>..oa.n

.0..

i. <...

... 4 n.. 4. o.. p.... b...

...l.........

.4...

O

...p.....t......p....y.

s. s

.. t.

..p....

..p..

...t""+puop""f""

" f "*o p u o t o Mun Q,

..p.......

?

...o..

g.....

7

...~o

.........o.

u..p....

p..o.

.'....'..'....b.

.4..

1 25

..o..

.a,..4....a..

...o....

J

E 300 O

.o....p...p....p

.s

...<.....p....<...

.n.p....t.....t.....p..

....p....p.....p....p...

-2

'..ne......1'....

?

.........;......,o..

........g.....g.....

...g......

...og..

m o o.. o u...

.n.......

.o.

.....4...4...

n...

......1..... w.... d. o.

.. n;..... u d....

+.

n i. o.. 4..... 4..... ;..

200 p..........p....p.

............u..

7........;.

. ng o o..go ogo

....g.....p.n..... g u n o o g

.o.

2,

.i i

.....b....b..4...4....

....... o:. o. 4....

.4

...o u..J.....i.....b....:n..

ob....i...

o 4

......in....n.....)......).

s..

.o

.n<o..e....<....(....

..t....t......t......q...

...In.olo...

2

3...")o..

100

~

1 p..

........o......

.o....

n.

..........o

.g..

.....'...4n..

l

.4.......4....

.........i......,

...4....

. o.b.

1...

4 I

. o p....p...b..4.....

... p...

.n.s....<.....<<n..<..o

...qi....

..p....p....g.....

..... p.

4.... o ).. o 4 n

'.. o?

.. o p..

.t o. n.p u..

o

.o

..s wu s,

e.

.o <o n.ong.o

.ngon.gon. gun.un

. u t.o

.?

p."

u.

.o 0

(

50 100 150 200 250 300 AUCTIONEERED LOW MEASURED RCS TEMPERATURE ('F)

FIGURE 3.4-4a N

MILLSTONE - UNIT 3 0535 3/4440 Amendment No 60.

ItJ5MT l6W St. T %WI WRd NcVG

,,,.....-~.

_i 800

.f

/

\\

/

s

\\

~

700 s

lE s

-/

!g g.

\\.

/

i

\\

600 g

s

. )

500 N

t x

E E

\\\\

/

\\

50 100 200 30 400 RTD - AUCTONEERED LOW MEASURED RCS TEMPER TURE ('F)

FIGURE 3.4-4b NOMINAL MAXIMUM ALLOWABLE PORY SETP0lNT FOR THE COLD DVERPRESSURE SYSTEM (THREE LOOP OPERATION) 1, WILLSTONE - UNIT 3 3/4 4-41 AmendmentNo.-(0 l

Low Setpoint PORV Curve.

N tec nm For the Cold OverpressureASystem i

t B00

,....;.....g....

g.... 7.... i.....i....

.... j....,

..... )i..... ;,.....

....;i.............................;,....,....j...

7

.....t.....

p....

.u.p....

..u.

u..

....g....................

.u..

i i

.....l....................

.....J.....

i i

i I

i i

i

.u.....4....

.. n o... 4. u..:.... 4........4....4....4.....g.....

.... <.....g.....

....i.....

.....g...4...,4...,4,,,

i i

700

.....i......i..........i.........i.....i......i......i. u...

i i

1 i..... g u...

i....i.,,,i, n i

g

..op....p.........g.....

..o

......u..

.o........

p t

.....I......i......i....

.....i......i...........1.....

.........a....

..............,1....4.....

4

.....,..........!...............!......i......!......!......

.... t..... t.... 1..... !......

....i.......

599 psla

........1....

I i

1 1

600 g

.....i.....

......,.....I......!

.. 1..... 1. u...:....

u...4....

. ~.......

....u.,

.t 4

2 i

1 1

1 i

i 1

1 1

..u.4....8..........l....

... 4.. u 4.....;.....g

..4........4....4....

....j....4..u......

.....,.....g.....g....

..o 1

....q....!s u...8 2

l 1

8 i

i 1

1 p..4....

...4uu4...4....4..

a....4....<.....g..o.

.... <...ng.....q.....

...up.....p.....p..4...,

1 e%

I i

1 1

1 3

I

.u..p o...t.n

......p.....pu...p...

....... y......u.-.

.... ;... u....

....p.o.

.......p....p....

..n un l

1 1

3 1

1 1

I c,,

500 i

1 I

i t

1 I

........t.....e.....<...

.a.....a.........a.....

... a..... a..... a..... a.....

41h O$i$

.....k.

v..

i..

.. h..... b.... f......

... f.. u.. j.. u. i.

.. i.....

.... i......l..... f,....f....

1 r

...,u..1.....p...p....

1 i

t t

4

..p....p...pu..go

...p u

...qu ugu.u nopo

. upou

.""t"*"t""

^

1' 1

l i

. u. p.... gun p.

.u.

.n

.u.. u.. ;...

4.

,m 400 e + o. 0... 4....

...4.u.4.u.4......4....

... 4....d... 4.o.4.u.

1

....d....de...d...4...

.. e

..t++..

.o

..u f

~

i i

.i

.... '..u..

p.....I

.....p.....p. up.

p..

...p....q....

..ui

...p....

.u.

i....

p...

.u.

A 1

1 u..

.u............

p..

....p....

un.,

i u d.... d. u. 4....

. 4... u a... 4..... a...

.......J.....

.u......l....n

....d...

.. u u.u...

uu.....

o i

1 l

1 1

l lE 300

..1p u.1 1

I t

1 O

.....!....t.o 1

u.upu..!pu...

p p

.....,u puu s....

. u p u.. <... u p.... p u..

....p..u p.o.p.u.p u.

p...p...

l 2

l 3'

1 l

l

..p....g......I....1'

9... o p....!..

...u..,.o.

.. n o u..p.u.;

u.

u.p......o.p...

p...

..p

..t.......

u...'.....r.....o..

.........1...u*..o :

u........!..........'...

.n i.

u lu. d. u... i u

.' n o 6.. o. 4... 4....

..t....4.....4......i....

.. i. u a..,. 4...

4

.. d.. o l

l 1

l 4

1 1

200 1

. -...i u

+

....}..........p...

.....u..

.. ;. o p.... p u.....

"..*.*t.'""'""t.""*t""

"l""?"*

1 4

...d..... u.4.....e...

.4..........*.....a..

J.. o J.......... i..

.tu..o...u....3........o

.u.

un.......

1 4

4 1

3

.4........4...4....

. 4..... u. 4. u..lu

. u.u. o 0..... o 4...

.u. 4... o 4. u. 4.n u l u..

. n un.

undnuduu a

I 4

3 1

1 1

......p... 1.... 1 i

r...'...4...

a

...o...........o.....n...

.1 i

1 2

i.

1 3

1 1

i 1

1 i

i.

... u" u.

.o...u.......u.u...

.u...u.

..o un

+

.o.

1 1

i 1

1 1

1

.........4.o.J.u...... 4.... 4. u. 4. u. 4... u

.u.. n o s.... 6.... i.u o

.u.< u..lu.o un.: on

.uu n..

. u 4.. u t

1 I

u np u "f "o

..... p.. o.p.. q..

4... n

. u..b.. 4. u...... +........<....0.....<..u.po.

.. o.. u.

..o on.

2

....,....,.....,.....p...

o

..........,..... >........p....e.............

...e....t,....

.. u.,p..i...

4 50 100 150 200 250 300

[

AUCTIONEERED LOW MEASURED RCS TEMPERATURE ('F)

FIGURE 3,4 4b MILLSTONE - UNIT 3 S

Amendment No. 60, om 3/4 4-41

EMER6ENCY CORE COOLING SYSTEMS 3/4.5.3 ECCS SUBSYSTEMS - T,yg LESS THAN 350*F DAN 01306 LIMITING CONDITION FOR OPERATION 3.5.3 As a minimum, one ECCS subsystem comprised of the following shall be OPERABLE:

, a.

One OPERABLE centrifugal charging pump b.

One OPERABLE RHR heat exchanger, c.

One OPERABLE RHR pump, d.

One OPERABLE containment recirculation heat exchanger, One OPERABLE containment recirculation pum;2, NJt A4' Joe e.

and p(:st, w,6 ms mo.l e esity me n t' ol" V< lvus is cysile o f r h, lh N (J s f.

An OPERABLE flow path ;gdi; efAtaking suction from the refueling water storage tank rn? 5 9 g = u:11; r::?!gr.:d and transferring 3

suction to the containment sump during the recirculation phase of operation.

APPLICABILITY:

MODE 4.

ACTION:

)

With no ECCS subsystem OPERABLE because of the inoperability of a.

the centrifugal charging pump, the containment recirculation pump, the containment recirculation heat exchanger, the flow path from the refueling water storage tank, or the flow path capable of taking suction from the containment sump, restore at least one ECCS sub-system to OPERABLE status within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or be in COLD SHUT 00WN within the next 20 hours2.314815e-4 days 0.00556 hours 3.306878e-5 weeks 7.61e-6 months , b.

With no ECCS subsystem OPERABLE because of the inoperability of either the residual heat removal heat exchanger or RHR pump, restore at least one ECCS subsystem to OPERABLE status or maintain the Reac-terCoolantSystemT"[9 1ess than 350*F by use of alternate heat removal methods.

c.

In the event the ECCS is actuated and injects water into the Reactor Coolant System, a Special Report shall be prepared and submitted to the Commission pursuant to Specification 6.9.2 within 90 days describ-ing the circumstances of the actuation and the total accumulated actuation cycles to date.

The current value of the usage factor for each affected Safety Injection nozzle shall be provided in this Special Report whenever-its value exceeds 0.70.

r "Ae imum of o centrifugal arging pump d one Safety Ipfection p mp all be OP BLE whenever e temperatur of one or mor of the RC cold

]

egs is 1 s than or equ to 350*F.

MILLSTONE - UNIT 3 3/4 5-7

4

EMERGENCY CORE COOLING SYSTEMS dA[l$1 j$C6

){*

$URVE!LL.ANCE REQUIREMENTS

=

4.5.3.1 requirements of specification 4.5.2 p (The ECCS subsystem shall be demonstra k.5 2 All ch ging pum nd Safety actionhmps,excep re ired OPE LE pumps, psall be dem strated operable b verifyin that-the abov IO l

t motor ci cult break s are secu d in the pen positi at leas once. peri

~

hours w never the emperature f one or te of the CS cold gs is less han or ual to 3.5p F.

m

\\vitsthe excep h'n th t valm mxy u out of a t:y8 m en t L., / cy<)//e.

o f bei,y w n w Il re a lipe y

(

9 4

0 l

(

MILLSTONE - UNIT 3 3

3/4 5-8

_ _ ^ _ _

3-zlcn Jur.a 13. Isso l

)

afsttiv1TV tovtoot tyritus satts mooreatoe TtwatuTutt tortrf tfrNT (Continued)

These corrections involved:

1 a conversion of the MDC used in the TSAR safety analyses to its equival(en)t MTC moderator density erith temperature at RATED based on the rate of change of THEA M POWER conditions, and (2) subtracting from this value the laroest differences in MTC observed between ICL all rods withdrawn. MTED THILM POWER conditions, ant those most adverse conditions of moderator temperature and pressure, rod insertion, axial power skewing, and zenon concentration that can occur in nortal operation and lead to a significantly sort negative ICL MTC at MTED THERKAL POWER.

These corrections transfomed the MDC value used in the F3AR safety analyses into the limiting End of Cycle Life (E0L)ive MTC value at a core condition of 300 MTC value.

The 300 ppm surveillance limit MTC value represents a conservat opm equilibrium boron concentration and is obtained by making corrections for burnup and soluble boron to the linIting ICL MTC value.

The-Surveillance Requirtiehts for measurament of the MTC at the beginning l

and near the end of the fuel cycle are adequate to confirm that the Mit remains within its limits since this coefficient changes slowly ove principally to the reduction in RCs boron concentration associatec with fuel burnup.

,)

3/4.1.1.4 w1NfwuM Trw*tuTuot too co1Titst1TY This specification ensures that the reactor will not be sace critical cith the Reactor Coolant System average temperature less inan !!).

This limitation 'is tvguired to ensure:

(1) the moderator temperature coefficient is within it analyzed temperature range.

(2 the )P 12the trip instrumentation is eithin its norr.a1 operating range.

setpoint. (4) the pressurizer is capable (3)f being in an Op!RAILE status with a interlock is above its o

steam bubble, and (5)the reactor vessel is above its tinmum RT temperature.

g.

1/4.1.1 $00T10NsysTtws The loren injection System ensurts that ne available during each mode of facility operation.gative reactivity control is The components required to perfom this function include:

1)beratedwatersources. 2 charging pumps.

power supply from OP:RAILE(4) ber(ic acid transfer pu=ps, an(d )(5)

(3) separate flow paths.

diesel generators.

[pg. Me s i. A..- Q i

With the mt: cc-Me tre-ete-- hec 2001 a minimum of two boron injection flev~ paths are requires to casure single functional casability in the event an assumed failure renders one of the flow paths inoperable.

The boration capbility of either flow path is sufficient to provide a IHl/TDOWN MILLSTQKI - UNIT 3 3 3/4 1 2 Amen:nent No. [. 50**

V.ay 17,1995 RUCTIVITY COWTROL SYSTUS b

1A.5ES BOP.aTION SYST MS (Continued)

MRGIN from expected opertting conditions Figure 3.1-5 af ter xenon decay and cooldown to 200'F.quivalent to that required by The maximum boration capkbility conservatively (sinimumbound expected operating ) conditions th boration volume requirement is established to life.

full power or hot zero power condition (peak xenon.The initial RCS boron con concentration assumes that the.most reactive control ) rod is not inserte The fint.1 RCS boren the core.

This set of conditions requires a sinimus usable volute of 21,802 gallons of 6600 ppe borated water from the boric acid storage tanks or 1,166,000 gallons of 2700 ppa borated water from the refueling water storage tank (D 1

A sinimum RVST volume of

, insed consisten with ECCS requirement.

1,166,000 st11ons is specified to be

'B W.t.2 9 +

rhol in N bf3 J da N )

Wi th t he * * * ' --- -" "-*

b:'. :: 200*h A one Boron Injection System is acceptable without single failure consideration on the basis of the stable reactivity condition of the reactor and the additional restrictions g,7 single Boron Injection Systec becomes inoperable. prohibiting CORE ALTEF DJ/blaM m 3,

,.o,..,.

.u_,,_,.....v, y

==, =wa;; r:..="sn.

...... u, n: nnme e, e ;;L W l'.LZM..'. 5"fyyQ'. "J.".',DN$: x::bm. _ _QQ@

t.. erp

(

2.7 " 3.'. J.. ',.{."':'_' C.. Z.. :. :.Z..,. Z..Z. C..i ?. '..: ::G...:..!A, G....: U.:~:: !q._z:":

y

,. yy

.. ~.

The boron capability required below 200*F is sufficient to provide a SHUTDOWN PARGIN of 1.3r. Ak/k after xenon decav 140*F. This condition requires either and cooldown from 200'F to E600 ppm borated water from the beric acid storage tanks orusabie volume of 4100 gallo a

250,000 gallons of 2700 ppm borated water from the RWST.

storage tank is 1300 gallons.

The unusable volume in each bo-ic acid The contained water volune limits include allowance for water not available because of discharge line location and other physical characteristics.

also ensure a pH value of between 7.0 and 7.5 for the solution recirc witliin containment after a LOCA.

iodihe and minimizes the effect *of chloride and caustic stress corrosion onThis pH archanical systems and components.

The minimum RWST solution temperature for MODES 5 and 6 is based on analysis assumptions in addition to freeze protection considerations.

Ginimur/ maximum RWST solution temperatures for M3 DES 1, 2, 3 and 4 are based on The analysis assumptions.

l t

MILLSTONE - UNIT 3 5 3/4 1-3 Amendment JE, Ff. F7, 115 0313

l l

l Insert 133/4.1.2.a With the plant in h10DE 4, one lloron injection flowpath is acceptable without single failure consideration for emergency boration requirements on the basis of the stable reactivity condition of the reactor, the emergency power supply requirement for the l

OPEllA131.E charging pump, and the fact that the plant is administratively borated to at least hiODE 5 requirements prior to cooldown to hiODE 4. Also, the primary grade water addition path to the charging pumps is surveilled to be locked closed to prevent a direct dilution accident in hiODE 4.

Insert 133/4.1.2.b The limitation for a maximum of one centrifugal charging pump to be OPERAllLE, when cold overpressure protection is in service, provides assurance that a mass addition pressure transient can be relieved by operation of a single PORV or RilR suction relief valve.

]

OcccmDer 2&,1994 3/4.4 traCTOP C00LaA*T SYSTEM EASES L

3/4.4.1 REACTOR CODL ANT LOOPS AND CODL ANT C1Ptut AT10N The plant is designed to operate in MDDES 1 and 2 with three or four reactor coolant loops in operation and maintain DNBR greater than the design limit during all nornal operations and anticipated transients. With less than the required reactor coolant loops in operation this specification requires that the plant be in at least HDT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

In HDDE 3 three reactor coolant loopc, and in Hode 4, two reactor coolant loops provide sufficient heat removal capability for removino core -

decay heat even in the event of a bank withdrawal accidentI however.fa single '"mp 3 l

reactor coolant loop provides sufficient heat removal capacity if a bank l

withdrawal accident etn be prevented, i.e., by opening the Reactor Trip System breakers.

ik a LvK w lkolewtl actis /ent en fs poenTe,/,

in MDDE 4, A tad in n;D: ; vith nector cocient loop i;11d, a single reactor coolant loop or RHR loop provides sufficient heat removal capabili,ty for removing decay heat but single fa;1ure considerations require that at 3 'P least two loops (e4%+r RHR or RCS) be OPERAB'.E.

P '/ M A9 tpy co-ame w or 53 In HDDE 5 with rsacter coolant loops not filled, a single RHR loop provides sufficient heat removal capability for removing decay heat; but single failure considerations, and the unavailability of the steam genertters as a bett removing component, require that at 1tast two RHR loops be OPERABLE.

The operation of one reactor coolant pump (RCP) or one RHR pump provides adequate flow to ensure mixing, prevent stratification ano produce gradual InJer reactivity changes during boren concentration redcetions in the Reactor g,w, Coolant System.

The reactivity change rate associated with boren reduction t7ill, therefore, be within the capability of operator recognition and control.

The restrictions on starting an RCP vith :r: er ere "3 ::1d hs: ha th:r er eqd t: M0*T are provided to prevent RCS pressure transients, caused by energy additions from the Secondary Coolant System, which could exceed the limits of Appendix G to 10 crR Part 50/

The RCS will be protected against overpressure transients and will no exceed the limits of Appendix C b-either; U) r:ctri:th; _W wa4 v ht: tr the pre ner4m Athewby-tmh; --- :

cche n for t-h a rp N cet!?nt te-expPd ht e,

- (2Y by reitricting starting of th RCP to wh:n-the secondary water temperature of each steam generator

.s les-. hen 00'i chave esch of the RCS cold leg ttmperatures.

ad y/e qc.

b ut.l vj en wb;ch cul,/ cwe pa nu. e e a ca n 4, s u.Wn m xM oe

.Sy1 Ytan v Ut h wu hl & cerl 14 e

</t J.j a, p ee awe of tic.SyJie m,

{IjlST0HE - UNIT 3 8 3/4 4-1 Ament. ment No. J', f9, 99 I

insert B3/4.4.1.a in MODE 5, with reactor coolant loops fille 6, a single RilR loop provides sufficient heat removal capability for removing decay hest; but single failure considerations. require that at least avo RilR loops or at least one RilR loop and two steam generators be OPERABLE.

4

)

i 4

4 4

Insert B3/4.4.1.b in MODE 5, during a planned heatup to MODE 4 with all RllR loops removed from

- operation, an RCS loop, OPERABLE and in operation, meets the requirement's of an l

OPERABLE and operating RilR loop to circulate reactor coolant. During the heatup there is no requirement for heat removal capability so the OPERABLE and operating RCS loop meets all of the required functions for the heatup condition. Since failure of the RCS loop, which is OPERABLE and operating, could also cause the associated steam l

generator to be inoperable, the associated steam generator cannot be used as one of the steam generators used to meet the requirement of LCO 3.4.1.4.1.b.

4

+<

1

March 11, 1991 p/4.4 REACTOR COOLANT SYSTEM BASES ' Continued)

, (:

The requirement to maintain the isolated loop stop valves shut with power removed ensures tha.t no reactivity addition to the core could occur due to the startup of an isoleted loop.

Verification of the boron concentration in an idle loop prior to openirip the stop valves provides a reassurance of the 4

adetJacy of the boron concentration in the isolated loop.

The 2600 ppm is suf ficient-to bound shuttiown margin requirements and provide for boron con:entr6 tion measurement uncertainty between the loo) and the RWST.

Draining

)

and.rurilling the isolated loop within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months prior to opening its stop vahes ensures adeq'JrTe airing of the coolant in this loop and prevents any resetivity effects ate te boron concentration stri.tifications.

i 4

fh StrY B 3)f.+, I 4 l

(

\\

@(( STONE-UNIT 3 B 3/4 4-la Amendment'No 7,60 l

1

)

Insert 113/4.4.1 The requirement to have all reactor coolant pumps de energized, prior to unisolating a loop, insures that the heat from the secondary side of the steam generator, in the loop being unisolated, does not result in an energy addition transient during the return of the loop to senice.

w--.

REACTOR COOLANT $YSTEM

(,,f BASE 5 SPECIFIC ACTIVITY (Continued)

Bas *ed upon the above considerations for excluding certain radionuclides from the sample analysis, the allowable time of 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months between sample taking and completing the initial anS1ysis is based upon a typical time necessary to perform the sampling, transport the sample, and perform the analysis of about 90 minutes.

After 90 ginutes, the gross count should be made in a reproducible geometry of sample and counter having reproGucible beta or gamma self shielding properties.

The counter should be reset to a reproducible efficiency versus It is not necessary to identify specific nuclides. The radiochemical energy.

determination of nuclides should be based on multiple counting of the sample within typical counting basis following sampling of less than 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , about 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , about 1 day, about I week, and about 1 month.

Reducing T to less than 500'F p* vents the release of activity should asteamgenerat$f9tube rupture since the saturation pressure of the reactor coolant is below the lif t. pressure of the atmospheric steam relief valves.

The surveillance Requirements provide adequate assurance that excessive specific activity levels in the reactor coolant will be detected in sufficient time to take corrective action.

A reduction in frequency of isotopic analyses following power changes may be permissible if justified by the data obtained.

F,,,,.. m e r.,.... -. - - s s-i v n s

,S m.

.~~; ;;;; p ;;;,,; >;--~~

~~7n,,3n;,,, ;,,,,7,g,,,,;~~

~~pr.:;77 ;;~~

t,'^ :10.:er:...'. ' t. ;e e-d.ve.t e:S,,.e-r ?:;t ;p;;,),n; '.._ lis^ i s i nc e th,,e

~~,n, a i i h f,.,,,~~

~~j t....~~

g'NDT '* '

4***

"~

"-r'-'--',

~~... _ _ l~~

~~_z_..~~

~~,t.~~

~~r ^ ^ 1..l F,.~~

~~__e.wi,.J L,i~~

~~.h. 10 0?* 50 rul;.~~

~~Hv....i, thi ;;t1dian ;;rr; ^h;an in~~

~~.....,,.......r.... t.,. u....,3.~~

~~_2 1,~~

Th; r ^;:ter **mina+ temeerature and pret tuce and =y:t;- $;';t;; =md :::ideen rktii ' Ith th; em;^;tfe ef pestaurif er) skall ha 14w<t96 in area-d*-**

