ML20207B250
| ML20207B250 | |
| Person / Time | |
|---|---|
| Site: | Palisades  |
| Issue date: | 03/01/1999 |
| From: | CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.) |
| To: | |
| Shared Package | |
| ML18068A534 | List: |
| References | |
| NUDOCS 9903080054 | |
| Download: ML20207B250 (200) | |
Text
PCS Pressure SLs B 2.1.2 BASES SAFETY LIMITS The maximum transient pressure allowable in the PCS pressure vessel under the ASME Code, Section III, is 110% of design pressure. The maximum transient pressure allowable in the PCS piping, valves, and fittings under 120% of design pressure (Ref. 6). The most limiting of these two allowances is the 110% of design pressure; therefore, the SL on maximum allowable PCS pressure is established at 2750 psia.
RhI p.0 M APPLICABILITY SL 2.1.2 applies in MODES 1, 2, 3, 4, 5, and 6 because this SL could be approached or exceeded in these MODES Jue to overpressurization events. 'he SL 1: 2pp!!:21: fn N0DE.6 ba4W 4L reca. b r t/ d A etb:::n: the reactor vessel head closure bolts =13 :st k had in 34cil<d a nd less than fully tensioned th:t th: oCS f Q=+ A k, ed"g it p:: ib!:
M:+41L See ena ouwfitssunpk esen+
SAFETY LIMIT The following SL violation responses are applicable to the VIOLATIONS PCS pressure SLs.
2.2.2.1 If the PCS pressure SL is violated when the reactor is in MODE 1 or 2, the requirement is to restore compliance and be in MODE 3 within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
/ With PCS pressure greater than the value specified in I SL 2.1.2 in MODE 1 or 2, the pressure must be reduced to below this value. A pressure greater than the value j specified in SL 2.1.2 exceeds 110% of the PCS design i pressure and may ' challenge system integrity.
The allowed Completion Time of 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> provides the operator time to omplete the necessary actions to reduce PCS pressure by terminating the cause of the pressure increase,
. removing mass or energy from the PCS, or a combination of l these actions, and to establish MODE 3 conditions.
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RCS Pressure SLl(MitAl B 2.1.2 I
BASES (continued)
APPLICABILITY SL 2.1.2 applies in MODES 1, 2, 3, 4, and because this SL could be approached or exceeg g,in these MODES due to overtressurization events. lThe SL is not pplicable in ,
' MODE 6 ecause the reac% r vessel head c sure bolts are o1 t fully ightened, makincf it unlikely tha the RCS can be [
/ pres urized.l SAFETY LIMIT The following SL violation responses are applicable to the VIOLATIONS (3CSpressureSLs.
0 lh in % b u44 & Mc4ep -
2.2.2.1 hee.d ms4 aid and t Itac.hir fcMel Clc6Wc kold eso %n
%II'I hASicned If the 5 pressure SL is violated when the reactor is in MODE 1 or 2, the requirement is to restore compliance and be lh j 4ns 9ch,nia l for in MODE 3 within I hour, a n C W rfec g3pe; 4.fon geni hLL CMU , AL& ugh With S pressure greater than the value specified in SL 2.1.2 in MODE 1 or 2, the pressure must be reduced to lh OutrM.Munph 66 of the Pc5 13 m A m M , o nce tne r
- H r below this value. A pressure greater (293Lthe specified in SL 2.1.2 exceeds 110% of the value g S' design lh-VW'l hocd i5 /tmotAd , fM, pressure and may challenge system integrit h
,Y M* O 6 blJ 3 L The allowed Completion Time of I hour provides the operator C ppfj Go e fwL time to complete the necessary actions to reduce S pressure by terminating the cause of the pressure increase, lh
'S lh ths /tceC4tf C.Lt fu (ps,) 6 (un Cn c <. removing mass or energy from the 5, or a combination of these actions, and to establish ODE 3 conditions.
lh P 1 (LfnWtd Nrsm % ha.chG YAs R virc.ntnd of 2.2.2.2 fv Q (-h SL 2. 2 Ap (n ger If the OCS pressure SL is exceeded in MODE 3, 4, RCS U/
U pressure must be restored to within the SL value wit in MPd. 5 minutes. g j Exceeding the 5 pressure SL in MODE 3, 4,9r 6)is potentially more severe than exceeding this SL in MODE 1 h
or 2, since the reactor vessel temperature may be lower and j the vessel material, consequently, less ductile. As such, pressure must be reduced to less than the SL within 5 minutes. This action does not require reducing MODES, since this would require reducing temperature, which would (continued) .)