~~+ka~~

~~u. n. s. t..y.-..-.s n o w'~~
,m
-- ' u'...,>i y...ww' v>-
= - -
= >
+-
., a ivi ey.weaiss vn.
.vu.
ns,.
,t,_
___t i.-.wsu
. v4 v c a se b w e e s wt yI.a s w s. 3 4 lw b.wg. f 5 4V 5 5 -.. e y. w t i TE gws k-0Ch^71$Cr; ^h;n,"0 7;k;". me,
~~w. syn mis~~
$b. 4 v
Wil b 91 Ene llfalk 14m.. akawn.
it.<+
i<n.. en,canigau..:t;; g;;..,7, t;,,;; p;;;:-t-d E *y b; ^ ' it i ^.; d $} I".t'rp el tt #? ' * ? ^d -
u
....i; 7-^"^**,*a ^,
iv.s.,
494 anu a.m a esI1ng Ilmite by ri*at w.
n:n-d;;t'i. i.'t.r; :.-ly.
ra* aa--e' ape-at'ea, ae-**$=***+
Cht *S:t** T!ti;;, ;.,,, p2";
t Odditi;n Onj ps;;;U T:;7 h;;ttr L
o MILLSTONE - UNIT 3 B 3/4 4-7
\\
3-21 97 March 6, 1990 1
1 REACTOR CODlh'T SYSTEM Batts
-)
i 4
t e-a..u.,,,usess,,mr n,us.,.e i, r.. >. a.
1
..n.......,
1 se.e.
.k.
..a c
..a
. t... m.,.......... -
..m
...ss.m.
.t..
m.
.. m.......m....._.............................. _...... _. -
1 5
e
,u.
,,.a.
,.....u
.......... u.. u.... a...,. a a... n.,,.... n..... u....a.
=,,,
7.
i
~~beW, r....o.-~~
4
.~.
a.
.,.. u...
........ u....
. 61. m v. n.. 6 v r u y.........,. a......,........
n.
.. r...,.,.,,..o.......r.......
v.
.n.
u..__-
r..e s s u r u e r,n e s.s w y a.nd d o own r.a.u s.s n u.. n n, o......
__2 s
.....m
=.;.
r.,o j..
,,.., f n,,,.,._.
....r......i.
.n.
. y..,
n...
m.
..y.......
u.m ___.. o.
.n.
u
.....o.......... u..
.a m...
. r..,....
u vi su...u.
m.m
.w-r, snu t
t.v e t em ere s e rd en h.vr'e^' * * + '
~~ad 'a~~
~~uiea n.. u n w~~
....a.....
'a*'.' ' ',"d.'.'.'++ a'.'."...',
.~ ~
y...wm.w e5
... ~. u o u r
.n
..,.,..........n
.n.
n.n n
..;......s......_.
m o.,
......a
~~
- -. -.......m.u..,
........e.. u...
,..>..,i.... _>..-
~~u. m.___.....~~
m.....,....
.. _,.._ a,,...........,
.....n
~~a... u.,..., L.........,........ s....._., a u.~~
m...
...... - _..... m.w..nas w m.
~~o. ___,.., m..~~
. m.
u u n. m. a.
u
,,..s....
.o.
~~r... _. _.. _....~~
.n
. v
.y.. o n... 6n.n....
.t u.... car.e.. a. 2 a.
a,
......m l
$.$... a..
..J
.. 3 M.Ls.
3d..
^ - - - - ^ ^--
-9.-$A J
^A
%.I JA'"*8.1..
u..._,,-m..,,,.,.. =__',".i'.**..*i.*..*.*.8**8***w ***DW bi3Em '" w '.
.. - - -.,- - r
'"1 n..
m.u..a n. gg,
1.
,, m,...,.. n........,......
, e e.m..
.. cre s u,... v
~~i.....o,..-~~
4
.t
,....~.
..... 6..s....,34.<,.... n..
.o.
~~3., s.. 7... 2~~
~~1..~~
46.
....t or,,,
at.
,i,........,._.i._<.u.
2,
., N D T ".'..*. * ' a"" 'm" ".8" " " ' ".' '... a... _,.,
-c..
.m.o
.n n.
'-' ~ - ' ' ' ' ' "
z u... -.....,,.
o n.
3
........n
.on
.....n.
m. N D T " " " ' " ' " " " ' ' ' ' ' " ' " * * ' " "'-
Pa.,,.. e.. s...
.,..u.._..u
.u...._.i..a...n..........m
.,,..s....
wi,6
.r...
..3,.,,... L.s.,.
r.2.
.._3....-...-...
m..
1 4
.-6....r+..
u...
3 k.m.
L..
... a..
s...
a..
. t.. a..
~~1.. a.., a.-~~
. --.. 41.,. - -............._
a w.
..,.,..,.i...
V..
.,.. e
. nvi..-
~~6. m..,~~
nv.n
,n...
-.b.pT,, u.a...., u..
~~a...~~
.n...
.n o...
... n...........
..n,.
..a_.
u Tx.. _.,. _ _
..s.
5.
sen..
....ou,
.n s.u.o__.....n,. u.. w.. u... a.,.
.....r..,o
.i.,.......
.a n.r,....u.....
s.
... u
... n a a.
.u.
AL-2 r.......
F1.....
,..l...
B t,F A,, A, t
1
.J
~~n. @. L.,...,.. J AMT~~
...A L
h--
A
.u..n.,..--.. v u.,...,o
.n.
i
~~n. sa.~~
.. u _. s, u.....,... V' n.
..a
.e u,.
~~e. y.s..,.~~
eu
.i.,.,......i'...u....u....
y
_..g
.i...
--.- ~
g, a........ -.
c
.a
.__2,.._,,
...,._ _ <,,,.. u
.v., h, DT..
. w r...
3, 2 7...... >,......
o
.n
..o 2.
m.33.,
... r e.o v......,.-.
y........
4....
,u.
. o n..,i v.
.u..
.u...
.m n
.m.
s..
,........u.
~~u...u...~~
7........
........m...
..a
.~...
4
~~c...... u.. n m g~~
4.
i MILLSTONE UNIT 3 B 3/4 4 8 b endment No. 48 r
i 1
2 4
..e
~. -. _ - ~. -,..
,..-..m m.--.e..-.-.v.~
.-.mv.g.s.
.--.-,~.,,.-e.--v_.,m-
y R.
A JAN 31 1986 TA8tE 8 3/4.4-1 2
5 REACTOR VESSEL FRACTURE TOUGMNESS PROPutTIES e
'm 50 ft-1b A
Shelf Cu P
T 35 Mil RT fede E
Component Code No.
Grade '
M
(*F)
~~F (ft-1b)~~
{ft-1b)
M Closure Need Dome 89812-1 A5338, CLI 0.08 0.010
-40 60 0
96.0 w
Closure Mead Torus 89813-1 A5338, CL1 0.11 0.010
-40 70 10 107.5 0.010 30
<30 30 121.C
(
Closure Nead Flange 3-1 A508. CL2 1
Vessel Flange 89
-1 A508, CL2 0.15 0.007
-40
-40 116.5 Inlet Nozzle 89806-3 508 CL2 0.09 0.009' 1
<70 10 162.0 Inlet Nozzle 89806-4 A
,CL2 0.09 0.009 0
<60 0
158.0 Inlet Nozzle RS-3 A508, L2 0.07 9.
-10
<50
-10 130.0 Inlet Nozzle RS-4 A508, C..
0.08 0.
O
<60 0
136.0 Outlet Nozzle R6-1 A508, CL2 0.012
-40
<20
-40 128.0 1
Outlet Nozzle R6-2 A508. CL2 0.006
-30
<30
-30 127.0
' 30 121.0 006
-30
<30 Outlet Nozzle 89807-1 A508. CL2 4
Outlet Nozzle 89007-2 A508, CL 0.
-30
<30
-30 125.0 Nozzle Shell 89804-1 A53, CL1 0.05 0.012 100 40 85.5 Nozzle Shell 89804-2 38, CLI 0.08 0.010
-40 90 20 104.5 Nozzle Shell 89004-3 338 CL1 0.05 0.009
-50 0
103.5 Inter. Shell
~
-1 A5338, CL1 0.05 0.010
-40 120 60 92.5 113.5 Inter.:Shell 2
A5338, CL1 0.05 0.014
-60 70 10 90.0 129.0 Inter. Shell 89005-3 A5338, CL1 0.04 0.009
-40 50 106.5 136.5 Lower Shell 89820-1 A5338, CL1 0.07 0.006
-50 70 10
.5 124.5 Lower Shell 89820-2 A5338 CLI 0.06 0.008
-30 100 40 114.5 tower Shell 89820-3 A5338, CL1 0.05 0.007
-30 80 20 79.
124.0 Bottom Ne Toros 89816-1 A5338 CL1 0.13 0.012
-50 20
-40 91.5 j
Botton.ad Dome 89817-1 A5338, CL1 0.15 0.012
-30
<30
-30 161.0 4
Int
& Lower Shell G1.59 SAW 0.07-0.011
-50
-50 200 I
& Girth Weld Seams e
NDTES:
Q.o i
HMWD = normal to major working direction NWD = major working direction
y
~l r,
E l
1
(
-u E
g E3 k
We R
E ~G W
WM l
a t
_ 'y?E
. u. 3 A 4$
4
- ein uS t j
)
g e.2 g
8 W W
gn
.m e
E g
N
=
m
~~!1111 1 1 I l~~
!!ll I l I I I
!!lI I l f f
lt119 i l 1 E
I N
R R
9 2
(gWD/N) 3DN3m.d NNAMN j
,_ MILLSTONE - UNIT 3 8 3/4 4 10
l nms r
l U Lbi5 hUnbinw.h-3 f._ _. _ 'u... e.,, a s @..
... a s. s, =- _ _ _.e'.i I ",d
',T ^.0 "..'*.a env be unad untH tha
.eniee
.' ^s's u.. vI E I.
e
...,'n.'.,
.wwi=.
.-a.
n
.u
- - - _. 2 1 4,
2
_ - - - - - - - - *,..- i.w
.n e.....
i ft.
sw.6 r i v s s en i.
"-=
-..,,(****A*:'*
2
-__-2 ou
.ici nr.
v.muv.w in uwwviv.nw. wi.n on.
i.. *-
,.,...a...
..y
.i..
gW.Y 1,99......., a... m a,. 6
~~a--~~
..a in een a -. a. n.a r1If.-. 4 A F]
l ll5
.m.e.rw rt e r _ve --.-
-. L m.. a. T.a. *..-
g
.g
,7__,_,,
-,.i i ns i s.v' i sw v.*-.-_4.
.s. :
it.
.a m 1- - - -
._ _ ni.,,k.a _. d
_. -f. a.
A. vu- *..
r
..n
.o.
,.,-e.
i
~~A.u~~
~~.L.~~
~~1.., m. '.. Ar 4ha_, a,m e n_i_ a_~~
J-k.4u...
.L.
4.(...Lt..
.. - - _ _a s.a i, i,.-... e. A. L......A.m T..L.. f,, A L..
..h&Nd..hI.
.a.m.m sw
~~n. v.g w.~~
_...n 9 h_3 4_ E Ahl_ k l.
J.
Q ( b.7 A...
d....e S
.a_
i.
_m_
1
.J X
A JJ_A s.
,3..***.^
}
p..,m,.
...v set.._n_w._
.g___
......,.y.w.pr.n., b.n ws,v..w_. u.u.
.. a.
~~u...~~
~~m. e.. t...a,,, m. i.~~
ya.vew.
m.,
.....n.
..i vi
_.i..__i.
a,-._
q e,
L.
.n.
~~v. m t......~~
...p 2_. _
ww=
-- g,,,,g,nq g g u.-
..s.wi..
.t,_
j n... = r = " * * *d i i m. ".......i.-..,. Je
.k.
, m i, i ti... s4 in
.u
..Y.
J.1._
t...a.
~~...-. k...~~
._.. i s u.
AB 4L.
i,......
..r-
2+
j
.i,.s..t.
~~s.... k._.... t. o.i s...~~
.-.a...- a. w i.*. i vi s. n y. -....
n.
y.
i
..n.*6,wy,.nv i______i.__
.-.. i s _ _ a, _ _
t e,..a,.
i m....,.=
r'***wiu 65'V5.: =*wiu
'*****w"*" ira. e v.i varivw.
_... s i.
."0-
.. T r,om Appt ng i A i
G. ^.,.;^.,-.
e,... e rs,s,a l.c u.l a s, s. a v a l ng st T,nDQ s s c r i v s g u
..v
.a w
..2 j
o.
.nw r e...wi. ^
8 u.o t....,..,,....
m.
.. i.i
^
.J - 3..a g i.
gn._
m.n.
.A..
A. t._..- m6n'pos are eiscusssu in
_J g,.._
A.
, r. m t
~~7..,...J. A,..,..__.wi.m,.~~
l n.
n
... n
~~2...~~
r..
en A..
...,_...i.J..<,,,
1 4.,, (......
.J
.. L... J
~~s..._,. i.,. n. i..s~~
~~.. L......~~
in mi._......
.-.. a.,.. L.re. m.
....7 i
i ot.
,..,.2 u.o o...
^
-.m_
_m g..a u...
...u...
.u
..... L.
,u.....r.2.......,..,
v.
2._... e...
iiiy...,
.._....1.,_,o.
____t..
. s. = m.. -
.m
^o-
--u wi+h.,.n+h Ar an..en.s.
.e
~~n....-. "i i~~
.nism.n.e.,
3, suw a i.ny.n h.
11 3,
g$ (hg n9 e",
e t it9st.k ( h h e.
~~a. l.~~
~~.... #. J ". ^ #. A.L.-......... '.~~
A.It.
^^
^^
. m.
u..
n yw..., :,6wie.,.:- ny
...,s. i u n.i.. -..n.. v i mi.n..o.n s vi..sni.
se,. :,, e, v,i_,...,e oc,i. n.yviisw ig i
.n.
-a,
...r.;^
,,,r,
.m.,y
(-
u nc m m n us s.a su...
vi
.wi
. y.m i. n...,..
..,6 c
4... o....
.....,u.._.
~~u...~~
.vii.n.
3 4 <,. _ _..u
.. r..,u
~~a....~~
m.
.,~.
4.m..
,n..._,_ s.o..s...a
~~s. o... u..~~
r.
u..
-1. u v.
... e r s. n.. o.
~~s..~~
..n...
.n a.
...u..,s.a.,..u.s...
s.
..t_1.........-
C,.,,, y
..s......
_,....., 21._
u.
4._<...
~~m.. 4 m.., s..~~
s.
._._.o.
. ~...
~~u.... -... -~~
.n u.u w.._.....
. a.,.o...n
...,...... n u.u 1.,.
..w
.n
> m om 3 4.,...,.,, i. n_.. e.
~~u.....1... - o w w u s u,~~
"' NDT..'
o.
,e, inwiv,ws.,6ns reaisuu.n inuusso snin, amgT,,,,.6orrespa,nwiny 6v.-
,a_.
.o.
.ni
,,,a,,,ia..
,n_
_r........ u..aL...
..o w.-
n....., - - - _ _.....
.r. L. a wr men a_.,k sa
.._1,_mi +a...a..... m.....a. '.. d '..u'. *. -
~~s..-. v.,. %~~
u
..a...
. - a
.....,o...,t......._.
_._ >ui, i. u - e,
- u.
.v.o u...
.n un L..._....,.
.n
.n..
.......r......
r..
a.u.....
~~m. _~~
2.._,.,.
L.._..m,.,
u.
.. n,
...u y,........... sin w 6nsi u s
.n.
y
-__....._ 4,a....,o
-..,2...,...... _
................ u.u.
._.., A-,
t.
., u..
i u s i. n u. -
.n
.u.
am v,,
.gu uure n 1,nn u m.
^ 1 R ' ' * * ' ' ' " ' " ' ' - " '- '
T, L
. u
.n.
^
^
V A.,4...e 4.
'w ine o m. w u, W
A.,.. J f. -
m k..
-a...-.. n a. k.....
en u.
w...-
2.._. m...
t.,.,. t...c.
ivu
..n u
B3/44-11 Mendment No. #, 89, M.lu.5 TONE - tlNIT 3
('
m k
3 -2 l.4 7 i
March 6. 1990 i
RfatTOR COOLWT tYtitM t
4 4
Ratti i
2 W......>,eusren,nse
,,u,,e d,
,*..u....n p..........,
.n.g
......, en m univ.,
,2 i
~~u..m~~
-(S m
. n m.__
og i u _
.u.
......o....rvas m en...
,..u...
.,u.
o....
n,'
j artil i.._ _ _._ _.... _ _.. _ 2 t
u m i.www. u.. m.
Thus,. u,_.....,... _......,u
..c...
.u.
..........,....r...___
~~u.... u...........y'~~
. L.
....n.....-
j
-.... g. v....... y 5....
...u 1
y-.-
1 l
e
. v i
- m ' an a gg
'iS l
1 m_-_
ug.
a.
~~n.... u..,.......~~
1
.a t..
-...u...,.... _.
j W6H,
.r.......,
4 i
.i r-n..+..
".m..m- "..,. 1 1"...
a
."...>.',e u. u -.... *.,,..a s.. a.
4%
4
)
~~u. a............u. u 2~~
~~n... e.. a.~~
La
.. - - ~ >--
.,....,-._.u_.
.2.... u.. u...,.,..,, u.u.
._,.,o....o.......... m e _.
a.
~~s.... s.~~
...a....me.
._ 1,.e..,
..a
,.....,... k,va m...,4, and 1..t 4..,
I i
>>_. s..m. n.., a.. 6...,, a a e a nide t ra n s i e n t. r.. t e d_.+.
8. ' '., -
m
.n u.........
....y........
u.,,
a n d n v a r ra nc e t,.., u....
m.. u.... ;,...ractu. *a".a...
."...--.... u
.,L..... _... o. _ %_,.
.m.
fa_. at -
^ ---*.._
.I a...--.-.
__....u.. o,
~~u.. m_._.m~~
,_ _,___. _ _ _... _...u....
u_
....r.... 2>,,.......<..
..m
.u
..a
.u..
.u.
.t..-.<.
t -
....r......
..,......, og
.n.
._2
.. u..
w w...y w...,.....u......,u....
. vm w....un.i o..u..yrvs.wrv......,. 3, n 3.,....,..
......~.
sne f. u.._.
..a
.a.
m
.a_
i teet96W
.._u...,..
a.u
.. u.. m. u.. u....
.,n_.1.w..,_.a.u............._............._........
7.......,......
.. ~.. ~.... >...........,...,....
4
.2.,.. u...
. u...........
m a_
n...
-...........m..
o.
....i. u. i. i..
Ph..,...
.n.
L.
...ia.<-
.L-
...... i.iIma 1 neat {An mi eL. #1 d*
- r s o.... _ _ _ u.. i _'m u
.. o......,...L.u-_._.__u..,._
m.u m. u -
u._.
...... a
. u
.1,..
me.................r........u...-
1
..t.
.....t
.2
.6
..u..,.....
4
............u.v.i.....--. u....--
_t
.....r...._._.._a.-..
m>
m a.
....~ >..-....
.,.,.......r..
. _ -. -.t.-..=....
m
.__.2. _.....-,-.....u.--...
- u
. u u.........
e.
..... u.
. a.
..... o...
BT.,hvua...b.
2 O.....
u..
W.bH
.......e l
.p.
, u.........u...
~~u.. n..~~
...u...._
4._
i-
.u........a...,.
u.o
..u....i.
.....sm..
.i.....
...7.
...e
.g..-
y -
~~n.. 6...e~~
......-.........a.--
.-...o....a mr........
3 2.
..a.... - - -.. '.
i-.
.n. u. m m.. u...u.
.r..,n,. a ri.w u_.
~~a..,...~~
~~,.. a... _.~~
~~u...~~
...n....
a u. u..., _ _. s.. -. -
..e u,,.....- _ _ m
~~u. o._~~
i.i.. u.a,....,.. u.~
. u v o...._ i.
.u.. u....-
...a.... u.. n.-
.. a e.. L.
,, y,.. _ _,,. n. go,, n...
.., m
.._am,..
..y
., m...
g
..a_
.u....n. u..u.
u..
, u....
~~u....a..a.~~
. ms. a.
.....r,.
6 v.r v,
-=
m m
a.
int g
.s......
2
..m....,
o... o....
i.au.vu....
..r..
.u.
n i
loe* Hon 3g.. -
i k
J l
n MILLSTONE. UNIT 3 8 3/4 4 12 Amendment NO. 48 4
i I
4 l
l
---.--.-+w.-w.r--..+.-.
t-'
M-
--w- - -
wm-i---.+-elg-.
~~---He wi-us-me-e-w--w.*w+-+--T---i-~~
J - g 1 qg
(
Harch 6. 1990
~~ltfACTOR CO$lANT tYiffM BAlft ly.._.......,r........~~
..u...
_a w
u.... s......n,
.t m m, m iv. o u -.
i m_-
m s s.......J..-
mem, eL.. am.
-g
-- e t a n d_.u......
....J.-
F-L.
m, v.,
sun..
__.. a n....._ s
.......s..o. um._a
.nw.....-
r-n e,. %.. ' >.' v " _'_. '
u='.="...
..i..-
n-u w
...n
. u s 6, tus
.n gg %.
.L u....
9.
y,....... J.. =. l..,a
.A.
~~1..J.. o--. 1. %. %~~
... n
.o.
'k
. k..i. ieb.L.
........v.,
t.
1.
(.. J.g..... _
w s.,.
,.whr.
yi b.bs ML.
.L
.. 2 --.
--LA L--
a.
..a vs..== -=.. ".".._1..
.c 2
.A
....1
=
v'u.o.===.'==..='=..f '. i, s n...6 s yn e.
s i i vu s w.. = = =_...."..= '. - i..
"*ww=w e.u.....
.a
. u-4 i -
,,s.--.v a ve s,
,,t,.......
tu n.
o as
.._...,r........
....o v. s o y. y.
.. s. i n...
~~a.. u.o~~
-. a..
sa...
2 a
..f s.
.....J
......n....
v u,
EurYs visa.eu s
n...,... i s ia s i u n g 5 6ppewvun r aiap e ins was v
.u
.-.L9.-
..J meno..,
1u.......,i 6ns,6 vwp D 3 H g ur at.

~~s.-~~

at.
r ' v e ' ' '"4. 2 4.
='"-*' =- '-
-r-'=='-"
'J 8 6 * "' '".
...,2-_
y._.-.
.n
{
t!! int'
~~m...............,...i..,...~~
~~...._,.. u.. u.. n. _,.,,...~~
~~s... s s. r..... - - _,n~~
.. #_ _ e _
se
~~d.., A_nna tm eka en_n_i s...~~
~~1.,o. 4..,, -~~
kr--.-='--'----
='*
'*M=is==
=.
nto
...a1 44...L8..
_ _ _ e. -...
.n i. i..v L t.
~
r---- '-
-a.,warmen..
me...n.*
-*---J
..J"
==55.wr*w a.A..-
.... s. a s.. I at w.
.. #4mHa- --- - -------'.r...=J'8.8..
L... o..
=.
.. i. a... g.. t.
a.
,- i n t.,
- s'
.t
..i g.g...
f..
y... s.
.... s b,.
w n..
. _,i analGR O.T IDE Yt35ti 6..u asil.
ins u.......u.....,,....../...
~~u. 7~~
,s u...n..
__.s.
1
.r......
r...e
.u.-.u..2.
~~u... m. n..~~
i 2..I.S.
..a_ sf, t e a_ d h
... e e_s..
Y. t..... *
( m f a. e.
.......i..i.
e k,., n e s...
.v
. L... r..._.
...-.e... -.... k.a u
qn inn A_ _..,.
~~t.... e L. e s...~~
. L......,n, e..... o
. i 1
u.o.
u.o
,.~.... p s...,.
~~u... u..,u,.~~
.,am..-...-
...lA.. _
....,..n,.
...r..
.o n
(.
......,. t........
.m
.m..
ns asms
.n.
.no
...........u...__.,..6vv......su..,,...,.
u,
,,, a u.m
...n
.... r..
........L......,...,,,,
g
___6 a.
...,.,,._o,s......,..
o.:
u...1L...
m
....,.....u
... n i v.
usser o6s+.-
.........a_
i
.t......... r....
m.
.u. o m._
...._.._,..._,,,,ny
, n,,m. 6...a.
.... nv6 4
.n 3..h une conomons no longer ruer,zos,a ;.,,. k.a M. -'_d! 9 "*.
i eveve+ a.
n.
...,....... a_. a....,n.
, <,.,,, o. s..a. x.....foM r
..n...... v.
~~g. a...~~
t....
... i o...,..,.
..a..
...$...i 6
sitetlys...
~~o. n z--i...~~
..-.,,nu
..m._.
~~n......~~
..e
.f..
i
.g.-.,
t.... _ a... _
u.u
....m....,.I..-
w wg
_w.....
r... - -.-
..m.
i..,..,.
e.6s6: ens unus neuve rus.
3 p..
......e
,4.._.
._ c a...
L.... o,.
_.i.o, e.,.
~~u...... s..,,s. u....~~
o_ r a, n u un. t.%.n t on_
m__*+..e....-
..- e.
a_ v, k
~~u....,,o i.n.t.~~
J.--
."a.';
..,........r..,.
. n...o...-.
.m. u.....o,,,n
,,,,,n z v s u s,i u-t.
a ~
..... n.
i nn w.
tt
~~u. o _.~~
.v..wisn6.
- w i 4..
i a__.
t.
o m..u,...._......,.
.___n._,
- nsu
..vy v,. n n.
.o...
m
..u4.
..,..s..
._.. _.rs tenn is..m nawr n.isn e
- enu i.nv.-
o........
.._ s,.. ;...,..
~~t.. u.~~
ins.w r,...n..
s n s m.,u tt
.....n.
.o t......
u.
no
......r
.s
~~u....~~
.--'...-.'..".....,,..a...'..-
6..,.
i
_ -.......... n.a o
..v,.
u.o
~~u.... g i..~~
.. M 444-4 r
-- -i
.... e. 3..a
....t.....

~~r' u.~~

nu i
h_, n_u e wi t h i ne n a '. =.., '..**"-.. *. -
',' s v's '. '. m'
. unw y
n
...t..
..t....... -..yi..
t..
t....
.n.w. m s.wu vs ns iusw vn en inw v.n..
m..
i C
i i.
i e
i
!{
1 i
MILLSTONE UNIT 3 83/4413
' Amendment No. 48 i
~~' A s~~
4 d
4
~. _ _ _. _
i
,3 3-21 97 February 8, 1997 PTSCR 3 4 97 i
REACTOR C00LANT SYSTEN Asts 1;
,,,,,,,...m..
r m a._.-_-
_..~.r.t L, rt i. a._
i p yn 6 g...ggry s
u..s.,..,..,.u..........,.................___
--,_ ma m.
,s._,.........i.7_.......
8t4a 8
_,....r..='*..u...."','*,.,,',_"',...','.m...... _.. e
..o...
m.
N'
.........m.,
n,. _. a... m.,.,n.,...
u...>..............
m_a..
m
.4..........,..,..,..,u.......
u.....
,e.. u,..'
".....r. '._'__ _-r.....
r-'.'.'
..,5W U.45e 1_.
h............,........-u.,-_..o..._-,_.....__>.....
l
.o
.. v.
.n.
..u.,.,,_,.__...u...
a..
..................u..
I I
~~n...~~
. m w m,,.. u...
im..........
.u.
nec m a_.,..
_.. _.. __.. m...
,,,.s u...
o o n........
o..
y
,._....,,u.......... 3.u..
...s,
.m.._
t u...,...... a u....u.,.,,
.u._........
..... _._.....u..,..,....u.,-
u,
a..
. m._ u...... _
m
...._u.
~,
u
.m.
.u
_,..t.
......,. v....wr...
m
.m
.....a a.,. 3 m.,.
F,..,,..,
u...._
m.
. m... t. m.
_.... a..&.._.. r.
2...
...,.u
...,2....
o_.,u.
m.
, -......,_o.a.,
u.
.a...,...... _.. _.
>a y....__
s... u........ m....
................ u....-
...r..........
u.-.. u. m..,.. u..._.. m_. n_..
r....__- _-
+
ms<,s,s...........,.........
4.
.u
. u.. <,.
......a.a u__...............,..,.. n,.u m,'. -7z-'-'....'.,-.-m.....,. r.....,, _ n..
o
..,...u.,s...,...
P.J,.._..,
. u... m.
ha. r f, Pea & d ( fi
.m A P r* m Pr4. n e_.. u4.h. x.._ A.t h.,. f....
~~e..., m 4... m. m m~~
~~g..~~
~~g.,,, m m..........~~
......,m.
m v inrntaavnt r n v i u s.
i. mi
,o
..........e n...
_.., ors......
.....,....,,,m_.3.,.,.,
..m
....m
.. u.- oa wu e,,,
.. on,y...
...m...n.....m...
.o
~
n --...............,.
.m.
non o.
s,.,,_,._..,..m..
m.u.1 v 5.. _. _._.
A.
m W
. us<,s,mnia_...._.a +x.._ i.....i os, pro m u,on o wm v..,...-e
.- -,<. +,
.-.ee.
mx._...._._.....
,.6._
5.
v ere..s.a,.......
7..r.*6""..
_......s..........
........s.
__..en.e.
..m_.~.
~~u.....-~~
...o
__m_m m
..m.
0,..m_____........,...,....
....,J.._r....,..,..r........,....,-.._......
6 O. u
.L..
.k.
1 _,
......J
..L.
.,......u....,.,.
.L-n......
.m
~~m....~~
.... i..,
r.
~~1... u.... _ _. 6..~~
. e.........gu. er., __,_m....,_m..__.
__ m. u_
-..i
.....n y
_ u m............
.s...
.a...
..,....o...
.,y
....u m
,u r. u.....,....
.,, m..,m,
. m.n n.on m u o. n... o u.
~~u...,,y..~~
~~u...,.~~
.m.....
_ _ _ m.,m..
m
...._-. ee u.. u..... u.-,
u v.- v.
..w m....-..m....
u oue
.m,. u.
._.u...,
m.
m.,
. u... m mu,. u,. eue _....u.
r-----
--s V.... _ _. _,
o..
. T.,.u..
~....
._,m
.m.r..,.... _. u
, -..... _.,....,.._....n..
,..... m o y e _.. r. nee,s a, a....n,._...
m.
....u..
... um
....,.. mu u..
2.
.a.
m u...a
,. r......m.
.___ m..m...
,....o....,.v...............,....m.u.. u..,.. m.
..m,_-_
veu.u.n..u.
<n
,.m.
.... m.
n
,...... ~...u.... u,_ m,..
..,.......u..~,_
......_,..,,m,. m...gu
.m.
.m
.r....1,..............
....m.~.,...,.,..m m
mm
...m
....m..
,.n,
....am.,...
or..
3.m.
....2.
u _,
_m
.._m n..,. u... u..
............m u..,
_..a.u,.,..,....
.... a u.+ u,,+
., m. _
...._a
___m
.. m,._.
. n u s,.-
__..m....
... u. m..,...,.,.m...
,m.,
.n.
-~
.t.............,...mm..,m.v.,_.
v
.m....u,
. _ ~.
.1,. -.
.~.... -
~....,.............
.u_ _._ u. -.
.o r..r
.m.
o.-
m
~~m...~~
. ~
,.,~.
...o
.......... 3 u.4
,.y.
.o~.,..,
M1u.5 TONE - UNIT 3 0474 8 3/4 4-14 Amendment No. JJ. 57.
~,-
February 8, 1997 PTSCR 3-4-97 REACTOR COOLANT SYSTEM BASE 5 O
installed and disallow start of an RCP if secondary temperature is more than 50',above primary temperature.
The requirement to lockout safety injection pumps and all but one centrifugal charging pum an be met by ensuring that a single failure or single action will no esult in an injection into the RCS his can be accomplished by empi ying at least two independent men o prevent a mass injection. The le actions which meet the requirement)following are five examples of acce I
for locking out a pump:
~~1) rac g down the pump breaker, 2) i i~~
placing the' pump in PTL and pulling its UC f s, 3) placing the pump in PTL and closi'ng the pump discharge valve (s) to e injection line, 4) closing the pump, discharge valve (s) to the injectio ine and either removing the power frofn the valve operator (s) or lockin anual valve (s) closed, and 5) clo ing the valve (s) from the injection s ce and either removing the power om the valve operator (s) or locking m al valve (s) closed.
A safety injection pump t is rendered incapable of in ing into the RCS by at least two indepe nt means may be energized for esting or for filling the Accumulators.
The Maximum All w PORY Setpoint for the CDP 11 be updated based on the results of examinations of reactor vess aterial irradiation surveillance specimens performed as required by 1 R Part 50, Appendix P and in accordance with the schedule in Table 4.4-5.
3/4.4.10 STRUCTURAL INTEGRITY The inservice inspection and testing programs for ASME Code Class 1, 2 and 3 components ensure that the structural integrity and operational readiness of these components will be maintained at an acceptable level throughout the life of the plant. These programs are in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR 50.55a(g).
l Components of the Reactor Coolant System were desigr.ed to provide access to permit inservice inspections in accordance with Section XI of the ASME Boiler and Pressure Vessel Code, 80 Edition and Addenda through Winter.
l 3/4.4.11 REACTOR COOLANT SYSTEM VENTS Reactor Coolant System vents are 3rovided to exhaust noncondensible gases and/or steam from the Reactor Loolant System that could inhibit natural circulation core cooling. The OPERABILITY of least one Reactor Coolant System vent path from the reactor vessel head and the pressurizer
(
N MILLSTONE - UNIT 3 5 3/4 4-15 Amendment No. pp, pp 138
f.A/S E R f 3. 4. 9 REACTOR COOLANT SYSTEM BASES 3/4.4.9 PRESSURl'/ TEMPERATURE LIMITS REACTOR COOLANT SYSTEM (EXCEPT THE PRESSURIZER)
BACKGROUND All components of the RCS are designed to withstand effects of cyclic loads due to system pressure and temperature changes. 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.
The LCO and Figures 3.4-2 and 3.4-3 contain P/T limit curves for heatup, cooldown, inservice leak and hydrostatic (ISLH) testing, and data for the maximum rate of change of reactor coolant temperature.
Each P/T limit curve defines an acceptable region for normal operation. The usual use of the curves is operational requirements during heatup or cooldown maneuv&.ing, when pressure and temperature indications are monitored and compared to the applicable curve to determine that operation is within the allowable region. A heatup or cooldown is defined as a temperature increase or decrease of greater than or equal to 10'F in any one hour period.
This definition of heatup and cooldown is based upon the ASME definition of isothermal conditions described in ASME,Section XI, Appendix E.
Steady state thermal conditions exist when temperature increases or decreases are <10'F in any one hour period and when the plant is not performing a planned heatup or cooldown in accordance with a procedure. During steady state thermal conditions, the limits of the heatup and cooldown curves do not apply.
Cold overpressure protection is adequate to protect the reactor coolant system.
I N
MILLSTONE - UNIT 3
-B-B/44-Amendment No. 17. JP 0528
Z hl5 E R T 3. % 9 REACTOR COOLANT SYSTEM BASES PRESSURE / TEMPERATURE LIMITS (continued)
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, and the LC0 limits apply mainly to the vessel. The limits do not apply to the Pressurizer, which has different design characteristics and operating functions which are addressed by LCO 3.4.9.2, " Pressurizer".
10 CFR 50, Appendix G (Ref.1), requires the establishment of P/T limits for specific 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 neutron embrittlement effect on the material toughness is reflected by increasing the nil ductility reference temperature (RTum) as exposure to neutron fluence increases.
The actual shift in the RT m of the vessel material will be established u
periodically by removing and evaluating the irradiated reactor vessel material specimens, in accordance with ASTM E 185 (Ref. 3) and Ap)endix H of 10 CFR 50 (Ref. 4). The operating P/T limit curves will be adjustec, as necessary, based on the evaluation findings and the recommendations of Regulatory Guide 1.99 (Ref. 5).
The P/T limit curves are composite curves established by superimposing limits derived from stress analyses of those portions of the reactor vessel and head that are the most restrictive. At any specific pressure, temperature, and temperature rate of change, one location within the reactor vessel will dictate the most restrictive limit. Across the span of the P/T limit curves, different locations may be more restrictive, and thus, the curves are composites of the most restrictive regions.
The heatup curve represents a different set of restrictions than the cooldown curve because the directions of the thermal gradients through the vessel wall are reversed.
The thermal gradient reversal alters the location of the tensile stress between the outer and inner walls.
The P/T limits include uncertainty margins to ensure that the calculated limits are not inadvertently exceeded. These margins include gauge and system loop uncertainties, elevation differences, containment pressure conditions and system pressure drops between the beltline region of the vessel and the pressure gauge or relief valve location.
In an effort to minimize the system frictional losses, additional restrictions on RCP operation below 160*F are provided in the LCO. These restrictions result in increased acceptable system pressures enabling a
greater operator flexibility during heatup and cooldown in MODE 5.
MILLSTONE - UNIT 3 B 3/4 4-6 Amendment No.
0528
l 1 AISEW 3, s 9 REACTOR.C00LANT SYSTEM BASES PRESSURE / TEMPERATURE LIMITS (continued)
The criticality limit curve includes the Reference 1 requirement that it be 140*F above the heatup curve or the cooldown curve, and not less than the minimum permissible temperature for ISLH testing. However, the criticality curve is not operationally limiting; a more restrictive limit exists in LCO 3.1.1.4,
" Minimum Temperature for Criticality."
The consequence of violating the LCO limits is that the RCS has been operated under conditions that can result in brittle failure of the RCPB, I
possibly leading to a nonisolable leak or loss of coolant accident. In the event these limits are exceeded, an evaluation must be performed to determine the effect on the structural integrity of the RCPB components.
The ASME Code,Section XI, Appendix E (Ref. 6) provides a recommended methodology for evaluating an operating event that causes an excursion outside the limits.
APPLICABLE SAFETY ANALYSIS The P/T limits are not derived from Design Basis Accident (DBA) analyses.
They are prescribed during normal operation to avoid encountering pressure, temperature, and temperature rate of change conditions that might cause undetected flaws to propagate and cause nonductile failure of the RCPB, an unanalyzed condition.
Reference 2 establishes the methodology for determining the P/T limits.
Although the P/T limits are not derived from any DBA, the P/T limits are acceptance limits since they preclude operation in an unanalyzed condition.
RCS P/T limits satisfy Criterion 2 of 10CFR50.36(c)(2)(ii).
LCD The two elements of this LC0 are:
a.
The limit curves for heatup, cooldown, and ISLH testing; and b.
Limits on the rate of change of temperature.
The LCO limits apply to all components of the RCS, except the Pressurizer. These limits define allowable operating regions while providing margin against nonductile failure.
The limits for the rate of change of temperature control the thermal gradient through the vessel wall and are used as inputs for calculating the heatup, cooldown, and ISLH testing P/T limit curve. Thus, the LCO for the rate of change of temperature restricts stresses caused by thermal gradients and also ensures the validity ~f thi~P/T~1imit curves.
o MILLSTONE - UNIT 3
" 3/4 M Amendment No.
0528
ZAlSERT 3. 4. 9 REACTOR COOLANT SYSTEM BASES PRESSURE / TEMPERATURE LIMITS (continued)
Violating the LCO limits places the reactor vessel outside of the bounds of the analyses and can increase stresses in other RCPB components. The consequences depend on several factors, as follows:
a.
The severity of the departure from the allowable operating P/T regime or the severity of the rate of change of temperature; b.
The length of time the limits were violated (longer violations allow the temperature gradient in the thick vessel walls to become more pronounced); and c.
The existences, sizes, and orientations of flaws in the vessel material.
MS grp #
APPLICABILITY a
The RCS P/T limits LC0 provides a definition of acceptable operation for prevention of nonductile failure in accordance wi 10CFR50, Appendix G (Ref.1).
The P/T limits were developed to provide.g for operation during heatup or cooldown (MODES 3, 4, and 5) or ISLH testing, in keeping with the concern for nonductile failure.
The limits do not apply to the Pressurizer.
During MODES I and 2, other Technical Specifications provide limits for operation that can be more restrictive than or can supplement these P/T limits. LC0 3.2.5, "DNB Parameters"; LC0 3.2.3.1 and 3.2.3.2, "RCS Flow Rate and Nuclear Enthalpy Rise Hot Channel Factor Four Loops Operating /Three Loops Operating";
LCO 3.1.1.4, " Minimum Temperature for Criticality"; and Safety Limit 2.1, " Safety Limits," also provide operational restrictions for pressure and temperature and maximum pressura. Furthermore, MODES 1 and 2 are above the temperature range of concern for nonductile failure, and stress analyses have been performed for normal maneuvering profiles, such as power ascension or decut"w ACTIONS Operation outside the P/T limits must be corrected so that the RCPB is returned to a condition that has been verified by stress analyses. The 30 minute Allowed Outage Time (A0T) reflects the urgency of restoring the parameters to within the analyzed range.
Most violations will not be severe, and the activity can be accomplished in this time in a controlled manner.
Besides restoring operation within limits, an evaluation is required to determine if RCS operation can continue.
The evaluation must verify the RCPB integrity remains acceptable and must be completed before continuing operation.
Several methods may be used, including comparison with pre-analyzed transients in the stress analyses, new analyses, or inspection of the components.
MILLSTONE - UNIT 3 B 3/C 4 10 Amendment No.
0528
TA/ SERT 3e % 9 REACTOR COOLANT SYSTEM BASES
~
PRESSURE / TEMPERATURE LIMITS (continued)
]
ASME Code,Section XI, Appendix E (Ref. 6), may be used to support the evaluation. However, its use is restricted to evaluation of the vessel beltline.
The.72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months A0T is reasonable to accom)lish the evaluation. The evaluation for a mild violation is possible within t11s time, but more severe violations may require special, event specific stress analyses or inspections.
A favorable evaluation must be completed before continuing to operate.
This evaluation must be completed whenever a limit is exceeded.
Restoration a
within 30 minutes alone is insufficient because higher than analyzed stresses may have occurred and may have affected the RCPB integrity.
If the required remedial actions are not completed within the allowed times, the plant must be placed in a lower MODE because either the RCS remained in an unacceptable P/T region for an extended period of increased stress or a sufficiently severe event caused entry into an unacceptable region.
Either possibility indicates a need for more careful examination of the event, best accomplished with the RCS at reduced pressure and temperature.
In reduced pressure and temperature conditions, the possibility of propagation with undetected flaws is decreased.
If the required restoration activity cannot be accomplished within 30 minutes, action must be implemented to reduce pressure and temperature as specified in the ACTION statement.
If the required evaluation for continued operation cannot be accomplished within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or the results are indeterminate or unfavorable, action must proceed to reduce pressure and temperature as specified in the Action statement.
A favorable evaluation must be completed and documented before returning to operating pressure and temperature conoitions.
Pressure and temperature are reduced by bringing the plant to MODE 3 within t
6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and to MODE 5 with RCS pressure-< 500 psia within the next 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .
The aMA0Ts 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.
SURVEILLANCE RE0VIREMENTS Verification that operation is within the LC0 limits as well as the limits of Figures 3.4-2 and 3.4-3 is required every 30 minutes when RCS pressure and temperature conditions are undergoing planned changes.
This Frequency is considered reasonable in view of the control room indication available to monitor RCS status.
MILLSTONE - UNIT 3 03/4441 Amendment No.
0628
l
[NSEg7* 3,f f REACTOR C00LANT SYSTEM BASES r
PRESSURE / TEMPERATURE LIMITS (continued) l-1 0
Surveillance for heatup, cooldown, or ISLH testing may be discontinued when the I
definition given in the relevant plant procedure for ending the activity is satisfied.
i This Surveillance Requirement is only required to be performed during system-heatup, cooldown, and-ISLH testing.
No Surveillance Requirement is given for criticality operations -because LCO 3.1.1.4 contains a more restrictive j
requirement.
l The Surveillance Requirement to remove and examine the reactor vessel material
~
irradiation surveillance specimens is in accordance with the requirements of 10CFR50, Appendix H.-
1 REFERENCES i
1.
.10CFR50, Appendix G.
2.
ASME, Boiler and Pressure Vessel Code,Section III, Appendix G.
3.
ASTM E 185-82, July 1982.
4.
10CFR50, Appendix H.
5.
P.egulatory Guide 1.99, Revision 2, May 1988.
6.
ASME,. Boiler and Pressure Vessel-Code,Section XI, Appendix E.
i PRESSURIZER BACKGROUND i
i
' The Pressurizer.is part of the RCPB, but is not subject to the same restrictions as the rest of the RCS.
This LC0 limits the temperature changes of the Pressurizer. and allowable temperature differentials, within the design assumptions and the stress limits for cyclic operation.
i 1
R i
r 1
4 I
MILLSTONE - UNIT 3 00/04-1:
Amendment No.
0528 i
i
Z N S G W.f. 4'. f -
REACTOR COOLANT SYSTEM BASES PRESSVRIZER (continued)
The LCO contains the Pressurizer limits for heatup, cooldown, and spray water temperature differential. Each temperature limit defines an acceptable region for normal operation. The limits that apply to the Pressurizer are as follows: The Pressurizer heatup and cooldown rates shall not exceed 100*F/hr and 200*F/hr, respectively. The spray shall not be used if the temperature difference between the Pressurizer and the spray fluid is greater than 320*F.
The heatup limit represents a different set of restrictions than the cooldown limit because the directions of the thermal gradients through the Pressurizer wall are reversed.
The thermal gradient reversal alters the location of the tensile stress between the outer and inner walls.
The consequence of violating the LC0 limits is that the Pressurizer has been operated under conditions that can result in failure, possibly leading to a nonisolable leak or loss of coolant accident.
In the event these limits are exceeded, an evaluation must be performed to determine the effect on the structural integrity of the Pressurizer.
APPLICABLE SAFETY ANALYSIS The Pressurizer temperature limits are not derived from Design Basis Accident (DBA) analyses.
They are prescribed during normal operation to avoid encountering temperature and temperature rate of change conditions that might cause the initiation / propagation of undetected cracks and cause failure of the pressure boundary.
LC.Q The two elements of this LCO are:
a.
Limits on the rate of change of temperature; and b.
Limits on the spray water differential temperature.
The LC0 limits apply to the Pressurizer. These limits define allowable operating regions and permit a large number of operating cycles while providing a margin to nonductile failure.
The limits for the rate of change of temperature control the thermal gradient through the Pressurizer wall and, therefore, restricts stresses caused by thermal gradients.
Violating the LCO limits places the Pressurizer outside of the bounds of the stress analyses. The consequences depend on several factors, as follow:
a.
The severity of the rate of change of temperature; MILLSTONE - UNIT 3 0 0/4 4 Amendment No.
0628
fAISEnf 3. '/. 9 REACTOR COOLANT SYSTEM BASES PRESSURI7ER (continued) b.
The length of time the limits were violated (longer violations allow the temperature gradient in the Pressurizer walls to become more pronounced); and c.
The existences, sizes, and orientations of flaws in the Pressurizer material.
APPLICABILITY The Pressurizer temperature limits LCO provides a definition of acceptable operation for prevention of failure. The temperature limits were developed to provide requirements for operation during heatup or cooldown, and their Applicability is at all times in keeping with the concern for failure.
ACTIONS Operation outside the temperature limits must be corrected so that the Pressurizer is returned to a condition that has been verified by stress analyses.
The 30 minute A0T reflects the urgency of restoring the parameters to within the analyzed range.
Most violations will not be severe, and the activity can be accomplished in this time in a controlled manner.
Besides restoring operation within limits, an evaluation is required to determine if Pressurizer operation can continue.
The evaluation must veri fy the Pressurizer integrity remains acceptable and must be completed before continuing operation. Several methods may be used, including comparison with pre-analyzed transients in the stress analyses, new analyses, or inspection of the components.
The 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months A0T is reasonable to accomplish the evaluation. The evaluation for a mild violation is possible within this time, but more severe violations may require special, event specific stress analyses or inspections.
A favorable evaluation must. be completed before continuing to operate.
This evaluation must be completed whenever a limit is exceeded.
Restoration within 30 minutes alone is insufficient because higher than analyzed stresses may have occurred and may have affected the Pressurizer integrity.
If the required remedial actions are not completed within the allowed times, the plant must be placed in a lower MODE because a sufficiently severe event may have caused entry into an unacceptable region. This possibility indicates a need for more careful examination of the event, best accomplished with the Pressurizer at reduced pressure. In reduced pressure conditions, the possibility of propagation with undetected flaws is decreased.
If the required restoration activity cannot be accomplished within 30 minutes, action must be implemented to reduce pressure as specified in the ACTION statement.
MILLSTONE - UNIT 3 4-af4-4=N Amendment No.
0528
ZNsur 3. 4. 9 REACTOR COOLANT SYSTEM BASES PRESSURIZER fcontinued)
If the required evaluation for continued operation cannot be accomplished within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or the results are indeterminate or unfavorable, action must proceed to reduce pressure as specified in the Action statement.
A favorable evaluation must be completed and documented before returning to operating pressure conditions.
Pressure is reduced by bringing the plant to MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . Pressure is further reduced by bringing the plant to MODE 4 or 5 and reducing Pressurizer pressure < 500 psia within the next 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .
The *n MOTS are reasonable,- based on operating experience, to reach the r
required plant conditions from full power conditions in an orderly manner and without challenging plant systems.
SURVEILLANCE RE0VIREMENTS Verification that operation is within the LC0 heatup and cooldown limits is required every 30 minutes when Pressurizer temperature conditions are undergoing
' planned changes. This Frequency is considered reasonable in view of the control room indication available to monitor Pressurizer status. Surveillance for heatup or cooldown may be discontinued when the definition given in the relevant plant procedure for ending the activity is satisfied. The Surveillance Requirement for heatup or cooldown is only required to be performed during system heatup and cooldown.
Verification that operation is within the LCO spray water temperature differential limit is required every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months when auxiliary spray is in operation. This Frequency is considered reasonable in view of the control room indication available to monitor Pressurizer status.
OVERPRESSURE PROTECTION SYSTEMS BACKGROUNQ The Cold Overpressure Protection System limits RCS pressure at low temperatures so the integrity of the reactor coolant pressure boundary (RCPB) is not compromised by violating the pressure and temperature (P/T) limits of 10CFR50,AppendixG(Ref.1). The reactor vessel is the limiting RCPB component for demonstrating such protection.
Cold Overpressure Protection consists of two PORVs with nominal lift setting as specified in Figures 3.4-4a and 3.4-4b, or two residual heat removal (RHR) suction relief valves, or one PORV ud one RHR suction relief valve, or a depressurized RCS and an RCS vent of sufficient size.
Two relief valves are required for redundancy. One relief valve has adequate relieving capability to prevent overpressurization of the RCS for the required mass input capability.
MILLSTONE - UNIT 3 3 0/' a '5-Amendment No.
0628 o
- _ -. -.. - _. -. ~ - - - ~ - ~ - -.
JZ A/S G A Y.3. 4. 1 D
REACTOR C00LANT SYSTEM i
BA
. SES-r I
QERRRES50RE PROTECTION SYSTEMS f continuedi-3 The reacior vessel material is less tough at low-te.9peratures?than at normal operating temperature. As the vessel neutron exposure occumulates, the material toughness decreases-and becomes less resistant to stress at-low temperaturos (Ref. 2)
RCS pressure, therefore, is maintained low at law temperatures and is i
sncreased only'as temperature is increased.
The potential for vessel overpressurization is most acute when the RCS is water.
solid, occurring while shutdown; a pressure fluctuation can oc. cur more quickly than an operator can react to relieve the condition.
Exceeding the RCS-P/T-limits by a significant. amount could cause nonductile cracking ef the reactor vessel.- LCO 3.4.9.1r " Pressure / Temperature Limits - Reactor Coolant System,"
requires administrative control of RCS pressure and temperature during heatup and cooldown to prevent exceeding _the limits provided in Figures 3.4-2 and 3.4-3.-
_This LCO provides RCS overpressure protection by limiting mass input capability 1
and requiring adequate pressure relief capacity.- Limiting mass input capebility l
requires all Safety Injection (SIH) pumps and-all but one centrifugal chcrging i
pump to be. incapable of injection into the RCS.
The pressure relief capacity requires either two redundant relief valves or a depressurized RCS and an RCS vento of sufficient size.
One relief valve or the open RCS. vent is the ii
- overpressure-protection device that acts,to terminate an increasing pressurc
- event.
l With minimum mass input capability, the ability to provide core coolant addition 4
i is restricted. -The LCO does not require the makeup control system' deactivated or the safety injection (SI) actuation circuits blocked.
Due to the lower pressures in the COP 7. MODES _ and the expected core decay heat levels, the makeup system can provide}dequate flow via the makeup control valve.
Y
<ew A7wr % tad ~
-PORV-Reauirements 1
i Astdesigned, the-PORY Cold Overpressure Protection (COPPS)-is signaled to open i
if the.RCS: pressure approaches a limit determined by the COPPS actuation logic.
The COPPS actuation logic monitors both RCS temperature and-RCS pressure and j
determines when -the nominal setpoint of Figure - 3.4-4a or Figure 3.4-4b is-approached.
The wide range RCS temperature indications are auctioneered to select the lowest temperature signal.
.The lowest temperature signal:is-processed through a function generator that calculates a pressure set)oint for_ that temperature.
The calculated pressure setpoint :is,then comparec with the indicated RCS pressure from. a wide' range e
pressure channel.
If the indicated-pressure meets or exceeds the calculated value,;a' PORV is signaled to open.
t
- Figure 3.4-4a and= Figure 3.4-4b present the PORV setpoints for COPPS.
Above.
_110*F, the setpoints are staggered so only one valve opens during a low 7
F l
2 MILLSTONE UNIT 3 4 U; 4c --
Amendment No.
0528 h
7
.[-,--.
,.,-m,,
,.,.,y w
. _., -. - + _
m
.s.~
y-_n
~
ZNSEX'T M. 9 REACTOR COOLANT SYSTEM BASES QJERPRESSURE PROTECTION SYSTEMS (continued) temperature overpressure transient. Setting both valves to the values of Figure 3.4-4a and Figure 3.4-4b within the tolerance allowed for the calibration accuracy, ensures that the' Reference 1 limits will not be exceeded for the analyzed isothermal events.
Whe a PORV is opened, the release of coolant will cause the pressure increase to slow and reverse. As the PORY releases coolant, the RCS pressure decreases until a reset pressure is reached and the valve is signaled to close.
The pressure continues to decrease below the reset pressure.as the valve closes.
RHR Suction Relief Valve Reouirements The isolation valves between the RCS and the RHR suction relief valves must be open to make the RHR suction relief valves OPERABLE for RCS overpressure mitigation. The RHR suction relief valves are spring loaded, bellows type water relief valves with setpoint tolerances and accumulation limits established by 4
Section III of the American Society of Mechanical Engineers (NJE) Code (Ref. 3) for Class 2 relief valves, When the RHR system is operated for decay heat removal or low pressure letdown d
control, the isolation valves between the RCS and the RHR suction relief valves are open, and the RHR suction relief valves are exposed to the RCS and are able to relieve pressure transients in the RCS.
RCS Vent Reouirements Once the RCS is depressurized, a vent exposed to the containment atmosphere will maintain the RCS at containment ambient pressure in an RCS overpressure transient, if the relieving requirements of the trannient do not exceed the capabilities of the vent. Thus, the vent path must be capable of relieving the flow resulting from the limiting mass or-heat input transient, and maintaining pressure below the P/T limits for the analyzed isothermal events.
For an RCS venE to meet the flow capacity requirement, it requires removing a l Pressurizer safety valve, removing a PORV and disabling its block valve in the open position, removing a Pressurizer manway, or similarly establishing a vent by opening an RCS vent valve provided - that the opening meets the size requirements.
The vent path must be above the level of reactor coolant, so as not to drain the RCS when open.
MILLSTONE - UNIT 3
@ J-Amendment No.
0528 5
za. ster 3.v.7 RfACTOR COOLANT SYSTEM BASES OVERPRESSURE PROTECTION SYSTEMS (continued)
APPLICABLE SAFETY ANALYSIS Safety analyses (Ref. 4) demonstrate that the reactor vessel is adequately protected against exceeding the P/T limits for the analyred isothermal events.
In MODES 1, 2, AND 3, and in MODE 4, with RCS cold leg temperature exceeding 275'F, the pressurizer safety valves will provide RCS overpressure protection in the ductile region. At 275'F and below, overpressure prevention is provided by two means:
(1) two OPERABLE relief valves, or (2) a depressurized RCS with a sufficiently sized RCS vent, as required by NUREG-0800, RSB 5-2 for temperatures less than RTuor + 90'F.
Each of these means has a limited overpressure relief capability.
The required RCS temperature for a given pressure increases as the reactor vessel material toughness decreases due to neutron embrittlement.
Each time the Technical Specification curves are revised, the cold overpressure protection must be re-evaluated to ensure its functional requirements continue to be met using the RCS relief valve method or the depressuri.:ed and vented RCS condition.
Transients capable of overpressurizing the RCS are categorized as either mass or heat input transients, examples of which follow:
(
Mass Inout Transients a.
Inadvertent safety injection; or b.
Charging / letdown flow mismatch Heat Inout Transients a.
Inadvertent actuation of Pressurizer heaters; b.
Loss of RHR cooling; or c.
Reactor coolant pump (RCP) startup with temperature asymmetry within the RCS or between the RCS and steam generators.
The Technical Specifications ensure that mass input transients beyond the operability of the cold overpressure protection means do not occur by rendering all Safety Injection Pumps and all but one centrifugal charging pump incapable of injecting into the RCS whenever an RHR suction relief valve is unisolated from the RCS or whenever any PORV has COPPS armed and its block valve open.
The Technical Specifications ensure that energy addition transients beyond the operability of the cold overpressure protection means do not occur by limiting reactor coolant pump starts. LC0 3.4.1.4.1, " Reactor Coolant Loops and Coolant Circulation - Cold Shutdown - Loops Filled," LC0 3.4.1.4.2, " Reactor Coolant MILLSTONE - UNIT 3 1 2/ '. ' O -
AmendhentNo.
0528 i
j
Z A/S 6 2 1 3.</. f REACTOR COOLANT SYSTEM BASES j
OVERPRESSURE PROTECTION SYSTEMS (continued) l Loops and Coolant Circulation Cold Shutdown Loops Not Filled," and LCO 3.4.1.3, " Reactor Coolant Loops and Coolant Circulation - Hot Shutdown" limit reactor coolant pump starts to one of the following plant conditions:
a.
An RCP is running, and The wide range cold leg temperature of any unisolated RCS loop is
>160*F, or b.
Two or more RCS loops are isolated, and An RCP is not running, and The secondary side water temperature of any steam generator in an unisolated loop is equal to or less than the wide range cold leg temperature of any unisolated RCS loop, or i
c.
No more than one RCS loop is isolated, and An RCP is not running, and Any RHR suction relief valve is unisolated from the RCS, and The secondary side water temperature of any steam generator in an unisolated loop is either:
>200'F and equal to or less than the wide range cold leg l
l temperature of any unisolated RCS loop, or 1200'F and 150*F hotter than the wide range cold leg temperature of any unisolated RCS loop. (Note: Reactor coolant pumps cannot I
be run with the wide range cold leg temperature of any unisolated l
RCS loop <160*F if any PORV has COPPS armed and has its block valve open.), or d.
No more than one RCS loop is isolated, and An RCP is not running, and The RHR suction relief valves are isolated from the RCS, and The wide range cold leg temperature of any unisolated RCS loop 2160*F, and Any PORV has COPPS armed and has its block valve open, and The secondary side water temperature of any steam generator in an unisolated loop is either:
equal to or less than the wide range cold leg temperature of any unisolated RCS loop, or
<250*F and 150*F hotter than the wide range cold leg temperature of any unisolated RCS loop, or e.
The RHR suction relief valves are isolated from the RCS, and Both PORVs are isolated or COPPS is blocked, and The wide range cold leg temperature of any unisolated RCS loop is
>275'F.
S MILLCTONE - UNIT 3 0 S/4 ?-3 Amendment No.
0528
y gf REACTOR COOLANT SYSTEM BASES OVERPRESSURE PROTECTION SYSTEMS (continued)
The cold overpressure transient analyses demonstrate that either one relief valve or the depressurized RCS and RCS vent can maintair HCS pressure below limits when RCS letdown is isolated and only one centrifug:
charging pump is operating.
Thus, the LCO allows only one centrifugal charging pump capable of injecting when cold overpressure protection is required.
The cold overpressure protection enabling temperature is conservatively established at a value 2 275'F based on the criteria described in Branch
' Technical Position RSB 5-2 provided in the Standard Review Plan (NUREG-0800).
PORV Performance The 10CFR50 Appendix G analyses show that the vessel is protected against non-ductile failure when the PORVs are set to open at the values shown in Figures 3.4-4a and 3.4-4b within the tolerance allowed for the calibration accuracy. The curves are derived by analyses that model the performance of the PORV cold overpressure protection system (COPPS), assuming the limiting mass and heat transients of one centrifugal charging pump injecting into the RCS, or the energy addition as a result of starting an RCP with temperature asymmetry between the RCS and the' steam generators.
These analyses consider pressure overshoot and undershoot beyond the PORV opening and closing, resulting from signal processing and valve stroke times.
' The PORV setpoints in Figures 3.4-4a and 3.4-4b will be updated when the P/T limits conflict with the cold overpressure analysis limits. The P/T limits are periodically modified as the reactor vessel material toughness decreases due to neutron embrittlement.
Revised limits are determined using neutron fluence projections and the results of testing of the reactor vessel material irradiation surveillance specimens. The Bases for LCO 3.4.9.1, " Pressure / Temperature Limits
- Reactor Coolant System (Except the Pressurizer)," discuss these evaluations.
The PORVs are considered active components.
Thus, the failure of one PORV is assumed to represent the worst case, single active failure.
~
RHR Suction Re1ief Valve Performance The RHR suction relief valves do not have variable pressure and temperature lift setpoints as do the PORVs. Analyses show that one RHR suction relief valve with a setpoint at or between 426.8 psig and 453.2 psig will pass flow greater than that required for the limiting cold overpressure transient while maintaining RCS pressure less than the isothermal P/T limit curve. Assuming maximum relief flow requirements during the limiting cold overpressure event, an RHR suction relief valve will maintain RCS pressure to 1110% of the nominal lift setpoint.
- Although each RHR suction relief valve is a passive spring loaded device, which meets single failure criteria, its location within the RHR System precludes meeting single failure criteria when spurious RHR suction isolation v 've or RHR suction valve closure is postulated. Thus the loss of an RHR suct ce relief MILLSTONE - UNIT 3 o c 4 20 Amendment No.
052%
$NS$RT 3. % 1 REACTOR COOLANT SYSTEM BASES OVERPRESSURE PROTECTION SYSTEMS (continued) valve is the worst case single failure. Also, as the RC