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CEOG STS B C5% Rev 1, 04/07/95 !
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CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS l I
RESPONSE TO THE DECEMBER 04, 1998 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.1, REACTIVITY CONTROL SYSTEM NRC REQUEST:
3.1-01 ITS 3.1.1 Shutdown Margin (SDM)
! LC0 3.1.1, SR 3.1.1.1 and SR 3.1.1.2 I DOC A.6 and DOC M.2 JFD 7 and JFD 9 ITS LC0 3.1.1 states that "SDM shall be within limits," without referring to a COLR or explicitly stating the SDM limits. The ITS 3.1.1 limits and their i applicability are defined in SR 3.1.1.1 and SR 3.1.1.2. TSTF-9 revised the STS from having the limits explicitly stated in the LC0 to referencing the j COLR in the LCO.
Coment: The ITS uses an unacceptable and cumbersome method to define LC0 4 limits. Recommend either including the limits in the LCO, thereby enabling
- the use of only one SR, or utilizing the COLi. as is done with other
- specifications.
Cgasumers Enerav ReSD0nSO:
J l The specific values for SDM have been removed from the CTS and re'ocated to !
! the COLR consistent with NUREG-1432 as modified by TSTF-9. By adopting TSTF-9, the cumbersome method of stipulating the limits for Shutdown Margin in multiple LCOs, or multiple Surveillance Reg Jirements, has been eliminated.
Justification for this change, as well as the related conforming changes, are 3 provided in the "Affected Submittal Pages" listed below. Reference to l l SR 3.1.1.2 in Section 3.3, " Instrumentation" will be deleted as part of Consumers Energy response to NRC's Request for Additional Information related to Section 3.3.
! As a result of relocating the SDM limits to the COLR, a revision has been made
- to Discussion of Change (DOC) 3.1.1, A.8.
This revision supersedes the response previously submitted by Consumers Energy to NRC RAI 5.6-02.
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l CONVERSION TO IMPROVED TECHNICAL SPCCIFICATIONS RESPONSE TO THE DECEMBER 04, 1998 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.1, REACTIVITY CONTROL SYSTEM Affected Submittal Paaes:
Att 1 ITS 3.1.1, page 3.1.1-1 Att 1 ITS 3.1.1, page 3.1.1-2 Att 2 ITS 3.1.1, page B 3.1.1-3 Att 2 ITS 3.1.1, page B 3.1.1-5 Att 2 ITS 3.1.1, page B 3.1.1-6 Att 3 CTS, page 3-50 (ITS 3.1.1 page 1 of 2)
Att 3 DOC 3.1.1, page 2 of 6 Att 3 DOC 3.1.1, page 3 of 6 Att 3 DOC 3.1.1, page 4 of 6 Att 3 DOC 3.1.1, page 5 of 6 Att 3 DOC 3-.l.1, page 6 of 6 Att 4 NSHC 3.1.1, page 1 of 3 Att 4 NSHC 3.1.1, page 2 of 3 Att 4 NSHC 3.1.1, page 3 of 3 Att 5 NUREG, page 3.1-1 Att 5 NUREG, page 3.1-1 Insert Att 5 NUREG, page B 3.1-4 Att 5 NUREG, page B 3.1-5 Att 5 NUREG, page B 3.1-6 Insert Att 6 JFD 3.1.1, page 2 of 3 Att 1 ITS, page 5.0-25 Att 3 CTS, page 6-20 l Att 3 DOC 5.0, page 2 of 7 Att 5 NUREG, page 5.0-21 Att 5 NUREG, page 5.0-21 Insert l
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SDM-l :3.1.1
!'i 3.1 REACTIVITY CONTROL SYSTEMS :
3.1.1 SHUTDOWNMARGIN(SDM)
% fi ol' LC0 3.1.1 SDM s' hall be within#11mitsy $kgg) . jn ihr. bd, f i r L
! APPLICABILITY: MODE 3, 4, and 5. ,
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a ACTIONS -
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CONDITION REQUIRED ACTION COMPLETION TIME l a
! A. SDM not within limit. A.1 Initiate boration to 15 minutes i i restore SDM to within +
! limit.
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- SURVEILLANCE REQUIREMENTS !