/T limits are revised to reflect change in toughness in the reactor vessel materials, the RHR suction relief valve's analyses must be re-evaluated to ensure continued accommodation of the design bases cold overpressure transients.
2 RCS Vent Performance With the RCS depressurized, analyses show a vent size of 15.4 square inches is capable of mitigating the allowed cold overpressure transient. The capacity of this vent size is greater than the flow of the limiting transient, while maintaining RCS pressure less than the maximum pressure on the isothermal P/T limit curve.
The RCS vent size will be re-evaluated for compliance each time the isothermal P/T limit curves are revised.
The RCS vent is a passive device and is not subject to active failure.
The RCS vent satisfies Criterion 2 of 10CFR50.36(c)(2)(ii).
RCP Seal Protection As described above, the analyses of the cold overpressure transients result in pressure overshoot and undershoot beyond the PORV opening and closing setpoints, resulting from signal processing and valve stroke times. The valve osarshoots l
are considered in the generation of the PORV setpoints presented in Figures 3.4-4a and 3.4-4b.
The valve undershoots are also evaluated in terms of potential damage to the RCP
~~1 seal. The minimum pressure, considering valve undershoot, must be higher than that required to maintain the RCP #1 seal as a film riding seal.~~
This requirement resulted in restrictions on the operation of pumps when the cold overpressure protection is being provided by one or two PORVs. Specifically, a.
When the RCS cold leg temperature of any unisolated loop is less than 160 degrees F, the PORY block valves are open, and the PORV's Cold Overpressure Protection System (COPPS) is armed, no RCPs may be in operation.
LC0 3.4.1.4.1,
" Reactor Coolant Loops and Coolant Circulation - Cold Shutdown - Loops Filled," and LC0 3.4.1.4.2,
" Reactor Coolant Loops and Coolant Circulation - Cold Shutdown - Loops Not Filled," provide this protection.
b.
When COPPS is armed, with the steam generator secondary side 2 250'F, heat injection transients due to the start of the first RCP with a tem)erature asymmetry between the RCS and the steam generators is pro 11bited.
LC0 3.4.1.3, " Reactor Coolant System - Hot Shutdown,"
provides this protection for MODE 4.
AmendNntNo.
MILLSTONE - UNIT 3 W+
0528
ENStrdT 3. */. 9 REACTOR COOLANT SYSTEM BASES OVERPRESSURE PROTECTION SYSTEMS (continued)
I c.
When COPPS is armed, PORY undershoot is analyzed for mass injection transients limited to one charging pump.
LCO 3.4.9.3,
" Reactor Coolant System - Overpressure Protection Systems," provides this 2
protection by requiring both safety injection pumps and all but one charging pump to be incapable of injection into the RCS.
In order to provide protection for the RCP #1 seal, a PORV setpoint of 1595 psia for temperatures 2 160 degrees F must be met. This minimum setpoint is derived by adding the applicable train uncertainty and valve undershoot to the required minimum RCS pressure required for seal integrity.
Due to the differing instrument uncertainties for the two trains of PORV COPPS, the train with the highest uncertainty is paired to the high setpoint curve.
LCD This LC0 requires that cold overpressure protection be OPERABLE and the maximum mass input be limited to one charging pump. failure to meet this LC0 could lead to the loss of low temperature overpressure mitigation and violation of the Reference 1 isothermal limits as a result of an operational transient.
To limit the mass input capability, the LCO requires a maximum of one centrifugal charging pump capable of injecting into the RCS.
The elements of the LC0 that provides low temperature overpressure mitigation through pressure relief are:
1.
Two OPERABLE PORVs; or
~
A PORV is OPERABLE for cold overpressure protection when its block valve is open, its lift setpoint is set to the nominal setpoints provided by Figure 3.4-4a or 3.4-4b and when the surveillance requirements are met.
2.
Two OPERABLE RHR suction relief valves; or An RHR suction relief valve is OPERABLE for cold overpressure protection when its isolation valves from the RCS are open and when its setpoint is at or between 426.8 psig and 453.2 psig, as verified by required testing.
3.
One OPERABLE PORV and one OPERAELE RHR suction relief valve; or 4.
A depressurized RCS and an RCS vent.
An RCS vent is OPERABLE when open with an area of 2 5.4 square inches.
Each of these methods of ovepressure prevention is capable of mitigating the limiting cold overpressure transient.
s MILLSTONE - UNIT 3
-S-.3/; 4-22 Amendment No.
0528
[MSERT 3.4 9 REACTOR COOLANT SYSTEM BASES OVERPRESSURE PROTECTION SYSTEMS (continued)
APPLICABILITY This LC0 is applicable in MODE 4 when any RCS cold leg temperature is 1275'F, in MODE 5, and in MODE 6 when the head is on the reactor vessel. The Pressurizer safety valves provide RCS overpressure protection in the ductile region (i.e.
>275'F). When the reactor head is off, overpressurization cannot occur.
o a
LC0 3.4.9.1 Pressure / Temperature Limits [provides the operational P/T limits for all MODES.
C0 3.4.2.2, " Safety Valves - Operating," requires the OPERABILITY of the Pressurizer safety valves that provide overpressure protection during MODES I, 2, and 3, and LC0 3.4.2.1, " Safety Valves - Shutdown," requires the OPERABILITY of the Pressurizer safety valves that provide overpressure protection during MODE 4.
Low temperature overpressure prevention is most critical during shutdown when the RCS is water solid, and a mass or heat input transient can cause a rapid increase in RCS pressure when little or no time exists for operator action to mitigate the event.
ACTIONS
~~a. and b.~~
l With two or more centrifugal charging pumps capable of injecting into the RCS, or with any SIH pump capable of injecting into the RCS, RCS overpressurization is possible.
To inmediately initiate action to restore restricted mass input capability to the i
RCS reflects the urgency of removing the RCS from this condition.
Required Action a. is modified by a Note that permits two centrifugal charging 4
pt"ups capable of RCS injection for i 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months to allow for pump swaps. This is a controlled evolution of short duration and the proceaure prevents having two charging pumps simultaneously out of pull-to-lock while both charging pumps are capable of injecting into the RCS, L.
In MODE 4 when any RCS cold leg temperature is s 275'F, with one required relief valve inoperable, the RCS relief valve must be restored to OPERABLE status within an allowed outage time (A0T) of 7 days. Two relief valves in any combination of the PORVs and the RHR suction relief valves are required to provide low temperature overpressure mitigation while withstanding a single failure of an active component.
4 3
MILLSTONE - UNIT 3 40/: 4-2r Amendment No.
0528
ZAl.SER7 3.4.1 REACTOR COOLANT SYSTEM BASES CVERPRESSURE PROTECTION SYSTEMS _(continued)
The A0T in MODE 4 considers the facts that only one of the relief valves is required to mitigate an overpressure transient and that the likelihood of an active failure of the remaining valve path during this time period is very low.
The RCS must be depressurized and a vent must be established within the following 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months if the required relief valve is not restored to OPERABLE within the required A0T of 7 days.
L The consequences of operational events that will overpressure the RCS are more severe at lower temperatures (Ref. 7). Thus, with one of the two required relief valves inoperable in MODE 5 or in MODE 6 with the head on, the A0T to restore two valves to OPERABLE status is 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .
The A0T represents a reasonable time to investigate and repair several types of relief valve failures without exposure to a lengthy period with only one OPERABLE relief valve to protect against overpressure events.
The RCS must be depressurized and a vent must be established within the following 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months if the required relief valve is not restored to OPERABLE within the required A0T of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .
L.
The RCS must be depressurized and a vent must be established within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months when both required COPS relief valves are inoperable.
The vent must be sized 2 5.4 square inches to ensure that the flow capacity is greater than that required for the worst case cold overpressure transient reasonable during the applicable MODES.
This action is needed to protect the RCPB from a low temperature overpressure event and a possible non-ductile failure of the reactor vessel.
The time required to place the plant in this Condition is based on the relatively low probability of an overpressure event during this tirae period due to increased operator awareness of administrative control requirements.
SVRVEILLANCE RE0VIREMENTS
[4.4.9.3.1 Performance of an ANALOG CHANNEL OPERATIONAL TEST is required within 31 days 1
prior to entering a condition in which the PORV is required to be OPERABLE and every 31 days on each required PORV to verify and, as necessary, adjust its lift setpoint.
The ANALOG CHANNEL OPERATIONAL TEST will verify the setpoint in accordance with the nominal values given in Figures 3.4-4a and 3.4-4b.
PORV actuation could depressurize the RCS; therefore, valve operation is not required.
1
- s MILLSTONE - UNIT 3
-t ?/' '-? F Amendment No.
0528
.TAIS Ett r 3. 4.1 REACTOR COOLANT SYSTEM BA3ES OVERPRESSURE PROTECTION SYSTEMS (continued)
Performance of a CHANNEL CALIBRATION on each required PORV actuation channel is recuired once each REFUELING INTERVAL to adjust the channel so that it responds i
anc the valve opens within the required range and accuracy to a known input.
The PORV block valve must be verified open and COPPS must be verified armed every 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months to provide a flow path and a cold overpressure protection actuation circuit for each required PORV to perform its function when required. The valve is remotely verified open in the main control room.
This Surveillance is performed if credit is being taken for the PORV to satisfy the LCO.
The block valve is a remotely controlled, motor operated valve. The power to the valve operator is not required to be removed, and the manual operator is not required to be locked in the open position. Thus, the block valve can be closed in the event the PORV develops excessive leakage or does not close (sticks open) after relieving an overpressure transient.
The 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Frequency is considered adequate in view of other administrative controls available to the operator in the control room, such as valve position indication, that verify the PORV block valve remains open.
[4.4.9.3.2 Each required RHR suction relief valve shall be demonstrated OPERABLE by l
verifying the RHR suction valves, 3RH5*MV6701A and 3RHS*M87010, are open when l
suction relief valve 3RHS*RV8708A is being used to meet the LC0 and by verifying i
the RHR suction valves 3RHS*MV8702B and 3RHS*MV8702C, are open when suction relief valve 3RHS*RV8708B is being used to meet the LCO.
Each required RHR suction relief valve shall also be demonstrated OPERABLE by testing it in accordance with 4.0.5. This Surveillance is only required to be performed if the RHR suction relief valve is being used to meet this LCO.
The RHR suction valves are verified to be open every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . The Frequency is considered adequate in view of other administrative controls such as valve status indications available to the operator in the control room that verify the RHR suction valves remain open.
The ASME Code,Section XI (Ref. 8), test per 4.0.5 verifies OPERABILITY by proving proper relief valve mechanical motion and by measuring and, if required, adjusting the lift setpoint.
De MILLSTONE - UNIT 3 3 3/a 5-Amendment No.
a 0528
Z ris S R r 3. 4 1i REACTQR COOLANT SYSTEM
-BASES DVERPRESSURE PROTECTION-SYSTEMS fcontinued)
.4.9.3.3:
.The ' RCS vent of 2 5.4-square inches is proven OPERABLE._by verifying its open-condition either:
a.-
Once every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months for a vent valve that cannot be locked ~ open.
b.
Once every 31 days for a valve that is locked, sealed, or secured in position or any-other passive vent path. A removed Pressurizer safety valve fits this category.
This - passive vent arrangement must only be open to be OPERABLE.
This Surveillance is required to be performed if the vent is being used to satisfy-the-pressure relief requirements of the LCO.
4.4.g.3.4 and
.4.9.3.5-To minimize the potential for a low temperature ~ overpressure event by limiting the mass input capability, all SIH; pumps and all but one centrifugal charging pump are verified incapable of injecting into the RCS.
The SIH pumps and charging pumps are rendered incapable of injecting into the RCS'
- through removing-the power from the. pumps-by racking the breakers out under administrative-control. : Alternate methods of control may be employed using at -
least:two.-independent means to. prevent an. injection:into the RCS. This may be accomplished through any of the following methods: 1) placing the pump in pull-
-to lock'(PTL) and pulling its UC fuses, 2) placing the pump in pull:to lock (PTL)--
and closing.the pump discharge-valve (s) to the injection line, 3) closing the pump. discharge valve (s) to the injection line and either removing power from the valve operator (s) or locking manual valves closed,. and 4) closing the valve (s)-
from the injection source and-either removing power from the valve operator (s) or locking manual ~ valves closed.
An SIH pump may'be energized for testing or for filling the Accumulators provided
~~it is incapable of injecting into the RCS.~~
The Frequency of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is sufficient, considering other indications and alarms avai_lable to-the operator in the enntrol room, to verify the required status of the equipment'.
' REFERENCES 1.
10CFR50,. Appendix G 2.-
-Generic Letter 88-11 3.
-ASME, Boiler _and Pressure _ Vessel Code,Section III 4.
FSAR, Chapter.15 5._
10CFR50, Section 50.46 l
6.
~~10CFR50, Appendix K 7.~~
Generic Letter 90-06 8.
ASME, Boileriand Pressure Vessel Code,Section XI MILLSTONE-- UNIT 3-40/444" Amendment No.
<osas
3/4.5 EMERGENCY CORE COOLING SYSTEMS-gfg 33 BASES-
~3/4.5.1~ ACCUMULATORS 1
-The CPCRASILITY of-each Reactor Coolant System'(RCS) accumulator ensures that a suf ficient. volume of borated water will be immediately forced into the reactor core through each of the cold legs in the event the RCS pressure falls below the pressure-of the accumulators. -This initial surge of water into the core provides the' initial cooling mechanism during large RCS pipe ruptures.
The limits-on accumulator volume, boron concentration and pressure ensure that the assumptions used for accumulator injection in the safety analysis are met.
The accumulator power operated isolation valves are considered to'be -
'" operating bypasses" in the context of IEEE Std. 279-1971, which requires that bypasses _ of a_ protective function be removed automatically whenever permissive l-conditions are-not met.
In addition, as these accumulator isolation valves f ail to meet single failure criteria,. removal of power to the valves is required.
The limits for operation with an accumulator inoperable for any-reason except an isolation valve closed minimizes the time exposure of the plant to a LOCA event-occurring concurrent with failure of an additieral accumulator which may result in unacceptable peak cladding temperatures.
If a closed 1
-isolation valve cannot be immediately opened, the full capability of one 1
accumulator.is not available and prompt action is required to place the reactor in a mode where this capability is not required.
3/4.5.2-and'3/4.5.3 ~ECCS SUBSYSTEMS The OPERAEILITY of two independent ECCS subsystems ensures that sufficient
-emergency core cooling capability'will be available in the event of a LOCA assuming the loss off one subsystem through any single failure-consideration.
Either subsystem operating in conjunction with the accumulators is capable of supplying sufficient core cooling to' limit the peak cladding temperatures within acceptable limits for all postulated break sizes ranging from thei
-double ended break of the largest RCS cold leg pipe downward.
In addition, each ECCS subsystem provides long-term core cooling capability in the recirculation moce during the accident recovery period.
N Vith the-RCS temperature b'elow 350*F, one OPERABLE ECCS sutsystem is acceptable without single f ailure considerati on the basis of the stable teactivity condition of the re. actor and the im ted core cooling requirements,
/ ul Ltx some nha ont 4 P norm / in f e c L',
/Nyj MILLSTONE - UNIT 3~
E 3/4 5-1
]
S
3 9 7 nrRCENeY CORE COOLING SYSTEMS January 3,1995 RASES I
ECCSSUBSYSTD5(Continued)
$ 1 +5 k[A'UNIIiN!?
%. m. u.... u. ;..u... u., z m..EI" E b 7 N S f!"8.f k a N b nhtv m = = m r m m _n A.n.a. - ami -" tim c,muyM :'-".c;;5EN;'p -"-,E9E'55
- - ~
U.E.' U,__E,5 3,U_5,M.N.,. M *"+_ X._.'.I".'.II' fE *2 n......., 5.. E...,.... _--
)
The Surveillance Requirements component ensures that at a minime,provided to ensure OPERABILITY of each the assumptions used in the safety analyses'are met'and that subsystem CPERABILITY is maintained. Surveillance Requirements for throttle valve position stops and flow balance testing provide assurance that proper ECCS flows will be maintained in the event of a LOCA. Maintenance of proper flow resistance and system to each ihjtdion point is necessary to: pressure drop in the piping f1 prevent total pump flow from exceeding runo.rt conditions when the system is)in its minim a resistance configuration, (2) provide the proper flow split between injection points in accordance w' th the assumptions used in the ECCS-LOCA analyses, and (3) provide an acceptabl6 level of total ECCS flow to all injection points equal to or above that assumed in the ECCS LOCA analyses.
Surveillance Requirement 4.5.2.C.t requires that the visual inspection of the containment be performed at least once daily if the containment has C-been entered that day and when the final containment entry is made. This will reduce the number of unnecessary inspections and also reduce personnel exposure.
3 /4. 5. 4 REFUEt1NG WATER STORAGE TANK The OPERABILITY of the refueling water storage tank as part of the ECCS ensures that a sufficient supply of borated water is a(RWST)ble for injec-vaila tion by the ECCS in the event of a LOCA.
The litits on RWST minima volume and boron concentration ensure that:
containment to permit recirculation (1) sufficient water is available within cooling flow to the core, and (2) the reactor will remain suberitical in the cold condition following sixing of the RWST and the.RCS water volumes with all control rods inserted except for the most reactive control assembly. These assaptions *are' consistent with the LOCA analyses.
The contained water volume limit includes an allowance for water not usable because of tank discharge line location or other physical characteristics.
The limits on contained water volume and boren concentration of the RWST also ensure a pH value of between 7.0 and 7.5 for the solution recirculated within containment after a LOCA. This pH band minimizes the effect of chloride and caustic stress corrosion on mechanical systems and components.
(
The maxinrJE/ minimum solution tecperatures for the RWST in MODES 1, 2, 3 and 4 are based on analysis assumptions.
M1LLSTONE - UNIT 3 8 3/4 5-2 Amendment No.100
d Docket No. 50-423 B16636 i -
1 i
j i
Millstone Nuclear Power Station Unit No. 3 Proposed Revision to Technical Specification j
Reduction in the Cold Overoressure Protection System Enablina Temperature l
(PTSCR 3-21-97)
Retvoed Paaes August 1997 l
U.S. Nucle:r Regul: tory Commission B16636%ttachment 3\\Page 1 RETYPE OF PROPOSED REVISION Refer to the attached retype of the proposed revision to the Technical Specifications.
The attached retype reflects the currently issued version of the Technical Specifications. Pending Technical Specification revisions or Technical Specification revisions issued subsequent to this submittal are not reflected in the enclosed retype.
The enclosed retype should be checked for continuity with Technical Specifications prior to issuance.
A matrix of pages is provided to assist the reviewer.
I i
Page #
Exis'Jng Amendment Pending Submittal 3/4 1-13 60 3/4 1-14 122 3/4 1-15 60 B16406 i
3/4 1-16 60 B16406 3/4 4-3 7
3/4 4-5 Base 3/4 4 Sa overflow 3/4 4-6 99 3/4 4-6a overflow 3/4 4-8 60 3/4 4-33 Base 3/4 4-34 60 3/4 4-35 60 3/4 4-36 48 3/4 4-37 Base 3/4 4-38 143 3/4 4-38a overflow 3/4 4-39 133 3/4 4-40 60 N4441 60 3/4 5-7 Base 3/4 5-8 Base B 3/4 1-2 50 B 3/4 1-3 113 B16447 8 3/4 4-1 99 B 3/4 4-1a 60 B 3/4 4-7 Base B 3/4 4-8 48 B 3/4 4 9 Base B 3/4 4-10 Base
U.S. Nuclear Regulatory Commission B16636%ttachment 3\\Page 2 i
Page #
Existing Amendment Pending Submittal B 3/4 4-11 89 83/44-12 48 I
B 3/4 413 48 l
83/44-14 88 B16246 B 3/4 4-15 138 B16246 8 3/4 4 16 88 83/4417 New B 3/4 4-18 New B 3/4 4-19 New B 3/4 4 20 New B 3/4 4-21 New B 3/4 4-22 New l
B 3/4 4 23 New j
B 3/4 4-24 New B 3/4 4-25 New B 3/4 4-26 New B 3/4 4-27 138 B 3/4 5-1 Base B 3/4 5-2 100
REACTIVITY CONTROL SYSTEMS 3/4.1.2 'B0 RATION SY$TEMS=
FLOW PATH - SHUTDOWN LIMITING COMITION FOR OPERATION
)
3.1.2.1 As a minimum, one of the following boron injection flow paths shall be OPERABLE and capable of being powered from an OPERABLE emergency power source:.
- A flow path from the boric acid storage system via either a boric a.
acid transfer pump or a gravity feed connection and a charging pump to the Reactor Coolant System if the boric acid storage
< system in Specification 3.1.2.5a. is OPERABLE, or b.
The flow path from the refueling water storage tank via a charging pus) to the-Reactor Coolant System if the refueling water storage tant in Specification 3.1.2.5b. is OPERABLE.
APPLICABILLT1: MODES 4, 5, and 6.
l ACTION:
With none of the above boron injection flow paths OPERABLE or a.
capable of being powered from an OPERABLE emergency power source in MODE 4, provide an OPERABLE flow path capable of being powered from -
an.0PERABLE emergency power source within I hour or be in-COLD SHUTDOWN within the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , b.-
With none of the above boron injection flow paths OPERABLE or capable of being powered from an OPERABLE emergency power source in MODES 5 or 6,' suspend all operations involving CORE ALTERATIONS or positive reactivity changes.
SURVEILLANCE REQUIREMENTS-4.1.2.1 At least.one of the above required flow paths shall be demonstrated OPERABLE:
a.
. At least once per 7 days by verifying that-the Boric Acid Transfer
' Pump _ Room temperature and the. boric acid storage tank solution temperature are greater than or equal-to 67'F when a flow path from the boric acid tanks is used, and b.-
At least once'per 31 days by verifying that each valve (manual, power-operated, or automatic) in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct position.-
~~MILLSTONE - UNIT-3 3/4 1-13 Amendment No. pp 0534~~
REACTIVITY CONTROL SYSTEMS FLOW PATlis - OPERATING LINITING CONDITION FOR OPERATION 3.1.2.2 At least two of the following three boron injection flow paths l
shall be OPERABLE:
a.
The flow path from the boric acid storage system via a boric acid transfer pump and a charging pump to the Reactor Coolant System (RCS),and b.
Two flow paths from the refueling water storage tank via charging pumps to the RCS.
APPLICABILIJJ: MODES 1, 2, and 3.
l ACTION:
l With only one of the above required boron injection flow paths to the RCS OPERABLE, restore at least two boron injection flow paths to the RCS to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY and borated to a SHUTDOWN MARGIN equivalent to at least the limits as shown in figure 3.1-4 at 200'F within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months restore at least two flow paths to OPERABLE status within the next 7 days or be in HOT SHUTDOWN within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .
$URVEILLANCE REQUIREMENTS 4.1.2.2 At least two of the above required flow paths shall be demonstrated OPERABLE:
a.
At least once per 7 days by verifying that the Boric Acid Transfer Pump Room temperature and the boric acid storage tank solutioa temperature are greater than or eoual to 67'F when it is a required water source; b.
At least once per 31 days by verifyf og that each valve (manual, power-operated, or automatic) in the P ow path that is not locked, sealed, or otherwise secured in position, is in its correct
~~position, c.~~
At least once each REFUELING INTERVAL by verifying that each automatic valve in the flow path actuates to its correct position on a Safety injection test signal; and d.
At least once each REFUELING INTERVAL by verifying that the flow path required by Specification 3.1.2.2a. delivers at least 33 gpm to the RCS.
]
1 NILLSTONE - UNIT 3 3/4 1-14 Amendment No. JP, pp, J77, 0624
~. -. -
REACTIVITY CONTROL $YSTEMS CHARGING PUNP - SHUTDO)Gi LINITING CONDITION FOR OPERATION 3.1.2.3 One charging pump in the baron injection flow path required by Specification 3.1.2.1 shall be OPEPABLE and capable of being powered from an OPERABLE emergency power source.
APPLICABillTY: H0 DES 4, 5, and 6.
l ACTION:
With no charging pump OPERABLE or capable of being powered from an a.
OPERABLE emergency power source in H0DE 4, provide an OPERABLE charging pump ca)able of being powered from an OPERABLE emergency power source wit 1in I hour or be in COLD SHUTDOWN within the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , b.
With no charging pump OPERABLE or capable of being powered from an OPERABLE emergency power source in H0 DES S and 6 suspend all o)erations involving CORE ALTERATIONS or positive reactivity c1anges.
SURVEILLANCE REQUIRENENTS 4.1.2.3.1 The above required charging pump shall be demonstrated OPERABLE by verifying, on recirculation flow, that a differential 3ressure across the pump of greater than or equal to 2411 psid is developed w1en tested pursuant to Specification 4.0.5.
NILLSTONE - UNIT 3 3/4 1-15 Amendment No. 77 79.
% 34
REACTIVITY CONTROL SYSTQii GingGlHG PUNPS - OPERATING LIMITING CONDITION FOR OPERATION 3.1.2.4 At least two charging pumps shall be OPERABLE.
l APPLICABILITY: MODES 1, 2, and 3.
l ACTION:
i With only one charging pum) OPERABLE, restore at least two charging pumps to OPERABLE status within 72 Tours or be in at least HOT STANDBY and borated to a SHUTDOWN MARGIN equivalent to at least the limit as shown in figure 3.1-4 at 200'F within the next 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months st restore at least two charging
) umps to OPERABLE status within the next 7 days or be in HOT SHUTDOWN wit 11n the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .
SURVEILLANCE REQUIRENENTS 4.1.2.4.1 At least two charging pumps shall be demonstrated OPERABLE by verifying, on recirculation flow, that a differentiai pressure across each pump of greater than or equal to 2411 psid is developed when tested pursuant to Specification 4.0.5.
MILL STONE - 11 NIT 3 3/4 1-16 Amendment No. JP, 0624
REACTOR COOLANT SYSTEM HQLJ1NIDOWN LIMITING CONDITION FOR OPERATION 3.4.1.3 Either:*,**
a.
With Reactor Trip System breakers closed, at least two RCS loops shall be OPERABLE and in operation, or b.
With Reactor Trip System breakers open, at least two loops consisting of any combination of RCS loops and residual heat removal (RHR) loops ***
shall be OPERABLE, and at least one of these loops shall be in operation.
For RCS loop (s) to be OPERABLE, at least one reactor coolant pump (RCP) shall be in operation.
APPLICABILITY: MODE 4.
ACTION:
a.
With less than the above required loops OPERABLE, immediately initiate corrective action to return the required loops to OPERABLE status as soon as possible; if the remaining OPERABLE loop is an RHR loop, be in COLD SHVTDOWN within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , b.
With no loop in operation, suspend all operations involving a reduction in boron concentration of the Reactor Coolant System and immediately initiate corrective action to return the required loop to operation.
~~All reactor coolant pumps and RHR pumps may be deenergized for up to I hour provided:~~
~~(1) no operations are permitted that would cause dilution of the Reactor Coolant System boron concentration, and (2) core outlet temperature is maintained at least 10*F below saturation temperature.~~
~~A reactor coolant pump (RCP) shall not be started unless one of the following conditions is met:~~
a.
At least one RCP is operating.
b.
The secondary side water temperature of each steam generator, not isolated from the RCS, is less than or equal to the lowest RCS wide range cold leg tem)erature of the unisolated RCS loops.
c.
Witi a maximum of one RCS loop isolated and with the RHR relief valves isolated from the RCS, the secondary side water temperature of each steam generator, not isolated from the RCS, is less than er equal to 250'F.
d.
All RCS wide range cold leg temperatures >275'F and no cold overpressure protection relief valves are in service as follows:
1)
COPPS is blocked or the PORV block valves are closed, and 2)
RHR relief valves are isolated from the RCS (3RHS*HV8701C or 3RHS*HV8701A is closed and 3RHS*HV8702B or 3RHS*MV8702C is closed).
~~Prior to opening 3RHS*HV8701C and 3RHS*HV8701A, or 3RHS*MV8702B and 3RHS*MV8702C, all safety injection pumps and all but one centrifugal charging pump shall be incapable of injecting into the RCS.~~
Surveillance Requirements 4.4.9.3.4 and 4.4.9.3.5 apply whenever any RHR relief valve is unisolated from the RCS.
MILLSTONE - UNIT 3 3/4 4-3 Amendment No. 7.
0625
KACTOR COOLANT SYSTEM COLD $HUTDOWN - LOOPS FI RER LIMITING CONDIY10N FOR OPERATION 3.4.1.4.1 At least one residual heat removal (RHR) loop shall be OPERABLE and i
in operation *, and either:
1 a.
One additional RHR loop shall be OPERABLE **, or l
b.
The secondary side water level of at least two steam generators i
shall be greater than 17%.
i APPLICABILITY:
MODE 5 with at least two reactor coolant loops filled ***.
a.
The RHR pump may be deenergized for up to I hour provided:
(1) no operations I are permitted that would cause dilution of the Reactor Coolant System boron concentration, and (2) core outlet temperature is maintained at least 10*F below saturation temperature,
b. All RHR loops may be removed from operation during a planned heatup to MODE 4 when at least one RCS loop is OPERABLE and in operation and when two additional steam generators are OPERABLE as required by LCO 3.4.1.4.1.b.
~~0ne RHR loop may be inoperable for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for surveillance testing provided l~~
~~the other RHR loop is OPERABLE and in operation.~~
a.
No reactor coolant pum)s (RCPs) may be in operation below 160'F unless COPPS is blocked or unless tie PORV block valves are closed.
b.
An RCP shall not be started unless one of the following conditions is met:
~~1. At least one RCP is operating and the lowest RCS wide range cold leg tem)erature of the unisolated RCS loops is >l60*F.~~
2. Witi two or more Reactor Coolant System (RCS) loops isolated, the first RCP shall not be started unless the secondary side water temperature of each steam generator not isolated from the RCS is less than or equal to the lowest RCS wide range cold leg teniperature of the unisolated RCS loo)s.
3. Witi a maximum of one RCS loop isolated, with the RHR relief valves isolated from the RCS, and with the PORVs providing cold overpressure protection, the first RCP shall not be started until the secondary side water temperature of each steam generator not isolated from the RCS is less than or equal to 50'F above the lowest RCS wide range cold leg temperature of the unisolated RCS loops.
4. With a maximum of one RCS loop isolated and with any RHR relief valve unisolated from the RCS, the first RCP shall not be started until the secondary side water temperature of each steam generator not isolated from the RCS is less than or equal to 200*F and less than or equal to 50'F above the lowest RCS wide range cold leg temperature of the unisolated RCS loops.
MILLSTONE - UNIT 3 3/4 4-5 Amendment No.
0526
REACTOR C0OLANT SYSTE8 COLD SHUTDOW - LOOPS FILLE _D LIMITING CONDITION FOR OPERATION ACTION:
a.
With less than the required RHR loop (s) OPERABLE or with less than the l required steam generator water level, immediately initiate corrective l
action to return the inoperable RHR loop to OPERABLE status or restore the required steam generator water level as soon as possible.
b.
With no RHR loop in operation, suspend all operations involving a reduction in boron concentration of the Reactor Coolant System and immediately initiate corrective action to return the required RHR loop to operation.
$URVEILLANCE REQUIREMENTS l
l 4.4.1.4.1.1 The secondary side water level of at least two steam generators when required shall be determined to be within limits at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .
4.4.1.4.1.2 At least one RHR loop shall be determined to be in operation and circulating reactor coolant at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .
i MILLSTONE - UNIT 3 3/4 4-Sa Amendment No.
0626
REACTOR COOLANT SYSTEN COLD SMTDOW - LOOPS N T FILLED LIMITING COWITION FOR OPERATION 3.4.1.4.2 no residual heat removal (RHR) loops shall be OPERABLE
and at least one RHR loop shall be in operation.**
APPLICABILITY: MODE 5 with less than two reactor coolant loops filled ***.
l ACTION:
a.
With less than the above required RHR loops OPERABLE, immediately initiate corrective action to return the required RHR loops to OPERABLE status as soon as possible.
b.
With no RHR loop in operation suspend all operations involving a reductioninboronconcentratIonoftheReactorCoolantSystemand
-immediately initiate corrective action to return the required RHR-loop to operation.
f'
0ne RHR loop may be inoperable for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for surveillance testing provided the other RHR loop is OPERABLE and in operation.
~~The RHR pump may be doenergized for up to I hour provided:~~
~~(1)noopera-tions are permitted that would cause dilution of the Reactor Coolant System boron concentration, and (2) core outlet temperature is maintained at least 10'F below saturation temperature.~~
a.
No reactor coolant pumps (RCPs) may be ih operation below 160'F unless COPPS is blocked or unless the PORV block valves are closed.
b.
An RCP shall not be started unless one of the following conditions is met:
1.
At least one RCP is operating and the lowest RCS wide range cold leg tem >erature of the unisolated RCS loops is >160'F.
2.
Witi two or more Reactor Coolant System (RCS) loops isolated, the first RCP shall not be started unless the secondary side water temperature of each steam generator not isolated from the RCS is less than or equal to the lowest RCS wide range cold leg temperature of the unisolated RCS loops.
3.
Witi a maximum of one-RCS loop isolated, with the RHR relief valves isolated from the RCS, and with the PORVs providing cold overpressure protection, the first RCP shall not be started until the secondary side water temperature of each steam generator not isolated from the RCS is less than or equal to 50'F above the lowest RCS wide range cold leg tem>erature of the unisolated RCS loops.
4.
Wit l) a maximum of one RCS loop isolated and with any RHR relief valve unisolated from the RCS, the first RCP shall not be started until the secnndary side water temperature of each steam generator not isolated from the RCS is less than or equal to 200*F and less than or equal to 50'F above the lowest RCS wide range cold leg temperature of the unisolated RCS loops.
NILLSTONE - UNIT 3 3/44-6 Amendment No. pp 77, nn
REACTOR COOLANT SYSTEM GQLD SHLITD0ldN - LOOP 5 NOT FILLED i
SURVEILLANCE REQUIRENENTS 4.4.1.4.2.1 The required RHR loops shall be demonstrated OPERABLE pursuant to l
Specification 4.0.5.
l 4.4.1.4.2.2 At least one RHR loop shall be determined to be in operation and circulating reactor coolant at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .
I i
)
l N!jlSTONE-UNIT 3 3/44-6a Amendment No. pp, 77,
REACTOR COOLANT $Y$1EN ISOLATED LOOP STARTUP LIMITING COM ITION FOR OPERATION 3.4.1.6 A reactor coolant loop shall remain isolated with power removed from the associated RCS loop stop valve operators until:
a.
The temperature at the cold leg of the isolated loop is within 20'F of the highest cold leg temperature of the operating loops, b.
The boron concentration of the isolated loop is greater than or equal to the boron concentration of the operating loops, or greater than 2600 ppm whichever is less, c.
All reactor coolant pumps are de energized.
l d.
The isolated portion of the loop has been drained and is refilled, and l
The reactor is subcritical by at least the value required bySpecifications3 e.
for Mode 6.
APPLICABILITY: MODES 5 and 6.
ACTION:
a.
With the requirements of the above specification not satisfied, do not open the isolated loop stop valves.
$URVEILLANCE REQUIREMENTS 4.4.1.6.1 The isolated loop cold leg temperature shall be determined to be within i
20'F of the highest cold leg temperature of the operating loops within 30 minutes prior to opening the cold leg stop valve.
4.4.1.6.2 The reactor shall be determined to be subcritical by at least the value required by Specifications 3.1.1.1.2 or 3.1.1.2 for Mode 5 or Specification 3.9.1.1 for Mode 6 within 30 minutes prior to opening the cold leg stop valve.
4.4.1.6.3 Within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months prior to opening the loop stop valves, the isolated loop shall be determined to:
a.
Be drained and refilled, and b.
Have a boron concentration greater than or equal to the boron concentration of the operating loops, or greater than 2600 ppm whichever is less.
MILLSTONE - UNIT 3 3/44-8 Amendment No. 17,77,77 0625
REACTOR COOLANT $YSTEN jf4.4.9 PRES $URE/ TEMPERATURE Linill LINITINR ColWITION FOR OPERATION P
3.4.9.1 The reactor coolant system (except the pressurizer) temperature and pressure shall be limited as follows:
a.
During an RCS heatup, the heatup limits of Figure 3.4 2 apply with the additional restriction that only one' reactor coolant pump can be o
operating when the lowest unisolated RCS loop wide range cold leg temperature is sl60'F.
b.
During an RCS cooldown, the limits of Figure 3.4 3 apply with the additional restriction that only one reactor coolant pump can be operating when the lowest unisolated RCS loop wide range cold leg temperature is sl60'F and no reactor coolant pump may be operated when the lowest unisolated RCS loop wide range cold leg temperature is $120'F.
c.
During steady state conditions, when the maximum temperature increase or decrease in any one hour period is <10'F and when the plant is not changing temperatures in accordance with a heatu) or cooldown procedure, only one reactor coolant pump can be operating w1en the lowest unisolated RCS loop wide range cold leg temperature is sl60'F.
The limits of Figures 3.4 2 and 3.4 3 do not apply during steady state conditions, d.
During RCS inservice leak and hydrostatic testing operations, the Hydrostatic and Leak Test limit of Figure 3.4 2 apply with the additional restrictions that within a one hour period prior to exceeding the heatup curve, and during each one hour period above the heatup curve, a maximum temperature increase or decrease of 5'F in any one hour period is allowed.
APPLICABillH: At all times.
E110N:
With any of the above limits exceeded, restore the temperature and/or pressure to within the limit within 30 minutes; perform an engineering evaluation to determine the effects of the out-of-limit condition on the structural integrity of the Reactor Coolant System; determine that the Reactor Coolant System remains acceptable for continued operation within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY l within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and reduce the RCS T and pressure to less than 200'F and500 psia,respectively,withinthefolloEfng30 hours.
4 SURVEILLANCE REQUIRENENTS 4.4.9.1.1 The Reactor Coolant System temperature and pressure shall be determined to be within the limits at least once per 30 minutes during system heatup and cooldown operations, and during the one hour period prior to and j'-
during inservice leak and hydrostatic testing operations.
4.4.9.1.2 The reactor vessel material irradiation surveillance specimens i
shall be removed and examined, to determine changes in material properties, as required by 10 CFR Part 50, Ap>endix H, in accordance with the schedule in Table 4.4-5.
The results of tiese examinations shall be used to update Figures 3.4-2 and 3.4-3 as required.
l 1
NILLSTONE - UNIT 3 3/44-33 Amendment No.
0626
-. ~.
1 I
i Millstone 3 Reactor Coolant System i
l l
Heatup Limits for up to 10 EFPY 2500 i
i i
i
/
i i
...v.......4.....l...
....4................:...o I
.4.....
.....,,...4..
.... 4.. g..
i t
i
.....j..'
...i....
.....!1..... '...:...... I...
l I
i i
i 1
/
- "t"
""t-
......!...........l,.....,:. Lea k Test Limit "'
- "t"" " " t ""
""I" -
..........,......,....4...........
.... I i
2000 l
4
{
{
{
' l:
I i
i i
1 i
i i
i i
i 1
.....[:...
.....:r.....:...
. ]. 3.....
....i..
r Heatup At a Maximum of 40'F in Any i
l :-
t 1 Hour Period Up To 160*F, and
-t-
- - r -- --[ -
- :/ ~ ~i --
At a Maximum of 80'F in Any 1 Hour i
..4......f....4......
..v..
....4.....
Period above 160*F.
i i.
(
... ;..... j..../..
i.
7
.. ;....;.. ).. ;
...;....4..
~~....j....~~
,j i
i i
i i
I fi i
1500 S
i I
/
e E
...................,i........,i... HeatuP i
i
......i.....
...i......,!.... !
/
i i
Limit i
i i
....,i..........:.
i
,......:... A.....i,.....
,.../.,.......,.....,i.....
i t
l i-
......g.....,.
....4
.........o p...4...........
i
/
i i
i Unacceptable i
.../.>.....<......g.....<.....
........ l M
f; w
1000 i
Onerat. ion i
i.
/ i.
i.
g-
..e.....n..4
...4....4..
.....g..../....4.....,
i
/
/
.....J.....:....4.....
.....I...
....a
....4....
I i
}
I
...4.........
.. '........4....
..4...
......j.... ;,..
Criticality Limit ll i
i i
i.
...,...........,....4....
!/
500 l l-
..>.....<......g.....
l :>...
i
{
g
.........4.....,...4.....
..4......t....4....4.....+
.........4
.........4....
.... t.... 4. o. 4.....
2 l
l l
...4....
..4....4.....
....4....4.....J......
.....:....4.....:...4.....
...4o...J...4......-...
j l
1 1:
i
.... i,!
.....t..........
i l:
0 O
100 200 300 400 Indicated Cold Leg Temperature (*F)
FIGURE 3.4 2 N!!.l. STONE - UNIT 3 3/44-34 Amendment No. 77.
Cf26
Millstone 3 Reactor Coolant System Cooldown Limitations for up to 10 EFPY 2500 1
i i
i i
i i
f i
i i
I 1
...4.....4.........4......a.....4.....'.....,
1 1
t t
t l
t l
....4.........4.....
...4.....j....4.....4......
t t
1 I
i l
l
..... j......{......i..... a.....
,..... j.....!......!....
I I
l
......I.....
.....7....i 7
..... 7...
I i
i i
i i
i i
.....I...... l..... 7..... l....
.....1.....1.....g.....
......i..........t...
.....;......t...........!...
I t
I 1
1 1
1 i
...1.....j......
....1.....1.....1.....
.....1........1..
...1......................
....7.....
2000 l
1 i
l t
1 i
l I
.....t.....
i....q.....
I
,i
~~......i.....~~
l......
.....p.....<
.....g.....g....
j i
1 1
1
?
....t......!.,..<i.......
......I....
p i
...4.....<.....I.....e.....
i
,,.........<....4.....<.....
i p.....
s....
1 i
1 8
I 1
t 8
}
l I
......!.. Cooldown Limit
}
l
....4....
.4....
..4....4.....
,...4....
1 i
1 i
i i
I 3
i
......j....
....g...,
. 4'....
. 9....
..4.....j..... i..... g!......
I 1
1 l
1 I
j 1
1 1
1
}
}
l j
1500 C1, V
i i
i i
m
.....A..........A..........
....L...
....A..
.....J....
....a.....<
C 1
D l
i f
l i
}
.....J....
.....J
...4....
i....<
I i
i i
Unaccentable
......i....i i
g-3
.. u 4..... q. u u q u. o (m.
l t
g O erat 10n
.t.... <.........,....,.....!..
.!..... i...... !..... i......
.....)......
P I
t 1000 i
i l
i I
......:......l.....
....4.....
.....4.....d....4......
4
}
...4....
....4....4 i
I I
1 l
..........I..
..?
......I..........t....
...u.!........u.'....
7 I
i l
l t
t j
i j
,... J... 1..
..;........A,..2'....
i i
i i
Cooldown At a Maximum of
......!.........t...
80'F in Any 1 He'ur Period to 2
~
160'F. Then At a Maximum of 500 20'F in Any 1 Hour Period
! /
i t
Below 160 F
......... j...... j......j..'.
.....j..... j......j..... j.....
,.....t
! f gy;..... ;.........
j p
7
......;..... u......;......,
Cooldown At a Maximum of l
i i
j i
80'F in Any 1 Hour Period to s.
.....j..... 3..........j............j......
160'F, Then At a Maximum of
......i............i.......i.....
......!.....,.....i..........
40'F in Any 1 Hour Period i
i i
! Below 160'F
?,
i, i
g 0
100 200 300 400 Indicated Cold Leg Temperature ('F)
Figure 3.4 3 NILLSTONE - UNIT 3 3/4 4-35 Amendment No. pp.
0626
l TABLE 4.4-5 gz xh REACTOR VESSEL MTERIAL SURVEILLANCE PROGRAM - WITunRAlpi $C!fEDULE E
m CAPSULE VESSEL LEAD APPROXIMTE l
NUMBER LOCATION FACTOR WIT}9 RAMI TlE (EFPY) g U
58.5*
3.98(a)
First Refueling (1.3 EFPY actual)
)
w Y
241' 3.74 9
Y 61*
3.74 16 W
121.5*
4.01 STAND 6Y X
238.5*
4.01 STAND 8Y R.
Z 301.5*
4.01 STAND 8Y M
a)
Plant specific evaluation.
f a
.F
REACTOR COOLANT SYSTEM PftE550R1ZIR LINITING CONDITION FOR OPERATION 3.4.9.2 The pressurizer temperature shall be limited to:
a.
A maximum heatup of 100'F in any 1-hour period, b.
A maximum cooldown of 200*F in any 1. hour period, and c.
A maximum sprty water temperature differential of 320'F.
APPLICABILITY: At all times.
ACTION:
With the pressurizer temperature limits in excess of any of the above limits, restore the temperature to within the limits within 30 minutest perform an engineering evaluation to determine the effects of the out-of-limit condition on the structural integrity of the pressurizert determine that the pressurizer remairs acceptable for continued operation within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least l
HOT STANDBY within the next 6 hurs and reduce the pressurizer pressure to i
less than 500 psia within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .
I SURVEILLANCE REQUIREMENTS 4.4.9.2 The pressurizer temperatures shall be determined to be within the limits at least once per 30 minutes during system heatup or cooldown. The spray water temperature differential shall be determined to be within the limit at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months during auxiliary spray operation.
NILL 5 TONE - UNIT 3 3/4 4-37 Amendment No.
0624
.m g qu-r
l REACTOR COOLANT SYSTEM OVERPRES$URE PROTECTION SYSTEMS LIMITING CONDITION FOR OPERATION 3.4.9.3 Cold Overpressure Protection shall be OPERABLE with a maximum of one centrifugal charging pump
~~and no Safety injection pump:. capable of injecting into the Reactor Coolant System (RCS) and one of the following pressure relief capabilities:~~
1.
One power operated relief valve (PORV) with a nominal lift setting established in Figure 3.4 4a and one PORV with a nominal lift setting established in Figure 3.4-4b, er 2.
Two residual heat removal (RHR) suction relief valves with setpoints 1 426.8 psig and 1 453.2 psig, or 3.
One PORV with a nominal lift setting established in Figure 3.4-4a or Figure 3.4 4b and one RHR suction relief valve with a setpoint 1 426.8 psig and 5 453.2 psig, or 4.-
RCS depressurized with an RCS vent of 2 5.4 square inches.
APPLICABILITY: MODE 4 when any RCS cold leg temperature is s 275'F**, MODE 5, and MODE 6 when the head is on the reactor vessel.
ACTION:
a.
With two or more centrifugal charging pumps capable of injecting into the RCS, immediately initiate action to establish that a maximum of one centrifugal-charging pump is capable of injecting into the RCS.
b.
With any Safety Injection pump capable of injecting into the RCS, immediately initiate action to establish that no Safety Injection pumps are capable of injecting into the RCS.
c.
With one required relief valve inoperable in MODE 4, restore the required relief valve to OPERABLE status within 7 days, or depressurize and vent the RCS through at least a 5.4 square inch vent within the next 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .
Two centrifugal charging pumps may be capable of injecting into the RCS for less than one hour, during pump swap operations. However, at no time will two charging pumps be simultaneously out of pull-to lock during pump swap operations.
When an RHR suction relief valve or a PORV, which is armed for COPPS, is unisolated from the RCS, and when the RCS cold leg temperature is greater than 275'F, a maximum of one centrifugal charging pump and no safety injection pumps shall be capable of injecting into the RCS.
M!gLSTONE-UNIT 3 3/4 4-38 Amendment No. Jp, 77, pp #7,
REACTOR COOLANT SYSTEN OVERPRES$URE PllQIEG110.. SYSTEMS LIMITING CONDITION FOR OPERATION d.
With one required relief valve inoperable in MODE 5 or 6, restore the required relief valve to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , or depressurize the RCS and establish an RCS vent of 15.4 square inches within the next 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , e.
With two required relief valves inoperable and with no RCS vent 15.4 square inches, dearessurize the RCS and establish an RCS vent of 1 5.4 square inches wit 11n 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , f.
In the event the PORVs, the RNR suction relief valves, or the RCS vent are used to mitigate an RCS pressure transient, a Special Report shall be prepared and sutaitted to the Commission pursuant to Specification 6.9.2 within 30 days.
The report shall describe the circumstances initiating the transient, the effect of the PORVs, the RHR suction relief valves, or RCS vent on the transient, and any corrective action l
necessary to prevent recurrence.
g.
Entry into an OPERATIONAL MODE is permitted while subject to these l
ACTION requirements.
NILLSTONE - UNIT 3 3/4 4-38a Amendment No. J7, 77, M. #7 osa
REACTOR _GQ0LST.SYSTElf DVERPRES$URE PROTECTION SYSTEM SURVEILLANCE REQUIREMENTS 4.4.9.3.1 Demonstrate that each required PORV is OPERABLE by:
l a.
Performance of an ANALOG CHANNEL OPERATIONAL TEST on the PORV actuation channel, but excluding valve operation, within 31 days prior to entering a condition in which the PORY is required OPERABLE and at least once per 31 days thereafter when the PORY is required OPERABLE; b.
Performance of a CHANNEL CAllBRATION on the PORV actuation channel at least once each REFUELING INTERVAlt and c.
Verifying the PORV block valve is open and the PORY Cold Overpressure Protection System (COPPS) is armed at least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months when the PORY is being used for overpressure protection.
4.4.9.3.2 Demonstrate that each required RHR suction relief valve is OPERABLE by:
a.
Verifying the isolation valves between the RCS and each required RHR suction relief valve are open at least once per 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months st and b.
Testing pursuant to Specification 4.0.5.
4.4.9.3.3 When com)1ying with 3.4.9.3.4, verify that the RCS is vented through a-vent pat 1way 1 5.4 scuare inches at least once per 31 days for a passive vent path anc at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months for unlocked open vent valves.
4.4.9.3.4 Verify that no Safety Injection pumps are capable of injecting into the RCS at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .
4.4.9.3.5 Verify that a maximum of one centrifugal charging pump is capable of injecting into the RCS at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .
MILLSTONE - UNIT 3 3/44-39 Amendment No. 77, pp, JPP, J77, osa
HICH $ETPOINT PCRV Cunvc FOR THE COLD OvanPatssunt PROTECTION SYSTEM 800
........<....s......,.
i.
1 I
.....t....g..
..o
...g.........
i i
i o
i 2
...a.
.a....$...
.. 4... 4,.... e..... a.
...4......
..4....a,,,..,
.u.
f I
i t
t i
1 i
..<....4......p....1
.. t......:.... 1 4....4.....,,.........
....t.........,
4....
700
....,i..... t...
662 psia
..i.....!,....,..
i i
t
....._....,i....
i
......e n.
u....n
..f.
1
~~4. +.. 4. n..t...<~~
~~u..........4......d...n u n d..... 4..... 4..... &....u. 1...~~
~~4..... a. n. 4,...~~
g o.. a.. o. 4.... 4....
n..
4
...1 3
n.q...q.....p.
.....p...>..
n<....q.....<..,.g.,
...q....g.o,q....q.,.
...q.,..po..q.,.q...
g 600
....1............l....
4
..J......!....
...s.....l......
~
2 4.... a... d.... a..
4... 4.,.~.
~~a....~~
.... /.... l... 4.. o j...
...4.....J.n.4,..8....
....t.o.4.....$...a....
1 1
1 1
?
i 4'
1 i.
~
..}n.q.,.a.....)..
..p.
.. 4... 4 n... (..... q....
..q..og...nb...qo
..q.,..)..4...qn
.q....."
en 8%
i.
'.....j 2
o.....g.....q...9...
,....g....p..
E 500 pn..p...p.
j
~
....;n ngo...,n
~
n9 4
... l.. o..
..ni....4..
.. o d. o m
.. 4...
...e...
J... 4.... n d....
4
. 4,....... i. n M (g O aI5 n;... +....;..
. n;...4......<...4....
... q.
4. p.... p... a. n.
og....p....)..
..p..
J B W
l
. a. n a
i.
t l
l i
1 oq.,
q..oq.
n. -.
s.
i 7..
n.p...
g
...p..
... q,.... g. n.. g o n. t. n.
.9n ogo.qn op
..I.
..'o
~~.............I..u~~
.q..
W 400
. n M
n.
..o l
l
't
.d...
4.n. 4.....b
. o.
. 4...
~~4.. o.. }.... i n~~
~~4...... } n. d... e d....~~
..4....<.
=
..<....q...
uq..q......p.q..
.... p.... g. n o g. n n t. o n ngo ngonqn
.s
~~i.. >....-..~~
...o J
1 l
1 1
.'.n..
i
... d.. o
,. a..,. 4...
....4..4....4....
.....4..
.. J... o 4
.. d. n.
4...
n aqn
}
l
...1...
2
,q.
.q...q.
nq...
.n<.....q..
c.
..q..
... t..... t..,. n t o... p...
...q...q...qo..on.
1 1
1
......'n I
t 1
.e.....1...
.. 1.... $.... q.... g..
1 i
2 n o; n ogn...
.q.ng.
4..
....gno ungn
..... ' -... '......1...
o;n r.o.
..$.n..'..
...!....'n n o..
....o...
. 4 n.. a...... l..
~~4..... 4.~~
~~4...~~
~~4...~~
...4..
. 4... o 4.. n.
. i.... 4... o. l.... n (..
4.. u.
200
..'.".."".'t.".'t."*
.o.'
.q......;....q.
.q.
..t.'
t.
..n l " t ' " " f. " '
....o.
.....u..
...n
..o...
.o 3
........4........J.o n. d n..d.. n d. 4 " +
o p... 4.... d... n 4 u..
...d....a..oan
.,.n.
.. n:. n...:...
I 1
1
.qu...p....p 4,.
.n..p.
o...+ n..
..<....<...ng....t...
..po..p...q.....p..
.. q n.. q. o o p...q o 1
I
{
l I
i l
100
.nn...n
.......,........n
...o o
n+....
...n..
.u.
.e
.n.o.n n.o o 3
t.
1
.d.ood.ood."
~~4...... l... 4.... 1.......4....4....4.,.....;....~~
. 4..... 4..... 4.. n.
. o J. o. o.. n. 1. o.. d... n u..
n.)" ".>n.
n o 4. n
~~4..... q. u. 4,.. n 4. n.~~
..4.n...:......>....+...
..4.....4..u..<..
.. {. o o t. o... p.. n l....
I 1
l l
1 1
... t. o.. 1
... q n... p.. 9.... 9.. n....np....
.... g.
.. l,n... q..... t....ot.o gn.
. " " t " "
~~t " ' 9. " " t " '~~
0 50 100 150 200 250 300 AUCTIONEERED LOW MEASURED RCS TEMPERATURE (*F)
FIGURE 3.4-4a NILLSTONE - UNIT 3 3/44-40 Amendment No, p.
0626
Low SETPOINT PORY CURVE FOR THE COLo OvEnentssunt Pn0TECTION SYSTEM 800 3
i
...n)..
.qn.q.u...p...
...p. o y o.
p...A
.u.t.......gu en'..".t......<....p..
...q
...p...q..q,.,q.,
....f..............
I 1
...o............
....l....lo.4...
.4...4o.4..
J..........oi....
.. 1.... J.... ;..
.o.
a..a.
u..
1 i.
i I
i.
3 I
i.
..q..,q.oq...q.
4...4...ud..........
,.. 4.. u. 4,.....,;o..
...&.......<....4....
un<.
.4,....q.,q....
700 1
e
.a n......o
.. 9... 9. u.. q.,. 9 o y
o 9..... g... o i. o
......g..,q.,
1 t.
l
.1
.o.....
n..
n..
o.
..o..
......o.......
...n
.... o...
!o
...u...
.u.
.u
. o d...
~~4..... d.... 4..~~
..4..
4.......
.. 4..... i. n
~~.. o. a.... a...~~
....q.....p. 4 4.... q...
o.)......
= a yo.
... f..
..p....p....<..
600
..g....q..
3g p...q....q..
]
o.
......o...n
.o
....1.......
.u
.n
,.....n
.oa..4.....
.1....
.,u
.,..4.
J. 4... 4.... J...
o. 4.... 1..... t.... J.,.
ut....d....d... 4...
e c.... o n o o. o p.
.og....4...t..,.g..
.4...q.,..q..
..)..
i
.....)u.q...q.
..). o
.q.
.q...
4
~
p.s
..q.o.q.,.,p...q..
...p...
.o e
op.
a,. 500
. q. u. q.... t. -
~~t. " " t..~~
+q"*"*"
g
...,..... a..... a.... a...
....4...
.4...
. a..... a.... a..... a..
hM
.. j.... 6.. n i.o.
.. j.... h.... j.... j..
n.i....i....i...'..
a F
. n o p
.4.*
.?
i
.qo
.o..
.q...q....
...q.
+
o,.
.p...pu.
. " t. " " t., " " ' t. " " t. "
" l " " f.. '
~~t " " 1. ".~~
(
m a
o.. o
.. 1 n
...o
.o..
.... o...
,M l
t
..g....
~~400 i~~
~~., 4..~~
~~4..... a....~~
6 ot.,
...<o
. 4..o<..
.4...
4
.. o )... o n.. 0...
...(....4......t,u.
~~4. o 1~~
...p.....,.
...t..
.n.
,3...
..<....p....<..
... t..... t. o. u t... o t. n
~~nt.u o3..op...oto s.~~
g
....po........
....p....;
... o n
.u.
.on
-u
........... o
.......q..o
.go 4
. 4... o a... 4..
. 4..
.4...
...4......,...
...J
.n.......
Z n
.o
.. d. o. 4.. o
.. i n u
300 t
.i l
n...p..qu.q...q...
V
..qo
.q.,.,q..,q..
s.. so.
4,
..so.q.....t...
np..
.....:.. o
..p....'.
?
..go,..noqo
...g....g....i.
tn.q.o
.q...
o,....
.....4
.o......
.n o...
.. d, o
.. d. o, 4,... ;.
~~4.....~~
.,......4..
n i.. n o l.. o un,
.o..
. 0..
4.....
.. d. o
.4...
200
"t""!""9" o.p....p....p.
~~3... 7 o.~~
n.
.no.
.. t. ' " t. " " ' t. ' ' ? " '
.n..
..u
..o.,.
.......o
.....ou...
+
1 o l. n.. 4. o n 1....
... d. ".. d... o d. o
~~4. n~~
.. i. n 3
..**....4.......4...
. d...
~~4. n.. d.... 6..~~
"....."..."...4....
o...
1 1
+
i p.o q o
~~4..... q. n.~~
~~4.... ny o,~~
c.o. &.... b o.. t..
...(.... t. o. n l.. u n t. n..
. + 4..... q u n o n. W o.
n. '..
g 4
...f.......u.....
..n.
t
.v.
~~o. n o....~~
.o.n..
.o e
1 1
1 k.. o a.' n. 4.... 4....
........4...
.. l... d...
~~4... u d.....~~
.. 4... u i.....J.....i...
n 4......
~~4. o. a. n i~~
t I
1 I
.qu.q.o.a..u 4..o.
>.. o 4.... u.;. o..
.. o. n u t.. u b..... (.....
.. t. u. n l. n o t...... t o.... d. o.qo n)n n.b n u.
.... t.... g..... :g.....t....
?....'%....p....
4 n..
g e..p..
.y
..p......j....
..j.....j.....j....j....
..)..
g..
O
~
g 3
50 100 150 200 250 300 AUCTIONEERED LOW MEASURED RCS TEMPERATURE (*F)
FIGURE 3.4-4b NILLSTONE - UNIT 3 3/4 4 41 Amendment No. pp.
0526
E K R&ENCY...ColLE COOLING SYSTEMS l
3/4.5.3 ECC5 $UBSYSTENS - T LESS THAN 350'F LINITING CONDITION FOR OPERATION t
3.5.3 As a minimum, one ECCS subsystem comprised of the following shall be OPERABLE:
a.
One OPERABLE centrifugal charging pump, l
l t
b.
One OPERABLE RHR heat exchanger, c.
One OPERABLE RHR pump, d.
One OPERABLE containment recirculation heat exchanger, e.
One OPERABLE containment recirculation pump, and f.
An OPERABLE flow path which, with manual realignment of valves, is capable of discharging to the RCS, taking suction from the refueling water storage tank, and transferring suction to the containment sump t
during the recirculation phase of operation.
APPLICABILITY: MODE 4.
ACTION:
With no ECCS subsystem OPERABLE because of the inoperability of a.
the centrifugal charging pump, the containment recirculation pump, the containment recirculation heat exchanger, the flow path from the refueling water storage tank, or the flow path capable of taking suction from the containment sump, restore at least one ECCS sub-system to OPERABLE status within I hour or be in COLD SHUTDOWN within the next 20 hours2.314815e-4 days 0.00556 hours 3.306878e-5 weeks 7.61e-6 months , b.
With no ECCS subsystem OPERABLE because of the inoperability of either the residual heat removal heat exchanger or RHR pump, restore at least one ECCS subsystem to OPERABLE status or maintain the Reac-tor Coolant System T., less than 350*F by use of alternate heat removal methods, In the event the ECCS is actuated and injects water into the Reactor c.
Coolant System, a Special Report shall be prepared and submitted to the Commission pursuant to Specification 6.9.2 within 90 days describ-ing the circumstances of the actuation and the total accumulated actuation cycles to date.
The current value of the usage factor for each affected Safety Injection nozzle shall be provided in this Special Report whenever its value exceeds 0.70.
NILLSTONE - UNIT 3 3/4 5-7 Amendment No.
05U
EEREEEY CORE C00 LINA SYSTEMS SURVEILLANCE REQUIRENENTS 4.5.3.1 The ECCS subsystem shall be demonstrated OPERABLE per the applicable requirements of Specification 4.5.2, with the exception that valves may be out of alignment but capable of being manually realigned.
I 4
gijLSTONE
' NIT 3 3/4 5-8 Amendment No.
J
l REACTIVITY. CONTROL SYSTEMS BASES MODERATOR TEMPERATURE COEFFICIENT (Continued)
These corrections involved:
(1) a conversion of the MDC used in the FSAR safety analyses to its equivalent HTC, based on the rate of change of moderator density with temperature at RATED THERMAL POWER conditions, and (2) subtracting from this value the largest differences in MTC observed between i'
EOL, all rods withdrawn, RATED THERMAL POWER conditions, and those most adverse conditions of moderator temperature and pressure, rod insertion, axial power skewing, and xenon concentration that can occur in normal operation and lead to a significantly more negative EOL MTC at RATED THERMAL POWER.
These corrections transformed the MDC value used in the FSAR safety analyses into the limiting End of Cycle Life (EOL) HTC value. The 300 ppm surveillance limit MTC value represents a conservative MTC value at a core condition of 300 apm equilibrium boron concentration, and is obtained by making corrections for aurnup and soluble boron to the limiting E0L MTC value.
The Surveillance Requirements for measurement of the MTC at the beginning and near the end of the fuel cycle are adequate to confirm that the MTC remains within its limits since this coefficient changes slowly due principally to the reduction in RCS boron concentration associated with fuel burnup.
3/4.1.1.4 MINIMUM TEMPERATURE FOR CRITICAllTY This specification ensures that the reactor will not be made critical i
with the Reactor Coolant System average temperature less than 551. This limitation is required to ensure:
(1) the moderator temperature coefficient is within it analyzed temperature ran e, (2) the trip instrumentation is within its normal operating range, (3 the P 12 interlock is above its setpoint (4) the pressurizer is capa le of being in an OPERABLE status with a steam bubble, and (5) the reactor vessel is above its minimum RTwor temperature.
3/4.1.2 BORAT10N SYSTEMS The Boron Injection System ensures that negative reactivity control is a M ble during each mode of facility operation. The components required to perfW m this function include:
(1) borated water sources, (2) charging pumps, (3) pparate flow paths, (4) boric acid transfer pumps, and (5) an emergency power supply from OPERABLE diesel generators.
With the plant in MODES 1, 2, or 3, a minimum of two boron injection l
flow paths are required to ensure single functional capability in the event an assumed failure renders one of the flow paths inoperable.
The boration capability of either flow path is sufficient to provide a SHUTDOWN MILL $ TONE - UNIT 3 B 3/4 1-2 Amendment No. U, pp, 0533
REACTIVITY CONTROL SYSTEMS BASES l
BORATION SYSTEMS (Continued)
MARGIN from expected operating conditions equivalent to that r after xenon decay and cooldown to 200'F.
The maximum boration capabilityequired by Figu boration volume conditions thro)ughout core operating life requirement is established to conservatively bound expected o minimum based on a minimum expected hot full power.
The initial RCS boron concentration is g
The final RCS boron concentration assumes that the most reactior hot ze inserted into the core.
21,802 gallons of 6600 ppm borated water from the boric a idThis ve contro" rod is not 1,166 000 gallons of 2700 ppm borated water from the re tank (,RWST).
c storage tanks or with ECCS requirement,A minimum RWST volume of 1,166,000 gallons is specified to be con i t s s ent i
With the plant in MODE 4, one boron injection flowpath is single failure consideration for emergency boration requirements on th stable reactivity condition of the reactor, the emergency power sup acceptable without the OPERABLE charging pump e basis of the at least MODE 5 requiremen,ts prior to cooldown to MODE 4and the fact th quirement for water addition path to the charging pumps is surveilled to be lockedAlso, t a direct dilution accident in MODE 4.
closed to prevent With the plant in MODES 5 and 6, one boron injection system is i
single failure consideration on the basis of the stable reactivity reactivity changes in the event the single boron inject i
condition of the and positive The limitation for a maximum of one centrifugal charging pump t esinoperable.l cold overpressure protection is in service, provides assurance that o be OPERABLE, when pressure transient can be relieved by operation of a single PORV or RHR
{
valve.
a mass addition suction relief The boron capability required below 200'F is sufficient t MARGIN of 1.3% Ak/k after xenon decay and cooldown from 200'F t o provide a SHUTDOWN requires either a usable volume of 4100 gallons of 6600 ppm borated o 140*F.
This condition boric acid storage tanks or 250,000 The unusable volume in each boric acid storage tank is 1300 gallons of 2700 p water from the water from the RWST.
e gallons.
The contained water volume limits include allowance for water not
{
available because of discharge line location and other physical char acteristics.
The limits on contained water volume and boron concentrat ensure a pH value of between 7.0 and 7.5 for the solution reci n of the RWST also containment after a LOCA. This pH band minimizes the evolution of iodin the effect of chloride rculated within and caustic stress corrosion on mechanicale and minimizes components.
systems and The minimum RWST solution temperature for MODES 5 and 6 analysis assumptions in addition to freeze protection considerations.
The ainimunVmaximum RWST solution temperatures for MODES 1, 2 analysis assumptions.
, 3 and 4 are based on MILLSTONE - UNIT 3 0633 B 3/4 1-3 Amendment 17, 77, 77 JJ7,
3/4.4 REACTOR COOLANT SYSTEM BASES 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION The plant is designed to operate in MODES 1 and 2 with three or four reactor coolant loops in operation and maintain DNBR greater than the design limit during all normal operations and anticipated transients. With less than the required reactor coolant loops in operation this specification requires that the plant be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .
In MODE 3, three reactor coolant loops, and in Mode 4, two reactor coolant loops provide sufficient heat removal capability for removing core decay heat even in the event uf a bank withdrawal accident; however, in MODE 3 l a cingle reactor coolant loop provides sufficient heat removal capacity if a bank withdrawal accident can be prevented, i.e., by opening the Reactor Trip System breakers.
in MODE 4, if a bank withdrawal accident can be prevented, a single l reactor coolant loop or RHR loop provides sufficient heat removal capability for removing decay heat; but single failure considerations require that at least two loops (any combination of RHR or RCS) be OPERABLE.
l l
In MODE 5, with reactor coolant loops filled, a single RHR loop provides sufficient heat removal capability for removing decay heat; but single failure considerations require that at least two RHR loops or at least one RHR loop and two steam generators be OPERABLE.
In MODE 5 with reactor coolant loops not filled, a single RHR loop provides sufficient heat removal capability for removing decay heat; but single failure considerations, and the unavailability of the steam generators as a heat removing component, require that at least two RHR loops be OPERABLE.
I l
in MODE 5, during a planned heatup to MODE 4 with all RHR loops removed from operation, an RCS loop, OPERABLE and in operation, meets the requirements of an OPERABLE and operating RHR loop to circulate reactor coolant. During the heatup there is no requirement for heat removal capability so the OPERABLE and operating RCS loop meets all of the required functions for the heatup condition.
Since failure of the RCS loop, which is OPERABLE and operating, could also cause the associated steam generator to be inoperable, the associated steam generator cannot be used as one of the steam generators used to meet the requirement of LC0 3.4.1.4.1.b.
The operation of one reactor coolant pump (RCP) or one RHR pump provides adequate flow to ensure mixing, prevent stratification and produce gradual reactivity changes during boron concentration reductions in the Reactor Coolant System.
The reactivity change rate associated with boron reduction will, therefore, be within the capability of operator recognition and control.
The restrictions on starting an RCP are provided to prevent RCS pressure transients, caused by energy additions from the Secondary Coolant System, which could exceed the limits of Appendix G to 10 CFR Part 50 or which could cause pressure excursions within the RHR system which would exceed the design pressure of the system.
The RCS will be protected against overpressure transients and will not exceed the limits of Appendix G by restricting starting of the RCPs based upon the secondary water temperature of each steam generator and the RCS wide range cold leg temperatures.
MILLSTONE - UNIT 3 B 3/4 4-1 Amendment No. 7. H #,
M32 l
l
3/4.4 KEACTOR COOLANT $YSTEN BASES (Continued) 4 The requirement to maintain the isola'ted loop stop valves shut with power 4
removed ensures that no reactivity addition to the core could occur due to the startup of an isolated loop. Verification of the boron concentration in an idle loop prior to opening the stop valves provides a reassurance of the adequacy of the boron concentration in the isolated loop. The 2600 ppm is sufficient to bound shutdown margin requirements and provide for boron concentration measurement uncertainty between the loop and the RWST. Draining and refilling the isolated loop within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months prior to opening its stop valves ensures adequate mixing of the coolant in this loop and prevents any reactivity effects due to boron concentration stratifications.
The requirement to have all reactor coolant pumps de-energized, prior to unisolating a loop, insures that the heat from the secondary side of the steam generator, in the loop being unisolated, does not result in an energy addition transient during the return of the loop to service.
NILLSTONE - UNIT 3 83/44-la Amendment No. 7. JP 0546
REACTOR COOLANT SYSTD1 BASES SPECIFIC ACTIA JJ (Continued)
Based upon the above considerations for excluding certain radionuclides from the sample analysis, the allowable time of 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months between sample taking and completing the initial analysis is based upon a typical time necessary to perform the sampling, transport the sample, and perform the analysis of about 90 minutes. After 90 minutes, the gross count should be made in a reproducible geometry of sample and counter having reproducible beta or gamma self-shielding properties.
The counter should be reset to a reproducible efficiency versus energy.
It is not necessary to identify specific nuclides.
The radiochemical determination of nuclides should be based on multiple counting of the sample within typical counting basis following sampling of less than I hour, about 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , about I day, about I week, and about 1 month.
Reducing T, to less than 500'F prevents the release of activity should a steam generator tube rupture since the saturation pressure of the reactor coolant is below the lift pressure of the atmospheric steam relief valves.
The Surveillance Requirements provide adequate assurance that excessive specific activity levels in the reactor coolant will be detected in sufficient time to take corrective action. A reduction in frequency of isotopic analyses following power changes may be permissible if justified by the data obtained.
3/4.4.9 PRESSVRE/ TEMPERATURE LIMITS REACTOR COOLANT SYSTEM (EXCEPT THE PRESSURIZER)
BACKGROU"D All components of the RCS are designed to withstand effects of cyclic loads due to system pressure and temperature changes.
These loads are introduced by startup (heatup) and shutdown (cooldown) operations, power transients, and reactor trips. This LC0 limits the pressure and temperature changes during RCS heatup and cooldown, within the design assumptions and the stress limits for cyclic operation.
The LCO and Figures 3.4-2 and 3.4-3 contain P/T limit curves for heatup, cooldown, inservice leak and hydrostatic (ISLH) testing, and data for the maximum rate of change of reactor coolant temperature.
Each P/T limit curve defines an acceptable region for normal operation. The usual use of the curves is operational requirements 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. A heatup or cooldown is defined as a temperature increase or decrease of greater than or equal to 10*F in any one hour period.
This definition of heatup and cooldown is based upon the ASME definition of isothermal conditions described in ASME,Section XI, Appendix E.
MILLSTONE - UNIT 3 B 3/4 4-7 Amendment No.
0528
i REACTOR COOLANT SYSTEN j
BASES f
I i
PRESSURE / TEMPERATURE LIMITS fcontinued) i Steady state thermal conditions exist when temperature increases or l
decreases are <10'F in any one hour period and when the plant is not performing a planned heatup or cooldown in accordance with a procedure. During steady state 3
thermal conditions, the limits of the heatup and cooldown curves do not apply.
i Cold overpressure protection is adequate to protect the reactor coolant system.
The LCO establishes operating limits that provide a margin to brittle j
failure of the reat. tor vessel and piping of the reactor coolant pressure boundary (RCPB). The vessel is the component most subject to brittle failure, and the LCO i
limits apply a.ainly to the vessel. The limits do not apply to the Pressurizer, 1-P which has different design characteristics and operating functions which are i
addressed by LC0 3.4.9.2, " Pressurizer".
d
(
10 CFR 50, Appendix G (Ref.1), requires the establishment of P/T limits for j
specific material fracture toughness requirements of the RCPB materials.
Reference 1 requires an adequate margin to brittle failure during normal operation, anticipated operatior,a1 occurrences, and system hydrostatic tests.
It mandates the use of the American Society of Mechanical Engineers (ASME) Code, l
SectionIII,AppendixG(Ref.2).
~
The neutron embrittlement effect on the material toughness is reflected by 1
l increasing the nil ductility reference temperature (RTwor) as exposure to neutron j
fluence increases.
1 The actual shif t in the RTuor of the vessel material will be established j,
periodically by removing and evaluating the irradiated reactor vessel material specimens, in accordance with ASTM E 185 (Ref. 3) and Ap>endix H of 10 CFR 50 (Ref. 4). The operating P/T limit curves will be adjustec, as necessary, based i
on the evaluation findings and the recommendations of Regulatory Guide 1.99 (Ref. 5).
i 1
l The P/T limit curves are composite curves established by superimposing L
limits derived from stress analyses of those portions of the reactor vessel and head that are the most restrictive. At any specific pressure, temperature, and temperaturo rate of change, one location within the reactor vessel will dictate the most restrictive limit. Across the s aan of the P/T. limit curves, different b
locations may be more restrictive, and tius, the curves are composites of tha l-most restrictive regions.
i The heatup curve represents a different set of restrictions than the
[
cooldown curve because the directions of the thermal gradients through the vessel wall are reversed.
The thermal gradient reversal alters the location of the l
tensile stress between the outer and inner walls.
i
.The P/T limits include uncertainty margins to ensure that the calculated J.
limits are not inadvertently exceeded. These margins include gauge and system loop uncertainties, elevation differences, containment pressure conditions and j-system pressure drops between the beltline region of the vessel and the pressure
. gauge or relief valve location.
In an effort to minimize the system frictional-i losses,' additional restrictions on RCP operation below 160*F are provided in the f.
LCO. These restrictions result in increased acceptable system pressures enabling-
{
f greater operator flexibility during heatup and cooldown in MODE 5.
e 4l NILLSTONE - UNIT 3 8 3/4 4-8 Amendment No..pp, om
-u
.a -
REACT 0_R COOLANT SYSTf3 BASES PRESSURE / TEMPERATURE LIMITS fcontinued)
The criticality limit curve includes the Reference 1 requirement that it be 140'F above the heatup curve or the cooldown curve, and not less than the minimum permissible temperature for ISLH testing. However, the criticality curve is not operationally limiting; a more restrictive limit exists in LCO 3.1.1.4,
" Minimum Temperature for Criticali+y."
The consequence of violating the LC0 limits is that the RCS has been operated under conditions that can result in brittle failure of the RCPB, possibly leading to a nonisolable leak or loss of coolant accident. In the event these limits are exceeded, an evaluation must be performed to determine the effect on the structural integrity of the RCPB components.
The ASME Code,Section XI, Appendix E (Ref 6) provides a recommended methodology for evaluating an operating event that causes an excursion outside the limits.
APPLICABLE SAFETY ANALYSIS The P/T linits are not derived from Design Basis Accident (DBA) analyses.
They are prescribed during normal operation to avoid encountering pressure, temperature, and temperature rate of change conditions that might cause undetected flaws to propagate and cause nonductile failure of the RCPB, an unanalyzed condition.
Reference 2 establishes the methodology for determining the P/T limits. Alti)ough the P/T limits are m.ot derived from any DBA, the P/T limits are acceptance limits since they preclude operation in an unanalyzed condition.
RCS P/T limits satisfy Criterion 2 of 10CFR50.36(c)(2)(ii).
LCD The two elements of this LC0 are:
a.
The limit curves for heatup, cooldown, and ISLH testing; and b.
Limits on the rate of change of temperature.
The LC0 limits apply to all components of the RCS, except the Pressurizer. These limits define allowable operating regions while providing margin against nonductile failure.
The limits for the rate of change of temperature control the thermal gradient through the vessel wall and are used as inputs for calculating the heatup, cooldown, and ISLH testing P/T limit curve. Thus, the LC0 for the rate of change of temperature restricts stresses caused by thermal gradients and also ensures the validity of the P/T limit curves.
MILLSTONE - UNIT 3 8 3/4 4-9 Amendment No.
0628
REACTOR COOLANT SY$IEH sASES PRESSURE / TEMPERATURE LIMITS fcontinued)
Violating the LC0 limits places the reactor vessel outside of the bounds of the analyses and can increase stresses in other RCPB components.