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- SVRVEILLANCE FREQUENCY j l i SR 3.1.1.1 l-------------- ----NOTE-------- -------- ~ 1 )
! Only require to be met in MOD 3 when T ,, $yol j l 1s a 525*F d four primary olant pu s i N i
, are operat g.
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I Verify SDM $/2 f.0% ao.\ 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> z
to be todhin 2+d5, !
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i Palisades Nuclear Plant
- 3.1.1 Amendment No. 01/20/98 [
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! -SOM 3.1.1 y .
SURVEILLANCE REQUIREMENTS / p SR 3.1.1.2
---------NOTE-------------------
Only reg red to be met in MODE 3 when T,,, g
- is 2 5 *F with less than four primary coola pumps operating, in MODE 3 when T ,, $,10 is < 25 F, and in MODES 4 and 5.
erify SDM is 2 3.5% ap. 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> i
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Palisades Nuclear Plant 3.1.1-2 Amendment No. 01/20/98 3-b
SDM B 3.1.1 -
r BASES APPLICABLE In addition'to the limiting MSLB transient, the SDM 4 SAFETY ANALYSES requirement for MODES 3 and 4 must also protect against an (continued) inadvertent boron dilution; (Ref. 3) and an uncontrolled control rod bank withdrawal from subcritical conditions (Ref.5). ,
< Each of these events is discussed below.
1 In the boron dilution analysis, the required SDM defines the reactivity difference between an initial subcritical boron l concentration and the corresponding critical boron 4 concentration. These values, in conjunction with the configuration of the PCS and the assumed dilution flow rate, directly affect the results of the analysis. This event is
!' most limiting at the beginning of core life when critical boron concentrations are highest.
The withdrawal of a control rod bank from subcritical
.4 conditions adds reactivity to the recctor core, causing both
- the core power level and heat flux to increase with i corresponding increases in reactor coolant temperatures and ,
pressure. The withdrawal of control rod banks also produce i
, a time dependent redistribution of core power. i
- Depending on the system initial conditions and reactivity j insertion rate, the uncontrolled control rod banks
< withdrawal transient is terminated by either a high power )
trip or a high pressurizer pressure trip. In all cases. l power level, PCS pressure, linear heat rate, and the DNBR do 1 not exceed allowable limits.
i SOM satisfies Criterion 2 of 10 CFR 50.36(c)(2). I LC0 The MSLB (Ref. 2) and the boron dilution (Ref. 3) efts. I are the most limiting analyses that establish the DM fv ni tb LCO. For MSLB accidents, if the LC0 is violated,
{ !
there is a potential to exceed the DNBR limit and to exceed i 10 CFR 100, " Reactor Site Criteria," limits (Ref. 4). For I I
the boron dilution accident, if the LCO is violated, then the minimum required time assumed for operator action to terminate dilution may no longer be applicable.
' SDM is a core physics design condition that can be ensured T I Et4-kM s throughxontrol rod positioning (regulating and shutdown g l
, rods) and through the soluble boron concentration. 1 l
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SDM i 7
B 3.1.1
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V BASES i
SURVEILLANCE SR i
3.l.1.1and_dR 3 ' y )
A I REQUIREMENTS SDM is verified by p:rf:rir.; a reactivity balance l calculation, considering the listed reactivity effecth: )
- b. Control rod positions;
- c. PCS average temperature; i
- d. Fuel burnup based on gross thermal energy generation;
- e. Xenon concentration; and
- f. Isothermal Temperature Coefficient (ITC). )
i Using the ITC accounts for Doppler reactivity in this calculation because the reactor is subcritical and the fuel temperature will be changing at the same rate as the PCS.
Samarium is not considered in the reactivity analysis since the analysis assumes that the negative reactivity due to l Samarium is offset by the positive reactivity of Plutonium i build in. l To mainta i consistency w'th the assumptions up6d in the)
MSLB anal sis, two value of SDM are specifief in the !
hl S 'D A
surveill nce reouiremen i
,. Q, l SR 3.1.1.1 requires SDM to be la v/ ocf. ljThis SDM value )
ensures the con quences or an PDLu as well as the oth r events descri d in the Applicable afety Analysis, wi 1 be l
. acceptable w never the plant is n MODE 3 with T,,, 3 525"F l and four Pr' ary Coolant Pumps cps) are in operat n. y Therefore SR 3.1.1.1 is modif' d by a Note which ly requires his SR to be met in ODE 3 when T.,, 2 5 "F and four PC are in operation.