The consequences depend on several factors, as follows:
The severity of the departure from the allowable operating P/T regime a.
or the severity of the rate of change of temperature; b.
The length of time the limits were violated (longer violations allow the temperature gradient in the thick vessel walls to become more pronot ned); and 4
c.
The existences, sizes, and orientations of flaws in the vessel material.
APPLICABILITY J
I The RCS P/T liniits LC0 provides a definition of acceptable operation for prevention of nonductile failure in accordance with 10CFR50, Appendix G (Ref.1).
The P/T limits were developed to provide requirements for operation during heatup or cooldown (MODES 3, 4, 3nd 5) or ISLH testing, in keeping with the concern for nonductile failure.
Tha limits do not apply to the Pressurizer.
During MODES 1 and 2, other Technical Specifications provide limits for operation that can be more restrictive than or can supplement these P/T limits. LC0 3.2.5, "DNB Parametors'; LCO 3.2.3.1 and 3.2.3.2, "RCS Flow Rate and Nuclear Enthalpy j
Rise Hot Channel Factor Four Loops Operating /Three Loops Operating";
LC0 3.1.1.4, "Minimu; %gerature for Criticality"; and Safety Limit 2.1, " Safety Limits," also proviu en%tional restrictions for pressure and temperature and maximum pressure. Ftn UK.more, MODES 1 and 2 are above the temperature range of cor.:: err. for nonductile failure, and stress analyses have been performed for normal maneuvering profiles, such as power ascension or descent.
ACTIONS Operation outside the P/T limits must be corrected so that the RCPB is returned to a condition that has ieen verified by stress analyses. The 30 minute Allowed Outage Time (A0T) reflet '.s the urgency of restoring the parameters to within the a ~
analyzed range.
Most violations will not be severe, and the activity can be accomplished in this time in a controlled manner.
Besides restoring operation within limits, an evaluation is required to determine if RCS operation can continue.
The evaluation must verify the RCPB integrity remains acceptable and must be completed before continuing operation.
Several methods may be used, including comparison with pre-analyzed transients in the stress analyses, new analyses, or inspection of the components.
MILLSTONE - UNIT 3 8 3/4 4-10 Amendment No.
0528
REACTOR COOLANT SYSTEM BASES PRESSURE / TEMPERATURE LIMITS (continuedi 4
the may he used to support 6),
f the vessel beltline.
Section XI, Appendix E (Ref.
evaluation. However, its use is restricted to evaluation o The evaluation for ASME Code, The 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months A0T is reasonable to accomplish the evaluation.
vere violations may a mild violation is possible within this time, but more se tions.
A favorable require special, event specific stress analyses or inspec t
evaluation must be completed before continuing to opera e.
dd Restoration This evaluation must be completed whenever a limit is excee e.
h alyzed stresses may within 30 minutes alone is insufficient because higher t an an have occurred and may have affected the RCPB integrity.
llowed times, the If the required remedial actions are not completed within the athe RCS r h
plant must be placed in a lower MODE because eit eran extended period of region for t ble region.
Either sufficiently severe event caused entry into an unaccep ation of the event, best unacceptable P/T i
In reduced possibility indicates a need for more careful exam naacco
{
ture.
and temperature conditions, the possibility pressure undetected flaws is decreased.
thin 30 minutes,
' f the required restoration activity cannot be accomplished wias speci t
action must be implemented to reduce pressure and tempera ure lished within ACTION statement.
If the required evaluation for continued operation cannot be accomp t proceed to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or the results are indeterminate or unfavorable, action mus i
statement.
A reduce pressure and temperature as specified in th turning to operating pressure and temperature conditions.
to MODE 3 within Pressure and temperature are reduced by bringing h
xt 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .
h the required The A0Ts are reasonable, based on operating experience to reac d l manner and without plant conditions from full power conditions in an or er y challenging plant systems.
SURVEILLANCE RE0UIREMENTS ll as the limits of Verification that operation is within the LCO limits a nd RCS pressure a is This frequency temperature conditions are undergoing planned chang available to monitor i
RCS status.
AmendmentNo.(p.77 B3/44-11 MILLSTONE - UNIT 3 0528
REACTOR COOLANT SYSTEM.
BASES PRESSURE / TEMPERATURE LIMITS (continued)
ASME Code,Section XI, Appendix E (Ref. 6), may be used to support the evaluation. However, its use is restricted to evaluation of the vessel beltline.
The 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months A0T is reasonable to accomplish the evaluation. The evaluation for a mild violation is possible within this time, but more severe violations may require special, event specific stress analyses or inspections.
A favorable evaluation must be completed before continuing to operate.
This evaluation must be completed whenever a limit is exceeded.
Restoration within 30 minutes alone is insufficient because higher than analyzed stresses may have occurred and may have affected the RCPB integrity.
If the required remedial actions are not completed within the allowed times, the plant must be placed in a lower MODE because either the RCS remained in an unacceptable P/T region for an extended period of increased stress or a sufficiently severe event caused entry into an unacceptable region.
Either possibility indicates a need for more careful examination of the event, best accomplished with the RCS at reduced pressure and temperature.
In reduced pressure and temperature conditions, the possibility of propagation with undetected flaws is decreased, If the required restoration activity cannot be accomplished within 30 minutes, action must be implemented to reduce pressure and temperature as specified in the ACTION statement.
If the required evaluation for continued operation cannot be accomplished within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or the results are indeterminate or unfavorable, action must proceed to reduce pressure and temperature as specified in the Action statement.
A favorable evaluation must be completed and documented before returning to operating pressure and temperature conditions.
Pressure and temperature are reduced by bringing the plant to MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and to MODE 5 with RCS pressure < 500 psia within the next 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .
The A0Ts 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.
SURVEILLANCE RE0VIREMENTS Verification that operation is within the LC0 limits as well as the limits of Figures 3.4-2 and 3.4-3 is required every 30 minutes when RCS pressure and temperature conditions are undergoing planned changer.
This frequency is considered reasonable in view of the control room indication available to monitor RCS status.
MILLSTONE - UNIT 3 B 3/4 4-11 Amendment No pp. 77 0528
i-REACTOR COOLANT SYSTEM BASES 4
PRESSURE / TEMPERATURE LIMITS (continued)
Surveillance for heatup, cooldown, or ISLH testing may be discontinued when the i
' definition given in the relevant plant procedure for ending the activity is satisfied, i
This Surveillance Requirement is only. required to be performed during system j
heatup, cooldown, and ISLH testing.
No Surveillance Requirement is given for i
criticality operations because LCO 3.1.1.4 contains a more restrictive requirement.
The Surveillance Requirement to remove and examine the reactor vessel material i
irradiation surveillance specimens is in accordance with the requirements of 10CFR50, Appendix H.
REFERENCES
(
l.
10CFR50, Appendix G.
2.
ASME, Boiler and Pressure Vessel Code,Section III, Appendix G.
j 3.
ASTM E 185-82, July 1982.
4.
10CFR50, Appendix H.
5.
Regulatory Guide 1.99, Revision 2, May 1988.
l 6.
ASME, Boiler and Pressure Vessel Code,Section XI, Appendix E, t
a l
PRESSURIZER LEACKGROUND i-The Pressurizer is part of the RCPB, but is not subject to the same rertrictions l-as the rest of the RCS.
This LC0 limits the temperature changes of the l
Pressurizer and allowable temperature differentials, within the design assumptions and the stress limits for cyclic operation.
1 i
i 4
4 i
MILLSTONE - UNIT 3 B 3/4 4-12 Amendment No. J7 0628
- = -
~
REACTOR COOLANT SYSTEM
- BASES PRESSURIZER (continued)
The LCO contains the Pressurizer limits for heatup, cooldown, and spray water temperature differential. Each temperature limit defines an acceptable region for normal operation. The limits that apply to the Pressurizer are as follows: The Pressurizer heatup and cooldown rates shall not exceed 100'F/hr and 200'F/hr, respectively. The spray shall not be used if the temperature difference between the Pressurizer and the spray fluid is greater than 320'F.
The heatup limit represents a different set of restrictions than the cooldown limit because the directions of the thermal gradients through the Pressurizer wall are reversed.
The thermal gradient reversal alters the location of the tensile stress between the outer and inner walls.
~
The consequence of violating the LC0 limits is that the Pressurizer has been operated under conditions that can result in failure, possibly leading to a nonisoiable leak or loss of coolant accident.
In the event these limits are exceeded, an evaluation must be performed to determine the effect on the structural integrity of the Pressurizer.
- APPLICABLE SAFETY ANALYSIS The Pressurizer temperature limits are not derived from Design Basis Accident (DBA) analyses.
They are prescribed during normal operation to avoid encountering temperature and temperature rate of change conditions that might cause the initiation / propagation of undetected cracks and cause failure of the pressure boundary.
LCO l
The two elements of this LC0 are:
a.
Limits on the rate of change of temperature; and 4
b.
Limits on the spray water differential-temperature.
The LC0 limits apply to the Pressurizer. These limits define allowable operating regions and permit a large number of operating cycles while providing a margin
~
to nonductile failure.
The limits for the rate of change of temperature control the thermal gradient 3
through the Pressurizer wall and, therefore, restricts stresses caused by thermal gradients.
Violating the LC0 limits places the Pressurizer outside of the bounds of the stress analyses.
The consequences depend on several factors, as follow:
The severity of the rate of change of temperature; a.
MILLSTONE - UNIT 3 B 3/4 4-13 Amendment No. #
0528 4
4
REACTOR COOLANT SYSTEM BASES-PRESSURIZER (continued) b.
The length of time the limits were violated (longer violations allow the temperature gradient in -the Pressurizer walls to become more pronounced); and c.
The existences, sizes, and crientations of flaws in the Pressurizer material.
APPLICABILITY The -Pressurizer temperature limits LCO provides a definition of acceptable operation-for prevention of failure. The temperature limits were developed to provide requirements for operation during heatup or cooldown, and their Applicability is at all times in keeping with the concern for failure.
ACTIONS Operation outside the temperature limits must be corrected so that the Pressurizer is returned to a condition that has been verified by stress analyses.
-The 30 minute A0T reflects the urgency of restoring the parameters to within the analyzed range.
Most violations will not be severe, and the activity can be accomplished in this time.in a controlled manner.
Elesides restoring operation within limits, an evaluation is required to determine if Pressurizer operation -can continue.
The evaluation must verify the Pressurizer integrity remains acceptable and must be completed before continuing operation. Several methods may be used, including comparison with pre-analyzed transients in the stress analyses, new analyses, or inspection of the components.
The 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months A0T is reasonable to accomplish the evaluation. The evaluation ur a mild violation is possible within this time, but more severe violations may require special, event specific stress analyses or inspections.
A favorable evaluation must be completed before continuing to operate.
This evaluation must be completed whenever a limit is exceeded.
Restoration within 30 minutes alone is insufficient because higher than analyzed stresses may have occurred and may have affected the Pressurizer integrity.
If the required remedial actions are not completed within the allowed times, the plant must be placed in a lower MODE because a sufficiently severe event may have caused entry into an unacceptable region. 'This possibility indicates a need for more careful examination of the event, best accomplished with the Pressurizer at reduced pressure. In reduced pressure conditions, the possibility of propagation with undetected flaws is decreased.
If the required restoration activity cannot be accomplished within 30 minutes, action must be implemented to reduce pressure as specified in the ACTION statement.
MILLSTONE - UNIT 3 B 3/4 4-14 Amendment No. pp. 77 0528
REACTOR COOLANT SYSTEM BASES PRESSURIZER (continued)
If the required evaluation for continued operation cannot be accomplished within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or the results are indeterminate or unfavorable, action must proceed to reduce pressure as specified in the Action statement.
A favorable evaluation must be completed and documented before returning to operating pressure conditions.
Pressure is reduced by bringing the plant to MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . Pressure is further reduced by bringing the plant to MODE 4 or 5 and reducing Pressurizer pressure < 500 psia within the next 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .
The A0Ts 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.
SVRVEILLANCE RE0VIREMENTS Verification that operation is within the LCO heatup and cooldown limits is required every 30 minutes when Pressurizer temperature conditions are undergoing planned changes. This Frequency is considered reasonable in view of the control room indication available to monitor Pressurizer status. Surveillance for heatup or cooldown may be discontinued when the definition given in the relevant plant procedure for ending the activity is satisfied. The Surveillance Requirement for heatup or cooldown is only required to be performed during system heatup and cooldown.
Verification that operation is within the LC0 spray water temperature differential limit is required every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months when auxiliary spray is in operation. This Frequency. is considered reasonable in view of the control room indication available to monitor Pressurizer status.
OVERPRESSURE PROTECTION SYSTEMS BACKGROUND The Cold Overpressure Protection System limits RCS pressure at low temperatures so the integrity of the reactor coolant pressure boundary (RCPB) is not compromised by violating the pressure and temperature (P/T) limits of 10CFR50, Appendix G (Ref. 1).
The reactor vessel is the limiting RCPB component for demonstrating such protection.
Cold Overpressure Protection consists of two PORVs with nominal lift setting as specified in Figures 3.4-4a and 3.4-4b, or two residual heat removal (RHR) suction relief valves, or one PORV and one RHR suction relief valve, or a depressurized RCS and an RCS vent of sufficient size.
Two relief valves are required for redundancy. One relief valve has adequate relieving capability to prevent overpressurization of the RCS for the required mass input capability.
MILLSTONE - UNIT 3 B 3/4 4-15 Amendment No. f7, 77. J77 0528
REACTOR COOLANT SYSTEM BASES I
OVERPRESSURE PROTECTION SYSTEMS (continued)
The reactor vessel material is less tough at low temperatures than at normal operating temperature. As the vessel neutron exposure accumulates, the material toughness decreases and becomes less resistant to stress at low temperatures (Ref. 2). RCS pressure, therefore, is maintained low at low temperatures and is increased only as temperature is increased.
The potential for vessel overpressurization is most acute when the RCS is water solid, occurring while shutdown; a pressure fluctuation can occur more quickly than an operator can react to relieve the condition.
Exceeding the RCS P/T limits by a significant amount could cause nonductile cracking of the reactor vessel.
LC0 3.4.9.1, " Pressure / Temperature Limits - Reactor Coolant System,"
requires administrative control of RCS pressure and temperature during heatup and cooldown to prevent exceeding the limits provided in Figures 3.4-2 and 3.4-3.
This LC0 provides RCS overpressure protection by limiting mass input capability and requiring idequate pressure relief capacity. Limiting mass input capability requires all Safety Injection (SIH) pumps and all but one centrifugal charging pump to be incapable.of injection into the RCS.
The pressure relief. capacity requires either two redundant relief valves or a depressurized RCS and an RCS vent of sufficient size.
One relief valve or the open RCS vent is the overpressure protection device that acts to terminate an increasing pressure event.
With minimum mass input capability, the ability to provide core coolant addition is restricted. The LCO does not require the makeup control system deactivated or the safety injection (SI) actuation circuits blocked.
Due to the lower l
pressures in the Cold Overpressure Protection MODES and the expected core decay heat levels, the makeup system can provide adequate flow via the makeup control valve.
PORV Reouirements As designed, the PORV Cold Overpressure Protection (COPPS) is signaled to open if the RCS pressure approaches a limit determined by the COPPS actuation logic.
The COPPS actuation logic monitors both RCS temperature and RCS pressure and determines when the nominal setpoint of Figure 3.4-4a or Figure 3.4-4b is approached.
The wide range RCS temperature indications are auctioneered to select the lowest temperature signal.
The lowest temperature signal is processed through a function generator that calculates a pressure setpoint for that temperature.
The calculated pressure setpoint is then compared with the indicated RCS pressure from a wide range pressure channel.
If the indicated pressure nieets or exceeds the calculated value, a PORV is signaled to open.
Figure 3.4-4a and Figure 3.4-4b present the PORV setpoints for COPPS.
Above 110*F, the setpoints are staggered so only one valve opens during a low MILLSTONE - UNIT 3 B 3/4 4-16 Amendment No. J7, 77 0528 1
REACTOR COOLANT SYSTEM BASES OVERPRESSURE PROTECTION SYSTEMS (continued) temperature overpressure transient. Setting both valves to the values of Figure 3.4-4a and Figure 3.4-4b within the tolerance allowed for the calibration accuracy, ensures that the Reference 1 limits will not be exceeded for the analyzed isothermal events.
When a PORV is opened, the release of coolant will cause the pressure increase to slow and reverse. As the PORV releases coolant, the RCS pressure decreases until a reset pressure is reached and the valve is signaled to close.
The pressure continues to decrease below the reset pressure as the valve closes.
ILT Suction Relief Valve Reauirements The isolation valves between the RCS and the RHR suction relief valves must be open to make the RHR suction relief valves OPERABLE for RCS overpressure mitigation. The RHR suction relief valves are spring loaded, bellows type water relief valves with setpoint tolerances and accumulation limits established by Section III of the American Society of Mechanical Engineers (ASME) Code (Ref. 3) for Class 2 relief valves, When the RHR system is operated for decay heat removal or low pressure letdown-control, the isolation valves between the RCS and the RHR suction relief valves are open, and the RHR suction relief valves are exposed to the RCS and are able to relieve pressure transients in the RCS, RCS Vent Reauirements Once the RCS is depressurized, a vent exposed to the containment atmosphere will maintain -the RCS at containment ambient pressure in an RCS overpressure transient, if the relieving requirements of the transient do not exceed the capabilities of the vent. Thus, the vent path must be capable of relieving the flow resulting from the limiting mass or heat input transient, and maintaining pressure below the P/T limits for the analyzed isothermal events.
For an RCS vent to meet the flow capacity requirement, it requires removing a Pressurizer safety valve, removing a PORV and disabling its block valve in the open position, removing a Pressurizer-manway, or similarly establishing a vent by opening an RCS vent valve provided that the opening neets the size requirements. The vent path must be above the level of reactor coolant, so as not to drain the RCS when open.
MILLSTONE - UNIT 3 B 3/4 4-17 Amendment No.
0528
REACTOR COOLANT SYSTEM BASES OVERPRESSURE PROTECTION SYSTEMS (continued)
APPLICABLE SAFETY ANALYSIS Safety' analyses (Ref. 4) demonstrate that the reactor vessel is adequately protected against exceeding the P/T limits for the analyzed isothermal events.
In MODES 1, 2, AND 3, and in MODE 4, with RCS cold leg temperature exceeding 275'F, the pressurizer safety valves will provide RCS overpressure protection in the ductile region. At 275'F and below, overpressure prevention is provided by two means:
(1) two OPERABLE relief valves, or (2) a depressurized RCS with a sufficiently sized RCS vent, as required by NUREG-0800, RSB 5-2 for temperatures less than.RTuor + 90*F.
Each of these means has a limited overpressure relief capability.
The required RCS temperature for a given pressure increases as the reactor vessel material toughness -decreases due to neutron embrittlement.
Each time the Technical-Specification curves are revised, the cold overpressure protection must be re evaluated to ensure its functional requirements continue to be met using the RCS relief valve method or the depressurized and vented RCS condition.
Transients capable of overpressurizing the RCS are categorized as either mass or heat input transients, examples of which follow:
Mass Inout Transients a.
Inadvertent safety injection; or b.
Charging / letdown flow mismatch Heat Inout Transients a.
Inadvertent actuation of Pressurizer heaters; b.
Loss of RHR cooling; or c.
Reactor coolant pump (RCP) startup with temperature asymmetry within the RCS or between the RCS and steam generators.
The Technical Specifications ensure that mass input transients beyond the operability of the cold overpressure protection means do not occur by rendering all Safety Injection Pumps and all but one centrifugal charging pump incapable of injecting into the RCS whenever an RHR suction relief valve is unisolated from the RCS or whenever any PORV has COPPS armed and its block valve open.
The Technical Specifications ensure that energy addition transients beyond the operability of the cold overpressure protection means do not occur by limiting reactor coolant pump starts. LC0 3.4.1.4.1, " Reactor Coolant Loops and Coolant Circulation - Cold Shutdown - Loops Filled," LC0 3.4.1.4.2, " Reactor Coolant MILLSTONE - UNIT 3 B 3/4 4-18 Amendment No.
0528
REACTOR COOLANT SYSTEM BASES.
OVERPRESSURE PROTECTION SYSTEMS (continued) i e
Loops and Coolant Circulation Cold Shutdown Loops Not Filled," and LCO 3.4.1.3, " Reactor Coolant Loops and Coolant Circulation - Hot Shutdown" limit reactor coolant pump starts to one of the following plant conditions:
a.
An RCP is running, and The wide range cold leg temperature of any unisolated RCS loop is l
>l60*F, or b.
Two or more RCS loops are isolated, and An RCP is not running, and The secondary side water temperature of any steam generator in an i
unisolated loop is equal to or less than the wide range cold leg temperature of any unisolated RCS loop, or c.
No more than one RCS loop is isolated, and An RCP is not running, and j
Any RHR suction relief valve is unisolated from the RCS, and The secondary side water temperature of any steam generator in an 3
unisolated loop is either:
>200*F and equal to or less than the wide range cold leg l
l temperature of any unisolated RCS loop, or
$200'F and $50*F hotter than the wide range cold leg temperature i
of any unisolated RCS loop. (Note: Reactor coolant pumps cannot be run with the wide range cold leg temperature of any unisolated
{
RCS loop <160*F 'if any PORV has C6PPS armed and has its block valve open.), or j
d.
No more than one RCS loop is isolated, and An RCP is not running, and The RHR suction relief valves are isolated from the RCS, and The wide range cold leg temperature of any unisolated RCS loop 2160*F, 3
and Any PORV has COPPS armed and has its block valve open, and The secondary side water temperature of any steam generator in an unisolated loop is either:
equal to or less than the wide range cold leg temperature of any unisolated RCS loop, or
<250*F and $50'F hotter than the wide range cold leg temperature of any unisolated RCS loop, or e.
The RHR suction relief valves are isolated from the RCS, and Both PORVs are isolated or COPPS is blocked, and The wide range cold leg temperature of any unisolated RCS loop is 1
j
>275'F.
MILLSTONE - UNIT 3 B 3/4 4-19 Amendment No.
l os2s
i REACTOR COOLANT SYSTEM i
- BASES OVERPRESSURE PROTECTION SYSTEMS (continued)
I-The cold overpressure transient analyses demonstrate that either one relief valve or the depressurized RCS and RCS vent can maintain RCS pressure below limits when RCS letdown is isolated and only one centrifugal charging pump is operating.
1 Thus, the LCO allows only one centrifugal charging pump capable of injecting when cold overpressure prot 9ction.is required.
J The cold overpressure protection enabling temperature is conservatively established at a value 2 275'F based on the criteria described in Branch Technical Position RSB 5-2 provided in the Standard Review Plan (NUREG-0800).
PORV Performance The 10CFR50 Appendix G analyses show that the vessel is protected against non-ductile failure when the PORVs are set to open at the values shown in Figures t
3.4-4a and 3.4-4b within the tolerance allowed for the calibration accuracy. The curves are derived by analyses that model the performance of the PORV cold overpressure protection system (COPPS), assuming the limi;ing mass and heat transients of one centrifugal charging' pump injecting into thi RCS, or the energy addition as a result of starting an RCP with temperature asymmetry between the RCS and the steam generators.
These analyses consider pressure overshoot and undershoot beyond the PORV opening and closing, resulting from signal processing and valve stroke times.
l The PORV setpoints in Figures 3.4-4a and 3.4-4b will be updated when the P/T i
limits conflict'with the cold overpressure analysis limits. The P/T limits are periodically modified as the reactor vessel material toughness decreases due to 1
neutron embrittlement.
Revised limits are determined using neutron fluence projections and the results of testing of the reactor vessel material irradiation surveillance specimens. The Bases for LCO 3.4.9.1, " Pressure / Temperature Limits
- Reactor Coolant System (Except the Pressurizer)," discuss these evaluations.
i The PORVs are considered active components.
Thus, the failure of one PORV is i
assumed to represent the worst case, single active failure.
RHR Suction Relief Valve Performance i
The RHR suction relief valves do not have variable pressure and temperature lift setpoints as do the PORVs. Analyses show that one RHR suction relief valve with a setpoint at or between 426.8 psig and 453.2 psig will pass flow greater than that required for the limiting cold overpressure transient while maintaining RCS pressure less than the isothermal P/T limit curve. Assuming maximum relief flow requirements during the limiting cold overpressure event, an RHR suction relief valve will maintain RCS pressure to s 110% of the nominal lift setpoint.
1 i
Although each RHR suction relief valve is a passive spring loaded device, which meets single failure criteria, its location within the RHR System precludes meeting single failure criteria when spurious RHR suction isolation valve or RHR suction valve closure is postulated. Thus the loss of an RHR suction relief t
MILLSTONE - UNIT 3 B 3/4 4-20 Amendment No.
0528
i REACTOR COOLANT SYSTEM i
BASES OVERPRESSURE PROTECTION SYSTEMS (continued) valve is the worst case single failure. Also, as the RCS P/T limits are revised to reflect change in toughness in the reactor vessel materials, the RHR suction relief valve's analyses must be re-evaluated to ensure continued accommodation of the design bases cold overpressure transients.
RCS Vent Performance With the RCS depressurized, analyses show a vent size of 2 5.4 square inches is capable of mitigating the allowed cold overpressure transient. The capacity of this vent size is greater than the flow of the limiting transient, while maintaining RCS pressure less than the maximum pressure on the isothermal P/T limit curve.
The RCS vent size will be re-evaluated for compliance each time the isothermal P/T limit curves are revised.
The RCS vent is a passive device and is not subject to active failure.
The RCS vent satisfies Criterion 2 of 10CFR50.36(c)(2)(ii).
RCP Seal Protection As described above, the analyses of the cold overpressure transients result in pressure overshoot and undershoot beyond the PORV opening and closing setpoints, resulting from signal processing and valve stroke times. The valve overshoots are considered in the generation of the PORV setpoints presented in. Figures 3.4-4a and 3.4-4b.
The valve undershoots are also evaluated in terms of potential damage to the RCP
~~1 seal. The minimum pressure, considering valve undershoot, must be higher than that required to maintain the RCP #1 seal as a film riding seal.~~
This requirement resulted in restrictions on the operation of pumps when the cold overpressure protection is being provided by one or two PORVs. Specifically, a.
When the RCS cold leg temperature of any unisolated loop is less than 160 degrees F, the PORV block valves are open, and the PORV's Cold Overpressure Protection System (COPPS) is armed, no RCPs may be in operation.
LC0 3.4.1.4.1,
" Reactor Coolant Loops and Cool ant Circulation - Cc.ld Shutdown Loops Filled," and LC0 3.4.1.4.2,
" Reactor Coolant Loops and Coolant Circulation - Cold Shutdown - Loops Not Filled," provide this protection.
b.
When COPPS is armed, with the steam generator secondary side 1250*F, heat injection transients due to the start of the first RCP with a temperature asymmetry between the RCS and the steam generators is prohibited.
LC0 3.4.1.3, " Reactor Coolant System - Hot Shutdown,"
provides this protection for MODE 4.
MILLSTONE - UNIT 3 B 3/4 4-21 Amendment No.
0528
REACTOR COOLANT SYSTEM BASES OVERPRESSURE PROTECTION SYSTEMS (continued) c.
When COPPS is armed, PORV un@r' hoot is analyzed for mass injection transients limited to one ch: Jing pump.
LCO 3.4.9.3,
" Reactor Coolant System - Overpressu:e Protection Systems," provides this protection by requiring both safety injection pumps and all but one charging pump to be incapable of injection into the RCS, In order to provide protection for the RCP #1 seal, a PORV setpoint of 2 595 psia j
for temperatures 2 160 degrees F must be met. This minimum setpoint is derived by adding the applicable train uncertainty and valve undershoot to the required i
minimum RCS pressure required for seal integrity.
Due to the differing instrument uncertainties for the two trains of PORV COPPS, the train with the j
highest uncertainty is paired to the high setpoint curve.
i g-l This LC0 requires that cold overpressure protection be OPERABLE and the maximum i
mass irput be limited to one charging pump. Failure to meet this LC0 could lead to the loss of low temperature overpressure mitigation and violation of the Reference 1 isothermal limits as a result of an operational transient.
To limit the mass input capability, the LC0 requires a maximum of one centrifugal charging pump capable of injecting into the RCS.
The elements of the LC0 that provides low temperature overpressure mitigation through pressure relief are:
1.
Two OPERABLE PORVs; or A PORV is OPERABLE for cold overpressure protection when its block valve is i.
open, its lift setpoint is set to the nominal setpoints provided by Figure 3.4-4a or' 3.4-4b and when the surveillance requirements are met.
2.
Two OPERABLE RHR suction relief valves; or An RfiR suction relief valve is OPERABLE for cold overpressure protection i
when its isolation valves from the RCS are open and when its setpoint is at or between 426.8 psig and 453.2 psig, as verified by requirtd testing.
3.
One OPERABLE PORV and one OPERABLE RHR suction relief valve; or 4.
A depressurized RCS and an RCS vent.
An RCS vent is OPERABLE when open with en area of 2 5.4 square inches.
Each of these methods of ovepressure prevention is capable of mitigating the t
limiting cold overpressure transient.
4 MILLSTONE - UNIT 3 B 3/4 4-22 Amendment No.
0528
REACTOR COOLANT SYSTEM BASES OVERPRESSURE PROTECTION SYSTEMS (continued)
APPLICABILITY This LC0 is applicable in MODE 4 when any RCS cold leg temperature is s 275'F, in MODE 5, and in MODE 6 when the head is on the reactor vessel. The Pressurizer safety valves provide RCS overpressure protection in the ductile region (i.e.
> 275'F). When the reactor head is off, overpressurization cannot occur.
LC0 3.4.9.1 " Pressure / Temperature Limits" provides the operational P/T limits for all MODES.
LCO 3.4.2.2, " Safety Valves - Operating," requires the OPERABILITY of the Pressurizer safety valves that provide overpressure protection during MODES 1, 2, and 3, and LC0 3.4.2.1, " Safety Valves - Shutdown," requires the OPERABILITY of the Pressurizer safety valves that provide overpressure protection during MODE 4.
Low temperature overpressure prevention is most critical during shutdown when the RCS is water solid, and a mass or heat input transient can cause a rapid increase in RCS pressure when little or no time exists for operator action to mitigate the event.
ACTIONS
~~a. and b.~~
With two or more centrifugal charging pumps capable of injecting into the RCS, or with any SlH pump capable of injecting into the RCS, RC5 overpressurization is possible.
To immediately initiate action to restore restricted mass input capability to the RCS reflects the urgency of removing the RCS from this condition.
Required Action a. is modified by a Note that permits two centrifugal charging pumps. capable of RCS injection for s 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months to allow for pump swaps. This is a controlled evolution of short duration and the procedure prevents _ having two charging pumps simultaneously out of pull-to-lock while both charging pumps are capable of injecting into the RCS.
L in MODE 4 when any RCS cold leg temperature is 1275'F, with one required relief valve inoperable, the RCS relief valve must be restored to OPERABLE status within an allowed outage time (A0T) of 7 days. Two relief valves in any combination of the PORVs and the RHR suction relief valves are required to provide low temperature overpressure mitigation while withstanding a single failure of an active component.
MILLSTONE - UNIT 3 8 3/4 4-23 Amendment No.
0528 4
4 REACTOR COOLANT SYSTEM BASES OVERPRESSURE PROTECTION SYSTEMS (continued)
The A0T in MODE 4 considers the facts that only one of the relief valves is required-to mitigate an overpressure transient and that the likelihood of an active failure of the remaining valve path during this time period is very low.
The RCS must be depressurized and a vent must be established within the following 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months if the required relief valve is not restored to OPERABLE within the required A0T of 7 days.
L The consequences of operational events that will overpressure the RCS are more severe at lower temperatures (Ref. 7). Thus, with one of the two required relief valves inoperable in MODE 5 or in MODE 6 with the head on, the A0T to restore two valves to OPERABLE status is 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .
The A0T represents a reasonable time to investigate and repair several types of relief valve failures without exposure to a lengthy period with only one OPERABLE relief valve to protect against overpressure events.
The RCS must be depressurized and a vent must be established within the following 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months if the required relief valve is not restored to OPERABLE within the required A0T of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .
L.
The RCS must be depressurized and a vent must be established within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months when both required Cold Overpressure Protection relief valves are inoperable.
The vent must be sized 2 5.4 square inches to ensure that the flow capacity is greater than that required for the worst case cold overpressure transient reasonable during the applicable MODES.
This action is needed to protect the RCPB from a low temperature overpressure event and a possible non-ductile failure of the reactor vessel.
The time required to place the plant in this Condition is based on the relatively low probability of an overpressure event during this time period due to increased operator awareness of administrative control requirements.
SURVEILLANCE RE0VIREMENTS 4.4.9.3.1 Performance of an ANALOG CHANNEL OPERATIONAL TEST is required within 31 days prior to entering a condition in which the PORV is required to be OPERABLE and every 31 days on each required PORV to verify and, as necessary, adjust its lift setpoint.
The ANALOG CHANNEL OPERATIONAL TEST will verify the setpoint in accordance with the nominal values given in Figures 3.4-4a and 3.4-4b.
PORV ar.tuation could depressurize the RCS; therefore, valve operation is not required.
MILLSTONE - UNIT 3 B 3/4 4-24 Amendment No.
0528 l
REACTOR COOLANT SYSTEM i
BASES OVERPRESSURE PROTECTION SYSTEMS (continued)
Performance of a CHANNEL CALIBRATION on each required PORV actuation channel is recuired once each REFUELING INTERVAL to adjust the channel so that it responds anc the valve opens within the required range and accuracy to a known input.
The PORY block valve must be verified open and COPPS must be verified armed every 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months to provide a flow path and a cold overpressure protection actuation circuit for each required PORV to perform its function when required. The valve is remotely verified open in the main control room.
This Surveillance is performed if credit is being taken for the PORV to satisfy the LCO.
The block valve is a remotely controlled, motor operated valve. The power to the valve operator is not required to be removed, and the manual operator is not required to be locked in the open position. Thus, the block valve can be closed in the event the PORV develops excessive leakage or does not close (sticks open) after relieving an overpressure transient.
The 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Frequency is considered adequate in view of other administrative controls available to the operator in the control room, such as valve position indication, that verify the PORV block valve remains open.
4.4.9.3.2 Each required RHR suction relief valve shall be demonstrated OPERABLE by verifying the RHR suction valves, 3RHS*MV8701A and 3RHS*M8701C, are open when suction relief valve 3RHS*RV8708A is being used to meet the LCO and by verifying the RHR suction valves, 3RHS*MV8702B and 3RHS*MV8702C, are open when suction relief valve 3RHS*RV8708B is being used to meet the LCO.
Each required RHR suction relief valve shall also be demonstrated OPERABLE by testing it in accordance with 4.0.5. This Surveillance is only required to be performed if the RHR suction relief valve is being used to meet this LCO.
The RHR suction valves are verified to be open every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . The Frequency is considered adequate in view of other administrative controls such as valve status indications available to the operator in the control room that verify the RHR suction valves remain open.
The ASME Code,Section XI (Ref. 8), test per 4.0.5 verifies OPERABILITY by proving proper relief valve mechanical motion and by measuring and, if required, adjusting the lift setpoint.
MILLSTONE - UNIT 3 8 3/4 4-25 Amendment No.
0528
REACTOR COOLANT SYSTEN BASES-OVERPRESSURE PROTECTION SYSTEMS (continued) 4.4.9.3.3 The RCS vent of 2 5.4 square-inches is proven OPERABLE by verifying its open condition either:
a.
Once every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months for a vent valve that cannot be locked open.
.b.
Once every 31 days for a valve that is locked, sealed, or secured in position or any other passive vent path. A removed Pressurizer safety valve fits this category.
This passive vent arrangement must only be open to be OPERABLE.
This Surveillance is required to be performed if the vent is being used to satisfy the pressure relief requirements of the LCO.
4.4.9.3.4 and 4.4.9.3.5 To minimize the potential for a low temperature overpressure event by limiting the mass input capability, all S!H pumps and all but one centrifugal charging pump are verified incapable of injecting into the RCS.
The SIH pumps and charging pumps are rendered incapable of injecting into the RCS through removing the power from the pumps by racking the breakers out under administrative control. - Alternate methods of control may be employed using at least two independent'means to prevent an injection into the RCS. This may be accomplished through any of the following methods: 1) placing the pump in pull to lock (PTL) and pulling its UC fuses; 2) placing the pump in pull to lock (PTL) and closing the pump discharge valve (s) to the injection line, 3) closing the pump discharge valve (s) to the injection line and either removing pover from the valve operator (s) or locking manual valves closed, and 4) closing the valve (s) from the injection source and either removing power from the valve operator (s) or locking manual valves closed.
An SIH pump may be energized for testing or for filling the Accumulators provided it is incapable of injecting into the RCS.
The Frequency of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is sufficient, considering other indications and alarms available to the operator in the control room, to verify the required status of the equipment.
REFERENCES 1.
10CFR50, Appendix G 2.
Generic Letter 88-11 3.
ASME, Boiler and-Pressure Vessel Code,Section III 4.
FSAR, Chapter 15 S.
10CFR50, Section 50.46 6.
10CFR50, Appendix K 7.
Generic Letter 90-06 8.
ASME, Boiler and Pressure Vessel Code,Section XI MILLSTONE - UNIT 3 B 3/4 4-26 Amendment No.
0529'
REACTOR COOLANT SYSTEM BASES 3/4.4.10 STRUCTURAL INTEGRITY The inservice inspection and testing programs for ASME Code Class 1, 2, l
and 3 components ensure that the structural integrity and operational readiness of these components will be maintained at an acceptable level throughout the life of the plant.
These programs are in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR 50.55a(g).
Components of the Reactor Coolant System were designed to provide access to permit inservice inspections in accordance with Section XI of the ASME Boiler and Pressure Vessel Code, 80 Edition and Addenda through Winter.
3/4.4.11 REACTOR COOLANT SYSTEM VENTS Reactor Coolant System vents are provided to exhaust noncondensible gases and/or steam from the Reactor Coolant System that could inhibit natural circulation core cooling.
The OPERABILITY of least one Reactor Coolant System vent path from the reactor vessel head and the pressurizer steam space ensures that the capability exists to perform this function.
The reactor vessel head vent path consists of two parallel flow paths with redundant isolation valves (3RCS*SV8095A, 3RCS*SV8096A and 3RCS*SV80958, 3RCS*SV80965B) in each flow path.
The pressurizer steam space vent path consists of two parallel paths with a power operated relief valve (PORV) and PORV block valve in series (3RCS*PCV455A, 3RCS*MV800A and 3RCS*PCV456, 3RCS*MV8000B).
The valve redundancy of the Reactor Coolant System vent paths serves to minimize the probability of inadvertent or irreversible actuation while ensuring that a single failure of a vent valve, power supply, or control system does not prevent isolation of the vent path.
The function, capabilities, and testing requirements of the Reactor Coolant System vents are consistent with the requirements of Item II.B.1 of NUREG-0737, " Clarification of TMI Action Plant Requirements," November 1980.
1 HILLSTONE - UNIT 3 B 3/4 4-27 Amendment No. JE, 77, UE 0528
2/4.5 EMERGENCY CORE COOLING SYSTEMS BASES 3/4.5.1 ACCUMULATORS The OPERABILITY of each Reactor Coolant System (RCS) accumulator ensures that a sufficient volume of borated water will be immediately forced into the reactor core through each of the cold legs in the event the RCS pressure falls below the pressure of the accumulators.
This initial surge of water into the core provides the initial cooling mechanism during large RCS pipe ruptures.
The limits on accumulator volume, boron concentration and pressure ensure that the assumptions used for accumulator injection in the safety analysis are j-met.
~
The accumulator power operated isolation valves are considered to be
" operating bypasses" in the context of IEEE Std. 279 1971, which requires that 3
bypasses of a protective function be removed autoratically whenever permissive conditions are not met.
In addition, as these sccumulator isolation valves fail to meet single failure criteria, removal of power to the valves is required.
The limits for operation with an accumulator inoperable for any reason except an isolation valve closed minimizes the time exposure of the plant to a LOCA event occurring concurrent with failure of an additional accumulator which may result in unacceptable peak cladding temperatures.
If a closed isolation valve cannot be immediately opened, the full capability of one accumulator is not available and prompt action is required to place the reactor in a mode where this capability is not required.
3/4.5.2 and 3/4.5.3 ECCS SUBSYSTEM The OPERABILITY of two independent ECCS subsystems ensures that sufficient emergency core cooling capability will be available in the event of a LOCA assuming the loss of one subsystem through any single failure consideration.
Either subsystem operating in conjunction with the accumulators is capable of supplying sufficient core cooling to limit the peak cladding temperatures within acceptable limits for all postulated break sizes ranging from the double ended break of the largest RCS cold leg pipe downward, in addition, each ECCS subsystem provides long-term core cooling capability in the recirculation mode during the accident recovery period.
With the RCS temperature below 350*F, one OPERABLE ECCS subsystem is acceptable without single failure consideration and with some valves out of normal injection lineup, on the basis of the stable reactivity condition of the reactor and the limited core cooling requirements.
MILLSTONE - UNIT 3 B 3/4 5-1 Amendment No.
0536
EMERGENCY CORE COOLING SYSTEMS BASES ECCS SUBSYSTEMS (Continued)
The Surveillance Requirements provided to ensure OPERABILITY of each component ensures that'at a minimum, the assumptions used in the safety-analyses are met and that subsystem OPERABILITY is maintained.
Surveillance
= Requirements for throttle. valve position stops and flow balance testing provide assurance that proper ECCS flows will be maintained in the event of a
.0CA. Maintenance of proper flow resistance and pressure drop in the piping system to each injection point is necessary to:
(1) prevent total pump flow from exceeding runout conditions when the system is in its minimum resistance configuration, (2) provide the proper flow split between injection points in accordance with the assumptions used in the ECCS-LOCA analyses, and (3) provide an acceptable level of total ECCS flow to all injection points equal to or above that assumed in the ECCS-LOCA analyses.
Surveillance Requirement 4.5.2.C.2 requires that the visual inspection of the containment be performed at least once daily if the containment has been entered that day and when the final containment entry is made.
This will reduce the number of unnecessary inspections and also reduce personnel exposure.
3/4.5.4 REFUELING WATER STORAGE TANK 4
The OPERABILITY _of the refueling water storage tank (RWST) as part of the ECCS ensures that a sufficient supply of borated water is available for injec-tion by.the ECCS in the event of a LOCA.
The limits on RWST minimum volume and boron concentration ensure that:
!1) sufficient water is available within containment to permit recirculathn cooling flow to the core, and (2) the reactor will remain subcritical in the cold condition following mixing of the RWST and the RCS water volumes with all control rods inserted.except for the most reactive control assembly.
These assumptions are consistent with the LOCA analyses.
The contained water volume limit includes an allowance for water not usable because of tank discharge line location or other physical characteristics.
The limits on contained water volume and boron concentration of the RWST also ensure a pH value of between 7.0 and 7.5 for the solution recirculated within containment after a LOCA. This pH band minimizes the effect of chloride and caustic stress corrosion on mechanical systems and components.
The maximum / minimum solution temperatures for the RWST in MODES 1, 2, 3 and 4 are based on analysis assumptions.
MILLSTONE - UNIT 3 B 3/4 5-2 Amendment No. Jpp osu
Docket No. 50-423 B16636 J
1 Millstone Nuclear Power Station Unit No. 3 Proposed Revision to Technical Specification l
Reduction in the Cold Overoressure Protection System Enablina Temperature l
(PTSCR 3-21-97)
Backaround and Safety Assessment August 1997 4
U.S. Nucl:ar R:guintory Commission B16636\\Attachm:nt 4\\ Pag: 1 BACKGROUND LERs96-038 and 97-030 committed to evaluate the inconsistencies in the MP3 Technical Specifications (TS) related to Cold Overpressurization Protection (COPS) enabling temperature and the Emergency Core Cooling System (ECCS)/ Charging System Mode 3 requirements. This review and subsequent calculations have resulted in the proposed reduction in the COPS Enabling Temperature from 350 F to 275*F As a result, the TS are being changed to incorporate the following:
new heatup and cooldown pressure /
temperature limit curves and PORV setpoint curves and their associated requirements as a result of assumptions made in these analyses; revisions to the Reactor Coolant Loops and Coolant Circulation, ECCS, Boration Systems and COPS to incorporate the lower enabling temperature and new restrictions for COPPS, PORV undershoot and RHR relief valve bellows; addition of a footnote to allow an RCP to substitute for an RHR pump during heatup from Mode 5 to 4 which is consistent with the improved standard technical specifications (STS); rewording 3/4.4.9.3 and its basis section to be consistent with the improved STS; and revision of the affected basis sections to be consistent with the proposed changes.
Heatuo Cooldown Pressure Temperature Limit Curves and PORV Setpoint Curves All components of the RCS are designed to withstand the effects of cyclic loads due to system pressure and temperature changes. This includes the requirements as set forth in 10CFR50 Appendix G to remain within the ASME code limits which prevent non-ductile failure of the vessel. Transients capable of overpressurizing the RCS are categorized as either mass input (e.g., inadvertent SI) or heat input (e.g., RCP startup with a temperature difference between the RCS and steam generators). The requiroments to ensure these transients are bounded are incorporated into the plant TS in the form of heatup and cooldown pressure / temperature (PTT) limit curves and COPS requirements. The COPS system consists of either two PORVs, or one PORV and one RHR relief valve or two RHR relief valves. In addition, restrictions on the number of RCPs, charging and safety injection pumps which may be running during heatup and cooldown are factored into the Appendix G calculations.
The Heatup Cooldown Pressure Temperature Limit Curves and PORV Setpoint Curves have been revised to incorporate the following issues not previously addressed:
1.
The elevation difference between the vessel mid-line and the RCS hot leg. NRC Information Notice (IN) 93-58 identified certain dynamic pressure drop parameters within the core which may not have been considered in the calculation of the RCS heatup and cooldown curves. The potential nonconservatism arises from the fact that the RCS pressure is measured in the RCS Hot Leg while the downcomer region of the vessel is the region most susceptible to non-ductile failure.
2.
The increased instrument and system uncertainties for the measurement of pressure and temperature. The instrument uncertainties previously used in the
i U.S. Nucirr Regul: tory Commission B16636\\Att chm:nt 4\\Pcg3 2 i
MP3 heatup and cooldown curves consisted of a 10'F margin on the RCS temperature and a 60 psi margin on the RCS pressure. Although these margins which were originally provided by Westinghouse were adequate to compensate i
for normal gauge uncertainties they did not include the overall system uncertainties under various operating conditions, in addition, the sub-atmospheric containment operating pressure of 10.6 psi had also not been included in the previous calculations. These additional considerations resulted in a temperature uncertainty of 22'F, a pressure uncertainty of 129 psi, and 10 psi to convert gauge to absolute prossure for sub-atmospheric containment.
3.
The updated reactor vessel chemistry data. The documentation reviews performed in response to the Generic Letter (GL) 92-01, identified additional Copper (Cu) and Nickel (Ni) content values which had not been previously considered in establishing the appropriate chemistry factor as required by Regulatory Guide 1.99, Revision 2.
4.
RCP Operational Restrictions (i.e., zero RCPs during cooldown below 120 F and one RCP between 120 F and 160 F). In order to gain margin on the heatup and cool-down curve, the allowable number of RCPs which can be operated at lower temperatures is being restricted to minimize the pressure drop across the core.
Reducing the pressure drop ac oss the core, makes the pressure at the RHR relief valves close to the pressure at the down-comer region and increases the available margin.
5.
Charging Pump and Safety injection Pump Restrictions. Acceptable results were demonstrated as long as the mass addition was limited to operation of one charging pump. The calculations take into account energy and mass addition transients and the capabilities of any of the above PORV and RHR relief valve combinations. For the PORVs, the available injection capacity from more than one charging pump could cause an overshoot that will allow the RCS pressure to exceed the Appendix G limits. For the RHR relief valves, the additional flow could exceed the relief capabilities of the relief valve (assuming the limiting single failure of one RHR relief valve) and damage the RHR relief valve bellows and exceed the capacity of the discharge piping, in addition to the requirements set forth as a result of the new Appendix G curves, additional restrictions have been placed on RCP starts and operation to address RCP seal and RHR relief pipe and bellows integrity. The RCP seals could be affected if a PORV actuation reduces RCS pressure below 300 psia. At less than 300 psia, the #1 RCP seal may cease to be film riding and result in failure. The #2 seal and the #1 seal leak-off isolation may not be capable of maintaining the integrity of the RCS boundary.
The RHR relief valve bellows could be damaged, possibly causing the valve to stick open, if flow is high enough to cause high back-pressure on the bellows. Since the RHR is considered part of the RCS when it is inservice, damage to the bellows could impact the integrity of the RCS boundary.
I
U.S. Nuclerr R gul tory Commission B16636\\Attcchm:nt 4\\Pcg3 3 These restrictions are as follows:
RCP operation is restricted below 160 F (unless PORVs are isolated).
RCP starts in Mode 4 are only allowed 1) when SG temperature 5 250 F and the RHR relief valves are isolated, or 2) when SG temperature 5 lowest RCS cold leg temperatura.
RCP starts in Mode 5 are only allowed 1) with two or more loops isolated and when the SG temperature 5 lowest RCS cold leg temperature, 2) with a maximum of one loop isolated and the SG temperature 5 50*F above the lowest RCS cold leg temperature and if the RHR relief valves are unisolated, SG temperature must be less than or equal to 200 F.
l Lower COPS Enable Temperature l
NRC Branch Technical Position RSB 5-2, requires that a Cold Overpressurization Protection System (COPS) be armed during low temperature reactor operation to ensure that the 10CFR50, Appendix G requirements are not inadvertently exceeded as a result of anticipated operational occurrences. This system is required to be operable whenever the RCS fluid temperature decreases below a value corresponding to the
(
most limiting vessel RTwoT at the t/4 flaw location, plus 90*F. The current enabling temperature of 350 F (RCS cold leg) was originally obtained from Westinghouse STS and was never adjusted for plant specific characteristics.
This enabling temperature is being lowered to facilitate transition between Modes and address a conflict between the ECCS/ Charging requimments for heat removal and reactivity control and the COPS requirement to prevent overpresssurization from the pumps injecting. Currently, the enabling temperature (Tc<350 F) is reached in Mode 3 because Tc reaches 350 F before Tavg (Mode 4 definition is Tavg<350 F). The ECCS TS require that two trains of ECCS to be operable in Modes 1-3 and one train to be operable in Mode 4 with a footnote requiring disabling of one train when below the COPS enabling temperature. The Charging System has a similar requirement where two trains are required for Modes 1-4 with a footnote requiring disabling of one train when below the COPS enabling temperature. This makes Mode 3/ Mode 4 transition difficult. By lowering the COPS enabling temperature, the TS are simplified and the transition conflict is eliminated.
The fluid temperature corresponding to RTwar (plus 90*F) for the t/4 and St/4 locations was determined to be N4*F (including instrument uncertainties). The COPS enable temperature is proposed to be lowered from the current 350*F to 275'F which provides a margin of 31*F above that required by RSB 5-2 and provides 75 F after the Mode 3/4 transition to isolate one train of ECCS.
4 U.S. Nucirr R::gulatory Commission B16636\\Attrchm:nt 4\\Pego 4 DESCRIPTION OF CHANGES To implement these changes the following specific changes to the TS Sections are being proposed:
3/4.4.9 Pressure / Temperature Limits 3.4.9.1 The current TS heatuo/cooldown curves are based on the brittle fracture limits and the maximum heatup/cooldown rate (100 F in any 1 hr. period) is based on the fatigue limit. However, the calculation of the heatup/cooldown curves is also based upon an assumption of maximum ieatup/cooldown rates. These assumptions are more limiting than the heatup/cooldown rates calculated for fatigue limits. This proposed change removes this fatigue limit maximum rate and now specifies only the brittle fracture heatup curves.
The heatup curve was changed from a curve specifying a constant 60 F/hr rate throughout the operating range to a compound curve specifying a rate of 40 F in any one hour period up to 160 F, and 80 F in any one hour period above 160 F. In order to minimize the effect of the dynamic pressure drop across the core, the restriction of a single RCP in operation at or below 160 F has been imposed during heatup. Above 160 F, there are no restrictions on the number of RCPs in operation, because the curves include a pressure margin corresponding to the AP across the core with four RCPs in service.
The cooldown curve was changed from a curve specifying multiple constant cooldown rates to a compound curve specifying a rate of 80 F in any one hour period above 160 F and 20 or 40 F in any one hour period below 160 F. In order to minimize the effect of the dynamic pressure drop across the core (where the operating window is smallest), the restriction of zero RCPs in operation below 120 F has been imposed during cooldown. From 120 F to 160 F, a maximum of one RCP is permitted during cooldown.
Pressure margins have been included in the P/T curves consistent with these RCP operation restrictions.
LCO 3.4.9.1.c has been added to expand the requirement that only one RCP can be operated to apply at all times when the RCS is less than or equal to 160 F, not just heatup and cooldown.
Temperature fluctuation requirements for hydrostatic and leak tests, renumbered to d, have been clarified to ensure that the 5 F temperature fluctuation is also applicable for the 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months period prior to as well as during the in-service test. This will ensure a uniform reactor vessel temperature during the in-service test.
U.S. Nucirr R:gul: tory Commission B16636\\Atttchm:nt 4\\Paga 5 Action A 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months time requirement has been added for performing an engineering evaluation when the limits have been exceeded. This adds a time limit where previously there was none. The 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months completion time is reasonable to accomplish this evaluation and is consistent with the improved STS. This assures that if the evaluation cannot be completed be in a timely manner, the plant will be placed in a lower and safer mode until this review is
~~complete, l~~
Surveillance Surveillance 4.4.9.1.1 is being revised to include the clarification above on the hydrostatic testing. Table 4.4-5 has been revised to state " approximate' l
withdrawal time. This is a clarification of the existing requirement and it is consistent with the intent of ASTM E-185.
3.4.9.2 A 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months time requirement has been added for performing an engineering evaluation when the limits have been exceeded. This adds a time limit where previously there was none. The 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months completion time is reasonable to accomplish this evaluation and is consistent with the improved STS. This assures that if the evaluation cannot be completed be in timely a manner, the plant will be placed in a lower and safer mode until this review is complete.
3/4.4.9.3 The existing LCO 3.4.9.3 has been replaced in its entirety to make it consistent with the improved STS. Specific wording has been changed for clarification. Most of these rewording changes do not change the intent. The enabling temperature for COPS has been reduced from 350 F to 275 F.
Figures 3.4-4a and b have been replaced with new cuntes based on the new Appendix G analysis.
This specification currently does not address disabling ECCS.
This requirement has been transferred from the ECCS TS to this section and expanded to cover Mode 5, and Mode 6 when the reactor vessel head is on.
The new analysis shows acceptable results for injection from one charging pump only. A footnote clarifies that this restriction on pumps applies whenever COPS is in service and not just when RCS cold leg temperature is below 275 F. The PORV lift setting limits have been revised to the values consistent with this analysis. Mode 3 has been deleted from the mode applicability for COPS since the enabling temperature has been lowered to 275 F.
A footnote is being added to TS 3.4.9.3 to allow 2 charging pumps to be capable of injecting for less than 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months during pump swap operations.
However, at no time will two charging pumps be simultaneously out of pull-to-lock during pump swap operations. Currently if the pumps need to be swapped, the first charging pump is stopped, placed in pull-to-lock and the UC fuse is removed before starting the second pump. With the proposed
)
U.S. Nucirr R:gul:. tory Commission B16G36\\Att: chm:nt 4\\Pcg3 6 change, the second charging pump except for being in pull-to-lock, would be i
capable of injecting prior to stopping the first pump. When the first pump is stopped, it will be placed in pull-to-lock and the second pump would be started.
The intent of this footnote is to ensure one pump will always be operable in the event that RCP seal flow and reactivity control are necessary.
When the RCP is in operation, the charging pump provides the preferred i
method for seal flow because it minimizes seal degradation. The proposed 1
change minimizes the time that this preferred method is interrupted. The fact that the plant is procedurally required to be borated to the highest required boron concentration for Modes 3,4, or 5 prior to entering Mode 4, makes reactivity control only necessary to reestablish the shutdown margin after a boron dilution event.
Actions Requiremerits have been added to immediately initiate actions to limit injection capability to a maximum of one charging pump. These are new actions which previously did not exist and are consistent with the improved STS.
Action b is relabeled as Action d and is being revised to make it clear that 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months are allowed to establish the required vent if the relief valve is not restored to operable status within the required time frame. This is an editorial change and does not effectively change the Action time. Action c is being relabeled as Action e and is being broadened to include any other condition that makes COPS inoperable. Current Action d has been deleted as an action and appropriately reworded as Surveillance 4.4.9.3.3 (verification of RCS venting).
Surveillances Surveillance 4.4.9.3.2 was consolidated into 2 subsections by deleting the specific references to the valve numbers and instead referring to the valves as RHR suction valves.
This simplfies the TS and tm specific valve numbers are provided in the bases for ciarification.
Surveillances 4.4.9.3.4 and 4.4.9.3.5 have been added to verify the charging and SI pumps are incapable of injecting once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . This surveillance is being transferred from ECCS TS 4.5.3.2. Further, the method for disabling the charging pump and Si pumps is being expanded to allow the options currently specified in the bases for TS 3/4.4.9.3. This includes placing the pump in pull-to-lock and pulling its UC fuses, placing the pump in pull-to-lock and closing the pump discharge valve (s), closing the pump discharge valve (s) to the injection line and either removing power from the valve operator (s) or locking manual valves closed or closing the valves (s) from the injection source by either removing power from the valve operator (s) or locking the manual valves closed.
U.S. Nucl:Cr R:gul: tory Ccmmission B16636%ttechm:nt 4\\Pcg3 7 Bases The bases have been replaced in their entirety with wording that is consistent with the improved STS. In general, the revised bases praides greater detail l
about the requirements and provides the basis for the changes being made to TS diswssed above.
The discussu and tables related to reactor vessel material testing and results have bet.n deleted it is replaced with a sentence which states the requirements for ree: tor vessel material irradiated surveillance specimens is in accordance with 100FR50 Appendix H.
The specific equations for calculating the limit curves havo been removed and replaced with references to the applicable standards that are to be followed.
3/4.1.2 Boration Systems The Boratior) Systems flow path and charging pump TS are being revised to change th' mode applicability and actions to be consistent with the new e
lower COPS enabling temperature which was changed to address an inconsistency in TS. The inconsistency in the TS relates to the requirement to initiate COPS in MODE 3 (Section 3.4.9.3) requiring no SI pumps and only one charging pump to be operable versus the requirement to have two ECCS trains (Section 3.5.2) and two boron injection flow paths (Section 3.1.2.4) operable in MODE 3. As a result of this conflict, the plant either violated the MODE 3 or MODE 4 requirements during heatups and cooldowns as reported in LER 96-038-00. The proposed revisions reduce the COPS enabling temperature to 275 *F, allowing the plant to entor MODE 4 and then line up the RCS injection paths prior to initiating COPS.
The new COPS enabling temperature makes the footnote for TS 3.1.2.2 no longer applicable to Mode 3. Since Mode 4 applicability is being relocated from TS 3.1.2.2 & 3,1.2.4 to 3,12.1 & 3.1.2.3, the footnote for TS 3.1.2.2 and Surveillance 4.1.2.3.2 are being deleted.
Previously, only one charging pump was required to be operable when the RCS cold leg temperature was below 350 F. As proposed only one charging pump will be required in Mode
~~4. The proposed change does not effectively change the charging pump requirement. Surveillance 4.1.2.3.2 is being replaced by surveillance 4.4.9.3.5.~~
In addition to opening the pump breakers-(current method specified in ECCS and Boration TS), the proposed surveillance provides additional options (currently specified in TS 3/4.4.9.3 bases) for assuring the inoperability of the Si and charging pumps.
n
U.S. Nucl:Cr R:gulatory Commission B16636\\Att chment 4\\Pcge 8 Bases The basis section is being revised consistent with these changes and is being clarified to state that a second method of boration is not required to be OPERABLE in Mode 4 for single failure considerations based on the stable reactivity condition of the reactor, the emergency power supply requirement for the operable charging pump, and the fact that the plant is procedurally required to be borated to the highest required boron concentration for Modes 3,4, or 5 prior to entering Mode 4. Procedures wili be revised to eliminate a xenon credit currently allowed when calculating shutdown margin for Modes 4 or5.
3/4.4.1 Reactor Coolant loops and Coolant Circulation Restrictions have been added on curting an RCP to prevent RCS pressure transients caused by energy additions from the Secondary Coolant System, which could exceed the limits of 10CFR50 Appendix G or which could cause pressure excursions within the RHR system which could exceed the RHR design pressure.
3/4.4.1.3 The LCO has been revised to be consistent with the new restrictions on RCP starts. A non operating RCP can no longer be considered a backup to an operating RCP because the :onditions could be such that an RCP could not be started once the operating pump trips.
As a result, the LCO has been revised to require two RCPs to be in operation when the Reactor Trip breakers are closed and require an RCP be in operation for the RCS Loop to be considered OPERABLE. This ensures that an unanalyzed heat addition transient will cot be initiated when an RCP is started in a loop required to meet the LCO but is not currently in operation.
The second footnote has been revised and is applicable to LCO 3.4.1.3.
This footnote no longer allows the SG secondary side temperature to be 50 F above the RCS temperature. However, it does allow the start of an RCP above 275'F regardless of the SG secondary side temperature if RHR relief valves are isolated and either the PORV Cold Overpressure Protection System (COPPS) is blocked or the PORV block valves are closed. These requirements are in place to ensure that energy addition transients do not result in a translent which would exceed the capacity of the RHR relief valves or which would cause a PORV undershoot which would challenge the RCP seals. As discussed above, there are no restrictions on RCP cperation above the COPS enabling temperature as long as RHR relief valves are isolated and either the PORV COPPS is blocked or the PORV block valves are closed.
A third footnote has been added that is applicable to the RHR loops. This footnote adds a restriction on Sl and charging pump operability when the RHR system is connected to the RCS This provides added assurance that
U.S. ilucl: r R:gul: tory Commission B1GG3G\\Att: chm:nt 4\\Page 9 l
the RHR piping will not be overpressurized by an inadvertent actuation of an l
Sl or charging pump. This change is considered safe as it is required to l
ensure that the structural integrity of the RHR relief valve bellows is maintained during a design basis mass injection transient.
3.4.1.4.1 The third footnote (*") in 3.4.1.4.1 has been revised to provide more stringent requirements for starting and operating an RCP in Mode 5. RCPs may not be in operation, with the RCS less than 160*F, unless COPPS is l
blocked or the PORV block valves are closed. The footnote provides four conditions that allow the RCP to be started: a) At least one RCP is operating. b) With two or more loops isolated, an RCP can only be started if l
both of the unisolated loops SG secondary side temperatures are less than l
or equal to the RCS cold leg temperature. c) With a maximum of one loop l
isolated and the RHR relief valves isolated from the RCS, an RCP can only I
be started if the secondary side water temperature of each SG is less than or equal to 50*F above the lowest RCS loop temperature. d) With a maximum of one loop isolated and the RHR relief valves unisolated from the RCS, an RCP can only be started if the secondary side water temperature of each SG is less than or equal to 200 F and less than or equal to 50*F above the lowest RCS loop temperature.
A footnote is being added to TS 3.4.1.4.1 to allow for all Residual Heat Removal (RHR) loops to be removed from operation during a planned heatup to Mods 4 when at least one Reactor Coolant System (RCS) loop is operable and in operation and when two additional steam generators are operable as required by LCO 3.4.14.1.b. Currently, there is no allowance for tuming off RHR in Mode 5 specifically for heatup During Mode 5, current heatup practices remove from service the RHR loop not in operation in order to perform isolation valve leak testing and to align it for safety injection mode. Credit is taken for two operable loops with at least 17%
steam generator level to meet the TS requirements.
During the latter stages of Mode 5, the operating RHR cooling loop is removed from service for the same purposes by using a time allowance allowed in the
"" footnote in TS 3.4.1.4.1 that permits the pump to be *de-energized" for up to one hour. However, this was determined to be a misapplication of this footnote. The proposed footnote will substitute an operable and operating RCS loop for RHR during heatup from Mode 5 to 4. During heatup to Mode 4, the first RHR pump will be operable to meet Mode 4 ECCS requirements and can be realigned for heat removal if necessary. The heatup procedure pre-requisite 2.1.1 (OP 3201) requires that four operable RCS loops be filled, swept and vented, with the associated steam generator secondary side water levels greater than 17 percent pr:or to RHR train removal. The RCS loop that is in operation will fulfi!! the function of RHR. Normally during Mode 4, the RHR function is to remove decay heat and circulate the RCS. However, during heatup, decay heat
U.S. Nucl:Cr R:gul tory Commission B1GG3G\\Att: chm:nt 4\\Pcg310 removal is not necessary, and as such the RCS loop in operation can l
substitute for RHR. If decay heat removal is necessary such as during a loss of offsite power (LOP), the plant will either realign RHR or heat up to Mode 4, establith natural circulation and remove decay heat using the RCS loops. The bases have been revised to discuss this change. As stated in the bases, the operating RCS loop can not be credited as one of the two steam generators to satisfy LCO 3.4.1.4.1.b.
3.4.1.4.2 The third footnote (~) in 3.4.1.4.1 has been added to 3.4.1.4.2.
3.4.1.6 LCO c is being added to require all RCPs to be de-energized prior to opening any RCS loop stop valves. This requirement will ensure that loop flow will not be initiated when the loop stop valves are opened. This provides added assurance that the temperature restrictions in TS 3.4.1.3-4 will be met before
}
those pumps are started. Existing LCOs e and d have been renumbered d and e.
Bases The basis section is being revised consistent with these changes and is being clarified to state that these requirements also prevent pressure excursions within the RHR system which could exceed the RHR relief valve bellows design pressure and exceed the capacity of the discharge piping.
3/4 5 3 Emeroency Core Coolino Systems (ECCS) Subsystems -Tavo less than 350 F The footnote allowing one safety injection (SI) pump and one charging pump to be operable when Tc is less than 350 F and the associated surveillance has been deleted. This information is now covered under TS 3.4.9.3 and 3.4.1.3. This TS previously ensured the COPS analysis is bounded and that the RHR piping was protected from overpressurization during an inadvertent Si actuation. These functions are covered in TS 3.4.9.3 and 3.4.1.3 including the surveillances. The second function is necessary when RHR is in service because the RHR relief valves are unable to prevent overpressurization of the RHR piping if more than one charging pump is started. Previously, an RHR interlock provided this protection. However, this interlock was removed as part of an industry initiative to reduce the risk of inadvertent isolation of the RHR system during cold shutdown and refueling operation. As part of the risk review, credit was taken for these pumps being incapable of injecting.
Adding this requirement to TS 3.4.1.3 will insure that these pumps are incapable of injecting between Ta 5350*F and Tc 5275'F when RHR is unisolated.