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SDM i B 3.1.1 3
BASES gel SURVEILLANCE SB_J.1.1.1(and_SR/3. --
(continued) X REQUIREMENTS i
IR 3.1/1.2 reauites SDM to Isb > 3.5% 401 This_ SDM valu4 - y
- ensures _the Ionsequences o an MSL eli as the'other.)
fev'erits_ ' describe , n _ ej i afet Analysis wiiT"he acceptable as a ~ result o a coo own t ich adds
_ positive reactivity in_the presence of a negative moderatgr temuerature coetticientW As such, the requirements of 4hiA y i Sg3.1.2 1must be met whenever the plant is in fMODE 3 ith
,,,, 525"P, ~ 1tn less than our FCPs oper ting, MODE 3 ith p
- T., < 525'F and MODES 4 a d 5. Therefo , SR 3.1.1. is m ified b a Note which nly requires is SR to b met in E 3 wi' T,,,
- t 525'F ith less than our PCPs o ratina.
DE 3 wihh T.m < 525*F anc MODES 4 and 5. g The Frequency of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> for the verification of SDM is aased on the generally slow change in required boron concentration, ar.d also allows sufficient time for the operator to collect the required data, whichdnclud q g performing a boron concentration analysis, and[complett the calculation. 9 REFERENCES 1. FSAR, Section 5.1
- 2. FSAR, Section 14.14
- 3. FSAR, Section 14.3
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- 4. 10 CFR 100
- 5. FSAR, Section 14.2 <
Palisades Nuclear Plant B 3.1.1-6 01/20/98 3-e
SL%n w SeAcunTl Care.d. 5thDns A.\ /
3.;o ~~ RONTRot ACO AND POWER DISTRIRuTLQN LIMITS)
'Acelicabilith .
Applies to operation of CONTROL R005 a hot channel factors during l operation 4[ j 1
Obiecti l l
To see ify limits of CONTROL R00 ement to assure an acceptable power )
distr Dution during power operatio , limit worth of individual r s to i valu s analyzed for accident cond tions, maintain adequate shut wn %g mar in after a reactor trip and specify acceptable power li ts for y 90 er tilt cond1tions. - .
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Scecifications Lh iI 9 I
- 3. (0.1 Shutdewn Marain Raouirements
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fo r prM W coolant pumps in operation)(at het shutdown and,'k
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i ag .4, Mb the shutdown margin shall be 2%. '
g g, g th lets than tour 6rimary coolant pumps in
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r D. cA(a t Eet f-I shutdown and abovd,thorition shall be t e diateiv_initiatedito l increase anQatattlTrtns sh'uTdown margin at 13 E . cjr, j ,
I A A A. ) l
,} K-v <\jw
- c. t less than the het shutdown condition),)with at least one primary 1 f[coolantpump in operation or at issu one shutdown cooling purp in 1
' operation, with a flow rate 22810 gom, the boron concentration ]
d .
snall be greater than the cold shutdown boron concentration for ; i 1
- (normal cooldowns and heatups, is, non emergency conditions.
"'g During non emergenc conditions, at less than the ho shutdown 5 0 condition with no erating primary coolant pumps a a primary system recirculat g flow rate < 2810 gpm but 2 65 gpm, then within one hour ther:
1 1. (a) Esta ish a shutdown margin of 2 3.5% nd (b) As re two of the three charging pu s are electrical.ly d abled.
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- 2. At east every 15 minutes verify th t no charging pumps are o rating, if one or more chargir pumps are determined to ot
, erating in any 15 minute survei lance period, terminate charging pump operation and ins e that the snutdown margin requirements are met and maint ned.
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-( ADO RAAl)
Amendment No. !!, ??, ',7, '", 70. '.'.", 152
-[ At O $( f%:Q A October 26, 1994 3.l.ll CN 3.so Y c f l O' 2-3-f q
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J ATTACHMENT 3 DISCUSSION OF CHANGES z SPECIFICATION 3.1.1, SHUTDOWN MARGIN A.3 CTS 3.10.1c specifies SHUTDOWN MARGIN requirements at "less than the hot shutdown condition" (below 525'F). In the proposed ITS this corresponds to i
MODE 3 <525'F, MODE 4, and MODE 5. The requirements for the refueling condition (MODE 6) are addressed in proposed ITS 3.9.1. This is an administrative change to reflect the NUREG-1432 defined MODES. This change is consistent with hpol the intent of NUREG-1432.
A.4 CTS 3.10.1c includes the statement "...with at least one primary coolant pump 'n operation or at least one shutdown cooling pump in operation, with a flow rate 2: 2810 gpm, the boron concentration shall be greater than the cold shutdown boronunN b p*g concentration." In the proposed ITS for operation with Tave < 525*F,16w SHUTDOWN MARGIN (SDM) rcyuusus will be@ regardlessgf the N [
primary system flow ratc6 _The6DM requement or 347eno wsl exstnfifoughout the temperature range as a cooldown occurs. /The lis definits of sum also allows kredit to taken for the most te tive rod, which was ass >m to be fully withdra , if all control rods c be verified to be inserted independent means. ,
This w uld allow from 1 to 1. % Ap credit for inserted C trol rod worth to be .
adde , depending on the ass ed reactivity from the most eactive rod which is fun" ion of core burnup.
I f erefore, adding this va e to the 2% Ap SDM requir at power, will appr imate 3.5%ap required S . Boron will be added as auired during the c down to account for the temper re defect. JOverall, this is considered to be an administrative
! change since the " cold shutdown boron concentration" requirement is replaced by the -
i requirement to haveg(SDM ff 3.?)RGbthroughout the temperature range. This ;
change could be more or less stri6tive depending on a particular primary coolant temperature evaluated, however overall the requirement is considered an i administrative " substitution" of one requirement for another while still preserving the i _SDM requirements.
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l ATTACHMENT 3.
l DISCUSSION OF CHANGES SPECIFICATION 3.1.1, SHUTDOWN MARGIN i
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>A. 5 CTS 3.10.1b states in part that "...boration shall be immediately initiated to increase !
j and maintain the shutdown margin at...." ' In the proposed ITS this statement becomes l Action A and the term "immediately" is changed to 15 minutes. In the proposed =
NUREG-1432, the time frame of 15 minutes is used in lieu of "immediately" to specify. >
a specific time in which an action must be started. The terminology conveys the same .
{. meaning in the CTS in that quick action must be taken. In NUREG-1432, a j Completion Time of "immediately" is defined in Section 1.3 as " pursue continuously i in a controlled manner without delay." Therefore, while a Completion Time of l "15 minutes" is used in the proposed ITS as compared to the CTS "Immediately" the '
- effective meaning is the same. Therefore, this is considered to be an Administrative Change. This change is consistent with NUREG-1432.
i RAI ;
. A.6 CTS 3.10.la, CTS 3.1.10.lb and CTS 3.1.10c contain the require.nents for %l-o f l SHUTDOWN MARGIN. The amount of required SHUTDOWN MARGIN is l l dependent on the plant operating conditions (e.g., above or bciow hot shutdown) and !
g C,et,p the number of primary coolant pumps in operation. To establish consistency with the l ormat and s_tyle of the ITS, the values of the required SHUTDOWN MARGIN have - \
l m the plant
. been moved t36prveiflance requiremodts (SR 3.1.1.Vand SR 3/1.1.2) j /g e4 specific operating conditions and pump configurations (have Keen Dlaced in
( LA.I f (survomance recurrement Notes 3 A new LCO sgement has been added which statesS f that the SHUTDOWN MAllGIN must be witlE unitsfand an Applicability of {
MODES 3,4, and 5 stipulated. These changes do not alter the actual CTS requirement
- for SHUTDOWN MARGIN, nor do they impose any additional requirements. These
! changes merely present the same information in a different format necessary to convert
. to the ITS. As such, these changes are considered administrative in nature.