U.S. Nucl:Cr R:gul: tory Commission B1GG36\\Att: chm:nt 4\\ Peg) 11 TS 3.5.3.f is being changed from *An OPERABLE flow path capable of taking suction from the RWST upon being manually realigned and transferring suction to the containment sump.* to *An OPERABLE flo*> path which, with manual realignment of valves, is capable of dischargin to the RCS, taking sudion from the RWST and transferring suction to the containment sump." The current wording implies that manual realignment is needed only for the suction valve. In order to prepare a second train of RHR for cooldown or to maintain one train of RHR in the ECCS mode while the other train is providing RCS cooling, the discharge valve for the RHR train in the ECCS mode or the RHR crosstle valves which limit flow to two loops, must be closed as well as the RWST suction valve. Since the ECCS mode I
of RHR operation assures injection into all four loops, manual realignment of the discharge and crosstie valves are needed as well as the suction valves in order for the train to function in the 4 loop ECCS injection mode. The bases have been revised to discuss this change.
Surveillance Surveillance 4.5.3.1 has been revised to allow for the different valve alignment allowed in LC0 3.5.3.f Bases The basis section is being revised to remove the discussion on the pump limits. This discussion is included in the bases for TS 3.4.9.3 and 3.4.1.3.
The bases is being clarified to allow for some valves to be out of the normal ECCS injection lineup, consistent with the revision to LCO 3.5.3.f.
SAFETY ASSESSMENT The main purpose of the proposed changes is to update the TS requirements for overpressurization'to reflect new analyses performed to demonstrate compliance with Appendix G. The requirements of 10CFR50 Appendix G provide assurance that a reactor pressure vessel failure will not occur. - These requirements are prescribed to avoid encountering pressure, temperature and temperature rate of change conditions that might cause undetected flaws to propagate and cause non ductile failure of the reactor vessel. The Cold Overpressurization Protection System and the TS restrictions provide assurance that the assumptions and initial conditions assumed in the calculations performed to demonstrate compliance with Appendix G remain bounding.
NRC Branch Technical Position RSB 5-2, requires that a
Cold Overpressurization Protection System (COPS) be armed during low temperature reactor operation to ensure that the 10CFR50, Appendix G requirements are not inadvertently exceeded as a result of anticipated operational occurrences. This system is required to be operable whenever the RCS fluid temperature decreases below a value corresponding to the most limiting vessel RTwor at the t/4 flaw location, plus 90*F. The fluid temperature corresponding to RTwot (plus 90'F) for the t/4 and 3t/4 locations was determined to be 244'F (including p
U.S. Nucl:Cr R:gulttory Commission B1663G\\Attechmsnt 4\\Pcg312 instrument uncertainties). The COPS enable temperature is proposed to be i
lowered from the current 350'F to 275'F which provides a margin of 31*F above that required by RSB 5 2. The reduction of the COPS enabling temperature eliminates the need for COPS to be operable in Mode 3. This will simplify the transition between Mode 3 and Mode 4.
The heatup/cooldovm curves address temperatures above 275'F and overpressure protection is assured by the TS limits and the plant procedures that implement these l
limits. The propose'd changes to the P/T limits reflect the new analyses that l
account for: instn> ment uncertainties associated with the measurement of pressure and temperature; the eleven difference between the vessel mid-line and the RCS hot leg; updated chemdy c@ end the dynamic pressure drop across the core which were previously rvt NAh Ensidered in these limits. As a result, the new proposed curves raise the luv er bound on RCS temperature, resulting in increased RCS ductility and therefore increased structural margin against non-ductile failure. The new curves take into account the dynamic pressure effects identified in NRC Information Notice 93-58 and are calculated in accordance with 10CFR50 Appendix G, ASME Section XI and Regulatory Guide 1.99, Revision 2.
These changes to the P/T limits are reflected in TS 3.4.9.1. Additional restrictions have been placed on RCP operation during heatup and cooldown to ensure that the pressure drop across the core assumptions used in developing the curves remain boundirig. These are also reflected in TS 3.4.1.3, 3.4.1.4.1, 3.4.1.4.2 and 3.4.1.6.
As such, the curves will continue to provide the required assurance for reactor vessel integrity.
Additional changes have been made to the Overpressure Protection TS to ensure that the assumptions made in the Appendix G calculations remain bounding. These include additional restrictions on charging pump and SI pump operability and the modification of the PORV setpoints. The pump requirements have been transferred from the ECCS specification and expanded to cover Modes 4,5 and G. In addition, these same pump restrictions have been included in TS 3.4.1.3 whenever RHR is in service. This provides added assurance that the RHR piping will not be overpressurized by an inadvertent actuation of an Sl or charging pump. Additional actions and surveillances have been provided to assure that assumptions on charging pump and Si pump operability will be met, in addition to opening the pump breakers (current method specified in ECCS and Boration TS), the proposed changes provide additional options (currently specified in TS 3/4.4.9.3 bases) for assuring the inoperability of the Si and charging pumps. These additional options require two distinct operator actions to restore injection capability from these pumps.
Thus, these options are equivalent in providing assurance that an inadvertent injection will not occur while at the same time allowing faster restoration if needed to
[
mitigate a loss of RHR.
Restrictions on starting RCP(s) when zero pumps are running are added to TS 3.4.1.3,3.4.1.4.1, and 3.4.1.4.2 to ensure that energy addition transients do not
U.S. Nucirr R:gul: tory Commission B16636%ttachm:nt 4\\Pcgo 13 result in RCS pressurization in excess of the RHR relief valve capabilities and to ensure that PORV undershoot does not result in RCS pressure lower than that required by the RCP #1 seal when an RCP is running and the PORVs are used for overpressure protection. These changes ensure that tne structural integrity of the RHR relief valve bellows and the RCP seals are not challenged.
A requirement to have all RCPs de-energized, prior to unisolating a loop is added to TS 3.4.1.6.c, to ensure that loop flow will not be initiated which results in an energy addition transient from the secondary side of the SG being unisolated. This change will preclude RCS overpressurization when an idled loop is returned to service and SG secondary side temperature is greater than the RCS temperature.
i Restrictions on temperature mismatch between the RCS and the SG secondary side have been revised to ensure that an unanalyzed energy (heat) addition event does not occur when an RCS loop is placed in operation.
The proposed change to TS 3.4.1.4.1 facilitates the operational transition from Mode 5 to Mode 4 during a plant heatup by allowing RHR to be removed from service to perform surveillances and alignment for ECCS injection. While in Mode 5, one RHR cooling train is currently required to be operable and to be in operation for decay heat removal and boron mixing. The RCS loop while in operation provides equivalent capability for-boron mixing and decay heat l
removal (by heating up to the point where steaming off the SG occurs). Since the heatup procedure pre-requisite requires that all four RCS loops be filled, swept and vented, with the associated steam generator secondary side water levels greater than 17 percent prior to RHR train removal, three steam generators will be available to meet the requirements of LCO 3.4.1.4.1.b.
However, in the event of a LOP, the RCP will trip and the RHR pump will need to be started to maintain Mode 5 conditions. Currently in Mode 5, in the event of a LOP, RHR cooling is re-established when the RHR pump is restarted after the diesel is running. With the proposed change, the RHR valves must be realigned, in addition to starting the RHR pump. Allowing normally open valves to be closed in Mode 5 to align RHR for ECCS injection, introduces ' failure to open' as a potential mechanistic failure malfunction in the RHR system. This is a malfunction of a different type since failure to open of these valves upon a LOP in Mode 5 has not been previously evaluated. However, these valves are stroked in order to establish the ECCS injection lineup. Thus, there is a very high degree of assurance that the valve re-alignment can be accomplished, in addition, since this footnote is only applicable during a heatup, decay heat will be very low. Thus, there is sufficient time to re align RHR for core cooling even if actions outside the control room are necessary. Further, with four RCS loops operable and a bubble drawn in the pressurizer and the RCS pressurized, the steam generators can be used for core cooling via natural circulation once the plant heats up to Mode 4, in the event the RHR cannot be re-established. Procedures will be revised to ensure o
U.S. Nucl:Cr R:gul: tory Commission B1GG3G%tttchm:nt 4\\Pcg) 14 the entry conditions include the loss of a RCP in Mode 5 as well as guidance to realign the RHR valves. The proposed note is consistent with the improved STS.
Based on this review, the change to TS 3.4.1.4.1 has been determined to not involve a significant hazards consideration and is judged to be safe.
The proposed change to TS 3.4.9.3 is being made to facilitate pump swap l
operations and provide an operable charging pump to ensure RCP seal flow and reactivity control will be available. At no time will two charging pumps will be running
)
simultaneously. Requiring at least one pump to be in pull-to-lock assures it will not
(
start on an inadvertent Si and exceed the maximum f;ow assumed in the Appendix G l
l analysis. If an operator were to inadvertently start the second pump, a failure of the charging throttle valve, FCV-121, and one relief valve credited for COPS would be necessary to exceed the assumptions in the Appendix G analysis. The procedure requirements to swap pumps and the likelihood of these multiple failures occurring during the short duration allowed in this footnote, piovide adequate assurance that an overpressurization event will not occur and challenge reactor vessel integrity. A slightly longer duration than currently in the improved STS (15 min. to 1 hr) for completion of this action is being proposed. The proposed increase in duration minimizes the potential for error in executing the swap over and allows for documented verification of the disabling of the required pump. Based on the above, there is no impact on the potential for an overpressure transient in the time interval.
l A clarification of the hydrostatic and leak test requirements ensures a uniform reactor vessel temperature for the test.
A 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months time limit is placed on the performance of engineering evaluations of out of specification condition. This provides added assurance for RPV integrity and is consistent with STS.
The changes also eliminate an inconsistency between the charging system operability requirements for boration and the charging system operability requirements for cold overpressure protection. These are reflected in TS 3.1.2.1, 3.1.2.2, 3.1.2.3 and 3.1.2.4.
The Bases requirement to maintain two charging pumps operable in Mode 4 will be reduced to one charging pump. As stated in the proposed basis section, a second method of boration is not required to be OPERABLE in Mode 4 for single failure considerations based on the stable reactivity condition of the reactor, the emergency power supply requirement for the operable charging pump, and the fact that the plant is procedurally required to be borated to the highest required boron concentration for Modes 3,4, or 5 prior to entering Mode 4. Procedures will be revised to eliminate a xenon credit currently allowed when calculating shutdown margin for Modes 4 or 5. This provides assurance that reactivity control will be maintained and stable while only one charging pump is operable for cold overpressure concems. The additional options for disabling the charging pump provided in TS 3/4.4.9.3 will allow for faster restoration when needed. These proposed changes clarify the TS requirements and provide added assurance that reactor vessel integrity will be maintained.
)
i
U.S Nucl:Cr R:gulatory Commission B1663G%ttschm:nt 4\\PCg] 15 The proposed change to TS 3.5.3.Iis being made to clarify that manual realignment is necessary for the discharge as well as suction valves for the RHR train aligned for l
l the ECCS mode. Since manual realignment is currently specified, this does not l
represent a new potential failure mode. In the event of a loss of offsite power, the RHR operating in the heat removal mode could be realigned from the control l
l room. Adequate time exists for RHR to be capable of performing its ECCS safety function.
In summary, the proposed changes have been reviewed and determined acceptable and safe.
e
Docket "o 50-423 B16636 Millstone Nuclear Power Station Unit No. 3 Proposed Revision to Technical Specification Reduction in the Cold Ovemressure Protection System Enablina Temperature (PTSCR 3-2197)
Sjanificant Hazards Consideration and Environmental Considerations August 1997
U.S. Nucirr R:gul: tory Commission B1GG36\\Att: chm:nt 5\\Pcg31 Slanificant Hazards Consideration NNECO has reviewed the proposed revision ir; sceordance with 10CFR50.92 and has concluded that the revision does not involve a significant hazards consideration (SHC).
The basis for this conclusion is that the three criteria of 10CFR50.92(c) are not satisfied. The proposed revision does not involve a SHC because the revision would not:
l l
1.
Involve a significant increase in the probability or consequence of an accident l
previously evaluated.
1 Probability of Occurrence of Previousiv Evaluated Accidents Since the PORV setpoints and the COPS enabling temperature have been calculated in accordance with 10CFR50, Appendix G and ASME Section XI, the change will not alter the probability that an overpressurization event will result in a loss of RV integrity. The new PORV setpoint curves are lower than the current curves in certain temperature ranges (below approximately 130'F and above approximately 220'F), and therefore the operating window is slightly decreased.
However, the reduced operating window is still sufficient for normal anticipated pressure fluctuations. Below 160*F, operation of Reactor Coolant Pumps are prohibited if the PORVs are armed for COPPS; therefore, PORV actuation will not occur below 160'F when the RCPs are running. In a water solid condition, RCS pressure is maintained via the letdown low pressure control valve, which, when in automatic mode, maintains the RCS pressure in a relatively narrow range. When the RCPs are not running, the PORV COPPS system can be actuated. However, for this condition, the allowable pressure range is 0 to 418 psia. This pressure range is sufficient to accommodato normal anticipated pressure fluctuations.
Above 220'F, the minimum pressure range is from 300 psia to 595 psia; this range is sufficient to accommodate normal anticipated pressure fluctuations. In this temperature range, a pressurizer bubble is normally present, which will minimize any pressure fluctuations, thereby limiting the possibility of a PORV actuation. Based on this, it is concluded that the proposed change will not impact the probability of occurrence that a PORV will be challenged.
When the RHR relief valves art used for COPS there is no credible scenario which would result in excessive relief valve undershoot. This is because these valves are spring loaded relief valves which are designed to close whenever the RCS pressure decreases below the nominal setpoint of 440 psig. This provides assurance that there will be no damage to the seal of a running RCP.
The proposed changes to the heatup/cooldown curves and the reduction in the enabling temperature for COPS only affect operational limits and can not be
U.S. Nucl::r R2gul:: tory Commission B1GG36\\Attcchm:nt 5\\Pcg3 2 initiators of an event. The restrictions on RC, RHR and ECCS pump operation can not result in an event initiator. Two separate operator actions are required to start an ECCS or RC pump. These two necessary actions as well as procedural controls are sufficient to prevent an inadvertent ECCS or RC pump start. De-eneralzing the RCPs when returning a loop to service can not initiate an event.
The proposed change will provide an opsrable charging pump to ensure RCP seal flow and reactivity control will be avatable. When the RCP is in operation, the charging pump provides the preferred method for seal flow. The proposed change minimizes the time that this preferred method is interrupted. A loss of charging pump seal flow will not cause a malfunction of an RCP because the pump is designed to use RCS flow as an alternate method at these conditions. Not allowing two charging pumps to run simultaneously and requiring at least one pump to be in pull-to-lock, assures a second pump will not start on an inadvertent Si and exceed the assumptions in the Appendix G analysis or initiate a Boron Dilution or CVCS Malfunction event. If an operator were to inadvertently start the second pump, a failure of the charging throttle valve, FCV-121, and one relief valve credited for COPS would be necessary to exceed the assumptions in the Appendix G analysis.
In addition, the actual time allowed for swapping the charging pumps is short. The remainder of the hour allows for documented verification of the disabling of the required pump.
The proposed change will not change any control systems for these pumps or alter the system configuration that would affect the probability of an uncontrolled increase in charging flow. The procedure requirements to swap pumps and the likelihood.of these multiple failures occurring during the short duration allowed in this footnote provide adequate assurance that an overpressurization event will not occur. Maintaining at least one pump always operable makes the system more reliable for reactivity control than the current method which disables both pumps simultaneously.
The proposed change to maintain one charging pump operable in Mode 4 can not initiate an event because of the stable reactivity condition of the reador, the emergency power supply requirement for the operable charging pump, and the fact that the plant is procedurally required to be borated to the highest required boron concentration for Modes 3,4, or 5 prior to entering Mode 4. These changes do not effectively change the availability of plant equipment or the way that the plant is operated.
l The proposed change to substitute an RCS loop for an RHR loop during a planned
)
heatup, can not initiate an event. The RCP will be verified as operating properly prior to stopping the RHR pump and as such will not initiate a loss of decay heat removal (by heating up to steam the SGs)/ loss of flow. While the RCP is in operation, it performs the RHR boron mixing function and the decay heat removal function is not required for heatup. Using the RCP to perform this function will not affect the probability that the RCP could fail because it will be operated within its normal operating design conditions.
Aligning RHR in the ECCS lineup will not
U.S. Nucl:Cr R:ptttory Commission B16G36\\Att: chm:nt 5\\Pcg] 3 affect the probability of a RHR pump to start. The pump will be operable in this lineup. Currently in Mode 5, RHR is lost on a LOP and is manually restarted once the diesel is running. With the proposed change, the RCP will be lost on a LOP and the RHR pump will have to be manually started. Thus, the proposed change does not affect the probability that the RHR pump could fail. Since the current response to a LOP is to manually restart the RHR pump, operator action is needed independent of this change. The proposed change allows normally open valves to be closed in Mode 5 to align RHR for ECCS injection. This introduces additional manual actions which could extend the time required to establish flow. In addition, if one l
diesel generator were to fail, manual operation of a valve in the ESF building would be necessary. The mechanistic ' failure to open' of valves that is introduced by the change as well as the need for manual operator action to realign these valves increases the time to establish heat removal. However, there is sufficient time to re-establish RHR because this note applies only for a heatup in which the plant will have been shutdown for at least several hours which causes decay heat to be low l
(as compared to high decay heat immediately following a plant trip). Thus, it is concluded that there is no impact on the probability of failure of RHR to perform its required func!!on.
The proposed change to the ECCS wording does not result in any new failure modes that could initiate an event since manual realignment from the control room is currently allowed. Nor can the manual alignment of RHR valves initiate an event because this alignment is only for accident mitigation.
Therefore, the proposed changes do not increase the probability of occurrence of previously evaluated accidents.
l l
)
U.S. Nucl:cr R:gul: tory Commission B1663G\\Atttchm:nt 5\\Pcg3 4 Consecuences of Previousiv Evaluated Accidents The revised Pressure / Temperature curves were calculated in accordance with 10CFR50, Appendix G, ASME Section XI, and Regula'.ory Guide 1.99, Revision 2.
This provides assurance that an inadvertent overpressurization event will not result in a loss of RV integrity. The restrictions on RCP operation and the requirement to de-energize the RCPs in Modes 5 and 6 when retuming a loop to service are consistent with the assumptions made in this Appendix G analysis and the RCPs are not required for accident mitigation for any previously evaluated accidents and therefore do not affect the consequences.
The COPS relieving capability is greater than the maximum RCS pressurization rate resulting from any allowed pump combinations, and the PORV setpoints have been adjusted to take into account instrumentation effects. This will provide assurance that COPS will continue to perform its safety function. Since the COPS enabling temperature has been demonstrated to be conservative at 275*F, allowing SI pump operability above 275'F will have no impact on vessel non-ductile failure.
i The restriction between 275*F and 350 F on the Si and charging pumps, has been appropriately moved to the reactor coolant loop section to provide protection for the RHR system (RCS protective boundary) and to the cold overpressure protection section to provide protection for the RHR relief valves and the RCP seals. By incorporating this requirement previously located in the ECCS TS, RCS integrity is ensured.
With the RCS tess than 160'F, the consequences of the PORV undershoot from the proposed PORV setpoints are that the RCS pressures may drop below the minimum requirement for RCP seal integrity. However, no seal damage will occur since a requirement has been added prohibiting the operation of RCPs below 160*F with the PORVs not isolated while in the low setpoint mode. With cold overpressure relief valves in service above the COPS enable temperature (275'F), restrictions are placed on the startup of an RCP and the number of ECCS pumps capable of injecting into the RCS to prevent unacceptable mass or energy addition transients. This provides assurance that the RHR relief valve capacity will not be exceeded and that PORV undershoot will not challenge the RCP #1 seal. The restriction on the maximum number of ECCS pumps ensures that the integrity of the RHR relief valve bellows and the RCP seals during mass injection transients (i.e., inadvertent SI).
The restrictions on RCS/SG secondary side temperature mismatch ensure that an unanalyzed energy addition event does not occur when an RCS loop is placed in operation.
The consequences of a small break LOCA in COPS Mode 4 are not affected because the plant will continue to maintain one charging pump operable in Mode
U.S. Nucirr R:guttory Commission B16G36\\Attrchment 5\\Pcg] 5 4.
In addition, additional options are provided in the bases of TS 3/4.4.9.3 for disabling the required charging and Si pumps that will allow faster restoration if required to mitigate a LOCA or loss of RHR in Modes 4,5 and 6.
An RHR pump will remain available in Mode 4 with manual realignment from the control room as required to perform its ECCS safety function. The changes have no impact on the cspability of RHR to function in the ECCS mode. RHR is credited during a safety grade cold shutdown. The proposed change assures that the RHR system will be available to perform its heat removal function during a safety grade cold shutdown and thus, there is no change in the analysis assumptions or consequences.
l The changes also eliminate an inconsistency between the charging system operability requirements for boration and the charging system operability requirements for cold overpressure protection. The requirement to maintain two charging pumps operable in Mode 4 will be reduced to one charging pump. As stated in the proposed basis section, a second method of boration is not required to be OPERABLE in Mode 4 for single failure considerations based on the stable reactivity condition of the reactor, the emergency power supply requirement for the operable charging pump, and the fact that the plant is procedurally required to be borated to the highest required boron concentration for Modes 3,4, or 5 prior to entering Mode 4. This provides assurance that reactivity control will be maintained and stable while only one charging pump is operable for cold overpressure concerns.
These changes do not effectively change the availability of plant equipment or the way that the plant is operated. The changes will not adversely impact the assumption for the limiting dilution flow path and flow rate and therefore, the consequences of a boron dilution event are not affected.
The proposed changes will maintain a charging pump operable for reactivity control while ensuring that the flow limits in the Appendix G analyses are not exceeded.
Remaining within the bounds of the Appendix G limits ensures reactor vessel integrity in Mode 4. Since the change maintains the reactor vessel integrity, it does not introduce any means of releasing radionuclides post-accident.
The consequences of a sma!! break LOCA in Mode 4 are not affected because the plant will continue to maintain one charging pump operable in Mode 4. These changes are reflected in TS 3.1.2.1, 3.1.2.2, 3.1.2.3 and 3.1.2.4.
Adequate protection is provided for reactor vessel integrity while maintaining reactivity control operability.
In Mode 5, RHR requirements are specified for decay heat removal in the case of a locs of offsite power bu' none are specified for ECCS accident mitigation. The first RHR train will be aligned for injection prior to taking the second train out of service.
This provides assurante that this train will be available if needed in Mode 5.
Currently in Mode 5, following a LOP the RHR system can be re-established by restarting the RHR pump once the diesel is running. No valve manipulations are
U.S. Nuclect R:gul: tory Commission B16636\\Atttchm:nt 5\\Pcgo 6 necessary. With the proposed change, when the operating RCP trips following a LOP, some of the RHR valves must be realigned from the ECCS to heat removal mode, if one diesel generator were to fail, manual operation of a valve in the ESF building would be necessary.
Since this footnote is only applicable during a heatup, decay heat will be low. There is sufficient time to re-establish RHR even if action outside the control room is necessary Since there are four operable RCS loops, a bubble drawn in the pressurizer and the RCS pressurized, the l
plant will heat up to Mode 4 and natural circulation will provide core cooling if the RHR system cannot be re-established. Thus, decay heat removal is assured and there is no affect on the consequences of a LOP.
Since the structural integrity of the RCS is maintained and adequate core cooling and reactivity control will be available for design basis events, the l
proposed changes will have no adverse impact on the consequences of previously evaluated accidents.
Therefore, the proposed revision does not involve a significant increase in the l
probability or consequence of an accident previously evaluated.
2.
Create the possibility of a new or different kind of accident from any accident previously evaluated.
The temperature / pressure limits will continue to meet the requirements of 10CFR50, Appendix G. Since the new limits continue to provide assurance of reactor vessel integrity, the proposed change does not create the possibility of an accident of a different type than. previously evaluated.
Adequate RCS pressure-relieving capabilities will continue to be maintained throughout the shutdown modes. No new malfunctions will be introduced which could result in a new accident postulated in Modes 3 5.
i The restrictions on RCP operation do not create the potential for unanalyzed heat injection transient as a result of an inadvertent RCP start because two operator actions are required to start a pump. The requirement to have all RCPs de-energized, prior to unisolating a loop adds additional assurance that an energy addition transient will not occur.
l The proposed change to allow 2 charging pumps to be operable does not create an accident of a different type because there will be adequate controls to ensure that the second pump does not inadvertently start and initiate an increase in RCS inventory or a boron dilution. Procedural controls will minimize the amount of time that both charging pumps are operable and at no time will two pumps be out of pull-to-lock.
The proposed footnote to TS 3.4.1.4.1 to remove RHR heat removal from operation allows normally open valves to be closed in Mode 5 to align RHR for 4
,e,-
U.S. Nucirr R:gulatory Commission B1GG36\\Att: chm:nt 5\\Pcgo 7 ECCS injection. This iritroduces ' failure to open' as a potential mechanistic failure malfunction in the RHR system. This is a malfunction of a different type since previously stroking of these valves was not needed to establish RHR. The current response to a LOP is to manually restart the RHR pump only, with no valve manipulations required. The proposed change adds the ri viual action of realigning the valves. Since operator action to re-establish RHR following a LOP is required independent of the proposed changes, crediting operator action does not create the potential for a malfunction of a different type. Allowing both trains of RHR to be out of service does not create a different accident because additional requirements have been specified for RCS loop operability and at least one RHR pump is operable for ECCS when the core cooling requirement is being met by crediting RCS loop operability Meeting the Mode 4 TS conditions prior to heatup, ensures two diesels are operable. As such, a single failure would only require one valve to be manually realigned in the ESF building. Adequate time is available to accomplish l
these actions since this note only applies during heatup, when decay heat is very low. Further, with four RCS loops operable and a bubble drawn in the pressurizer and the RCS pressurized, the steam generators can be used for core cooling via natural circulation once the plant heats up to Mode 4, in the event the RHR cannot be re-established. Since core cooling will be assured if a LOP occurred during heatup in Mode 5, the change in plant response to this event does not constitute an accident of a different type.
The proposed changes to TS 3.5.3. f to manually realign the ECCS valves is no different from what is currently evaluated.
During a Mode 4 LOCA adequate procedural guidance is provided to er sure that RHR will be realigned for injection.
The proposed change allows RHR to be aligned to perform its safety grado cold shutdown heat removal function.
Therefore, the proposed revision does not create the possibility of a new or different kind of accident from any accident previously evaluated.
3.
Involve a significant reducticn in a margin of safety.
The new proposed curves raises the lower bound on RCS temperature, resulting in increased RCS ductility and therefore increased structural margin against non-ductile failure. The now curves take into account the dynamic pressure effects identified in NRC Information Notice 93-58 and are calculated in accordance with 10CFR50 Appendix G, ASME Section XI and Regulatory Guide 1.99, Revision 2.
These changes to the P/T limits are reflected in TS 3.4.9.1. Additional restrictions have been placed on RCP operation to ensure that assumptions used in developing the curves remain bounding. These are also reflected in TS 3.4.1.3, 3.4.1.4.1.,
3.4,1.4.2 and 3.4.1.6. As such, the curves will continue to provide the required assurance for reactor vessel integrity.
U.S. Nucirr Regul: tory Commission B1GG3G\\Att: chm:nt 5\\Pcga 8 The COPS enable temperature is proposed to be lowered from the current 350*F to 275'F which provides a margin of 31*F above that required by NRC Branch Technical Position RSB 5-2. The reduction of the COPS enabling temperature eliminates the need for COPS to be operable in Mode 3. This will simplify the transition between Mode 3 and Mode 4.
Additional changes have been made to the Overpressure Protection TS to ensure that the assumptions made in the Appendix G calculations remain bounding. These include additional restrictions on charging pump and SI pump operability and the modification of the PORV setpoints. The pump requirements have been transferred from the ECCS specification and expanded to cover Modes 4,5 and 6. In addition, these same pump restrictions have been included in TS 3.4.1.3 whenever RHR is in service. This provides added assurance that the RHR piping will not be overpressurized by an inadvertent actuation of an Si or charging pump. Additional actions and surveillances have been provided to assure that assumptions on charging pump and Si pump operability will be met. The additional options for assuring the inoperability of the Si and charging pumps require two distinct operator actions to restore injection capability from these pumps. Thus, these options are equivalent in providing assurance that an inadvertent injection will not occur while at the same time allowing faster restoration if needed to mitigate a loss of RHR.
A requirement to have all RCPs de-energized, prior to unisolating a loop is added to TS 3.4.1.6.c, to ensure that loop flow will not be initiated which results in an energy addition transient from the secondary side of the SG being unisolated. This change will preclude RCS cverpressurization when an idled loop is returned to service and SG secondary side temperature is greater than the RCS temperature.
The PORV setpoints were established to ensure that the P/T limit curves are not exceeded as a res0lt of a single operator action or as a result of a single equipment malfunction, as required by the current system design basis criteria (i.e., SRP Branch Technical Position RSB 5-2).
A clarification of the hydrostatic and leak test requirements ensures a uniform reactor vessel temperature for the test.
A 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months time limit is placed on the performance of engineering evaluations of out of specification condition. This provides added assurance for RPV integrity.
The changes also eliminate an inconsistency between the charging system operability requirements for boration and the charging system operability requirements for cold overpressure protection. These are reflected in TS 3.1.2.1, 3.1.2.2, 3.1.2.3 and 3.1.2.4.
Tne Bases requirement to maintain two charging pumps operable in Mode 4 will be reduced to one charging pumo. As stated in the proposed basis section, a second method of boration is not required to be OPERABLE in Mode 4 for single failure considerations based on the stable
U.S. Nucl::r R:gul: tory Commission B1GG3G\\Att: chm:nt 5\\Pago 9 reactivity condition of the reactor, the emergency power supply requirement for the operable charging pump, and the fact that the plant is procedurally required to be borated to the highe' t required boron concentration for Modes 3, 4, or 5 prior to s
entering Mode 4. This provides assurance that reactivity control will be maintained and stable while only one charging pump is available. The additional options for disabling the charging pump (provided in the bases for TS 4.4.9.3.5) will allow for faster restoration when needed while maintaining two distinct operator actions to prevent a second pump from being started. This provides added assurance that j
i reactor vessel integrity will be maintained.
J l
Procedures will minimize the amount of time that both charging pumps are operable and having at least one pump in pull-to-lock will ensure that the second pump does l
not inadvertently start and exceed the Appendix G analysis limits and thus, ensure l
reactor vesselintegrity.
The TS bases for requiring RHR in Mode 5 is to remove decay heat and provide RCS circulation. Since the RCP can perform the RHR circulation function and the decay heat removal function is not required during heatup, the proposed change is consistent with the bases. Since this option is only allowed during heatup where decay heat is low, sufficient time will be available to re-establish RHR heat removal as required to mitigate a LOP in Mode 5. Further, with the RCS pressurized, four RCS loops operable and the SG filled, core cooling can be accomplished by the steam generators via natural circulation once the plant heats up to Mode 4, in the event that RHR cannot be re-established. Therefore, the design basis analyses remain limiting and the margin of safety is not reduced.
The criginal plant design allows the RHR pumps to be available for both heat
'emoval while shutdown and ECCS. As such, an allowance, TS 3.5.3.f, was provided to allow manual realignment from heat removal to ECCS mode. The specific wording of TS 3.5.3.f implies that this realignment only involves the suction valves. Since discharge valves must also be realigned, the TS is being reworded to apply for the discharge as well as suction valves. Therefore, this change is a clarification of the existing TS.
The proposed changes do not impact the protective boundaries (reactor vessel integrity) nor any of the design basis accidents.
Therefore, the proposed revision does not involve a significant reduction in a margin of safety.
In conclusion, based on the information provided, it is determined that the proposed revision does not involve an SHC.
U.S. Nucl:Cr R:gul: tory Commission B16636\\Att chment 5\\Pcg310 Environmental Considerations NNECO has reviewed the proposed license amendment against the criteria of 10CFR51.22 for environmental considerations. The proposed revision does not involve a SHC, does not significantly increase the type and amounts of effluents that may be released offsite, nor significantly increase individual or cumulative occupational radiation exposures. Based on the foregoing, NNECO concludes that the proposed revision meets the criteria delineated in 10CFR51.22(c)(9) for categorical exclusion from the requirements for environmental review.

ML20217R299

Text

Navigation menu

Search