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I ATTACHMENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.1.1, SHUTDOWN MARGIN MORE RESTRICTIVE CHANGES (M)
M.1 CTS 3.10.la specifies "With four primary coolant pumps in operation at hot shutdown and above, the shutdown margin shall be 2%." However there is no action specified in the CTS if the shutdown margin is found to be less than 2% and so the plant would have to enter LCO 3.0.3. In the proposed ITS, if the SHUTDOWN MARGIN is found to be below the limit, boration must be initiated within 15 minutes. This is similar to the restoration action specified in CTS 3.10.lb which specifies if shutdown margin is below the required amount that "boration shall be immediately initiated to increase and maintain the shutdown margin." Since in the CTS, LCO 3.0.3 would be have to be entered if the SHUTDOWN MARGIN was
- found to be below the 2% limit, the 15 minutes to initiate boration is considered to be a more restrictive change. Initiating boration to restore the required amount of SHUTDOWN MARGIN is the appropriate action to take in this situation to return the plant to a safe condition. Furthermore, CTS 3.10.lc does not specify actions to take if flow is 2 2810 and the shutdown margin requirements (boron concentration greater than the cold shutdown boron concentration) have not been met. Therefore. if the SHUTDOWN MARGIN was not met, and the plant was above the CTS Cold Shutdown (210*F) then the plant would have to be shutdown in accordance with LCO l 3.0.3. In the proposed ITS, ACTION A requires that if the SHUTDOWN MARGIN l (SDM) requirement is not within limit, then boration must be initiated within 15 j minutes to restore SDM to within limit. Therefore, since the proposed ITS requires I that action be taken with 15 minutes, it is considered to be a more restrictive action. l This change is consistent with NUREG-1432.
W M.2 The Palisades Nuclear Plant CTS does not contain an explicit surveillance requirement 3,1-0 / j for SHUTDOWN MARGIN even though there was a requirement that the limits _ be met as specified in 3.10.1. Proposed ITS 3.1.1 adds SR 3.1.1.1(ana YK 3.1/l.2)to y verify SHUTDOWN MARGIN "every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />." Since the requirement to verify SHUTDOWN MARGIN was not explicitly required in the CTS, the addition of the proposed Frequency is considered a "more restrictive" change. This change is consistent with NUREG-1432.
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' ATTACHMENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.1.1, SHUTDOWN MARGIN .
M.3 CTS 3.10.7 includes an exception which allows a deviation fr'om the requirement for
- . shutdown margin during performance of CRDM exerciset Proposed ITS 3.1.1 does not contain this same exception since violation of the LCO is not expected during the performance of the control rod drive exercise surveillance (SR 3.1.4.4). During the'.-
performance of SR 3.1.4.4, control rods will be exercised between 6 inches and 8 inches. The change in reactivity as a result of this' movement is small due to the 1 relative worth of the control rods which is largely determined by their position in the' core at the time this SR is performed. This small change in reactivity is not enough to cause a violation of the Shutdown Margin requirements of ITS 3.1.1. Thus, reliance on the exception contained in CTS 3.10.7 is not needed. This change is
- consistent with NUREG-1432.
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1 i RESTRICTIVE CHANGES - REMOVAL OF DETAILS TO LICENSEE i CONTROLLED DOCUMENTS (LA)
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j (There were a6 " Removal of Details" channes in thM specifientME] [\ %\' D i
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! CTS 3.10.1 contains the requirements for Shutdown Margin including specific values i based on plant conditions and configuration. This proposed change relocates the values for Shutdown Margin to the COLR in order to provide core design and operational ficxibility that can be used for improved fuel management and to solve plant specific l
issues. Placing the Shutdown Margin values in the COLR allows the core design to be l
finalized after shutdown when the actual end of cycle burnup is known. This would save redesign efforts if the actual burnup differs from the projected value. Current reload design efforts and the resolution of plant specific issues are restricted by the guidelines to not change the Shutdown Margin since it'would result in a License Amendment Request.
Although the actual value of Shutdown Margin is not derived through calculations, it is assumed to be an initial input in the plant safety analyses. As such, a change in Shutdown Margin must be evaluated for its impact on the safety analyses to determine if the revised value results in an unreviewed safety question. Placing the Shutdown Margin limits in the COLR does not result in a significant impact on plant safety since changes to 3 the safety analyses (including a change in Shutdown Margin limits) are done in l
accordance with NRC approved methodologies.
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i ATTACHMENT 3" i DISCUSSION OF CHANGES i SPECIFICATION 3.1.1, SHUTDOWN MARGIN i
LESS RESTRICTIVE W e. usen.no CHANGES t.cs (L) ResilvM , cQw ui +n it cPecSCMW1 K intained ;2: 3.75% whenev r' L.1 CTS 3.10.lb req es SHUIDUWN MARUIN to be
- less than four P ry Coolant Pumps (PCPs) are in rations and the plant is i ot g j shutdown or abo e. The basis for this requirement is o ensure an adequate amo t i of SHUTDOW MARGIN is available to prevent a turn to power following ain
! Steam Line B (MSLB). Inclusion of this requ' ment in the CTS was appr ved /fAl i
in Amendmen 31 to the Provisional Operating Lic for the Palisades Plant g_
(November 1 1977) which authorized power leve up to 2530 Mwt. In sup rt of
. Amendment 1, an analysis of the MSLB with t o operating PCPs was pe rmed to .
address ope tion of the plant with less than fo operating PCPs since thi i configurati was permitted by the technical s ifications during plant tups and cooldowns and for a restricted period of tim at reduced power level. s part of the i conversio to the ISTS and to establish a si le value for SHUTDOW MARGIN in Mode 3 ith less than four PCPs in operati n, a re-evaluation of the SLB with two l operatin PCPs was performed assuming minimum SHUTDOWN ARGIN of l 3.5 % . e evaluation shows that this ev t does not present as gre a challenge to l DNB fuel centerline melt as the ste line break analysis of re rd. As such, ITS 3. 1, "SDM" is proposed with a inimum SHUTDOWN GIN limit of j 3.5%. Relaxing the requirement of S 3.10.1b to maintain a inimum _
SHU OWN MARGIN of 3.75% never less than four PC are operating in hot -)
stand' y (ITS Mode 3) is acceptable ince the consequences of MSLB with less j i thanfour PCPs operating with a S UTDOWN MARGIN of .5% are bounded by the <
MSIB analysis of record.
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[Al i s.\4 ATTACHMENT 4 j NO SIGNIFICANT HAZARDS CONSIDERATION SPECIFICATION 3.1.1, SHUTDOWN MARGIN 4
I
- LESS RESTRICTIVE CHANGES L.1
! Tht.rr. unt.no Ms S:siedw' CQ a.oooc.jdd wrk 40 54ubt. e/\-
CT5 3.iG.i'u mouw S !UTOO"'?! .". "C!?! :: b .x:..: : c.,: ;. 3.75 % . m. ... :m
"[ l l
han four Primary Coolant Pumps (PCPs) are in operations and the plant is in hot shutdown 4
j 3r soove. The basis for this requirement is to ensure an adequate amount of SHUTDOW j WARGIN is available to prevent a return to power following a Main Steam Line Brea !
'MSLB). Inclusion of this requirement in the CTS was approved in Amendment 3 o the l
}
1 Provisional Operating License for the Palisades Plant (November 1,1977) whic uthorized ~
>ower levels up to 2530 Mwt. In support of Amendment 31, an analysis of MSLB with i l wo operating PCPs was performed to address operation of the plant with s than four l j 1perating PCPs since this configuration was permitted by the technical cifications during l
! alant heatups and cooldowns, and for a restricted period of time at r uced power level. As
) art of the conversion to the ISTS and to establish a single value r SHUTDOWN MARGIN I
i n MODE 3 with less than four PCPs in operation, a re-evalu n of the MSLB with two i >perating PCPs was performed assuming a minimum SH OWN MARGIN of 3.5%. The l 1 : valuation shows that this event does not present as great challenge to DNB and fuel
{ :enterline melt as the steamline break analysis of reco . As such, ITS 3.1.1,
' SHUTDOWN MARGIN (SDM)" is proposed wi minimum SHUTDOWN MARGIN l
4 imit of 3.5%. Relaxing the requirement of CT .10.lb to maintain a minimum SHUTDOWN MARGIN of 3.75% whenever ss than four PCPs are operating in hot l standby (ITS MODE 3) is acceptable sinc e consequences of an MSLB with less than four j PCPs operating with a SHUTDOWN GIN of 3.5% are bounded by the MSLB analysis j af record.
- l. Does the change invol a significant increase in the probability or consequence
! of an accident prev usly evaluated?
i l Analyzed even are assumed to be initiated by the failure of plant structures, systems
] or compone . The proposed change relaxes the required SHUTDOWN MARGIN i from 3.75 to 3.5% when less than four PCPs are in operation in MODE 3.
j SHUTD WN MARGIN is neither'a accident initiator, nor accident precursor and i there re can not affect the probability of an accident. Therefore, the proposed cha e does not result in a significant increase in the probability of an accident
{ previously evaluated.
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ATTACHMENT 4 NO SIGNIFICANT HAZARDS CONSIDERATION SPECIFICATION 3.1.1, SHUTDOWN MARGIN E1. (continued )
The consequences of a previously analyzed event are dependent on the initial conditions assumed for the analysis, and the availability and successful func ' ning of the equipment assumed to operate in response to the analyzed e~ent, an . e setpoints at which these actions are initiated. The proposed change relaxes th equired SHUTDOWN MARGIN from 3.75% to 3.5% when less than fo cps are in operation in MODE 3. The minimum required SHUTDOWN ARGIN is an initial assumption used in the MSLB accident which ensures specif' acceptable fuel design limits are not exceeded. A minimum SHUTDOWN M IN value of 3.75%
prevents a return to power in the event of the worst st line break assuming less than four operating PCPs. The maximum return to wer with a 3.5% SHUTDOWN MARGIN is approximately 150 MWt. Although reduction in available SHUTDOWN MARGIN from 3.75% to 3.5% sults in a higher return to power following a MSLB, the consequences of a M B with less than four PCPs operating is bounded by the analysis of record for a SLB. As such, the acceptable fuel design limits and radiological consequences res ing from a change in SHUTDOWN MARGIN are with the limits derived om Standard Review Plan section 15.1.5 appendix A, and 10 CFR 100. Th fore, the proposed change does not involve a f significant increase in the conse nces of an accident previously evaluated.
J
- 2. Does the change create th sibility of a new or different kind of accident from
- any accident previousi valuated?
The proposed ch e does not involve a physical alteration of the plant. No new equipment is mg introduced, and no installed equipment is being operated in a new or differe manner. The proposed change relaxes the required SHUTDOWN MAR from 3.75% to 3.5% when less than four PCPs are in operation in
( MO 3. Therefore, the change does not create the possibility of a new or different L k
- . _' d of accident from any accident previously evaluated.
,s RRi 5 l cl Palisades Nuclear Plant Page 2 of 3 01/20/98 3-m
i ATTACHMENT 4 NO SIGNIFICANT HAZARDS CONSIDERATION SPECIFICATION 3.1.I, SHUTDOWN MARGIN E.3 Does this change mvvive a a;sadicant reduction in a margin of sapf The margin of safety is determined by the design and qua ' tion of the plant
, equipment, the operation of the plant within analy its, and the point at which' protective or mitigative actions are initiated. e proposed change relaxes the
- required SHUTDOWN MARGIN from % to 3.5% when less than four PCPs a e in operation in MODE 3. The sed change does not effect established safety
- limits, operating restrictio r design assumptions. The margin of safety for an ,
i MSLB is established e event described in the FSAR which considers the most limiting case init' from hot full power. This case bounds the consequences (radiologica fuel cladding failure) from other initial operating states including
- operati ith less than four PCPs and an initial SHUTDOWN MARGIN of 3.5 %. t
, the margin of safety previously established for the MSLB accident of record as remained unchanged. Therefore, this change does not involve agant ,
u_ reduction in a marginALSafety._- --
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4 SOMET) > SdO'F finalog) /
-h 4.41-l 3.1 REACTIVITY CONTROL SYSTEMS Fg 3.1.1 SHUTOOWN MARGIN (SON / M > 20f*F (A % og LCO 3.1.1 SOM shall bedd5#4kl Lamin irnd54 @r0gadd 14 fht,,
Yg W7f APPLICA81LITY: MODES 3, Q 4; 'uJ 5 i ACTIONS
! CONDITION REQUIRED ACTION COMPLETION TIME
)
A. SOM not within limit. A.1 Initiate boration to 15 minutes
' restore SOM to within limit.
I i s l i
71TF 'I SURVEILLANCE REQUIREMENTS l
\ , SURVEILLANCE FREQUENCY 3.1.1.1 Verify SOM ( Wi+H 4 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />s-Lith $
4 /
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- T Scd h
i CEOG STS 3.1-1 Rev 1, 04/07/95 s.
7 3-o
gR1 ,
1 SECTION 3.1 3Y .
INSERT 1
. NOTE- -.- -- .
Only required to be met in ODE 3 when T,,, is a 525'F and fo primary coolant pumps are operating.
i i
ERT2 J
......... ... . -.- .-.- -NOTE..- - - - . - - - - . -
. Only required to be met in MOD 3 when T,,,is 2 52 with less than four primary coolant pumps operati , in Mode 3 when T,,, is 525'F, and in MODES 4 and 5.
i SR 3.1.1.2 Verify DM is 2 3.5% Ap. 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 1
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_ _ - - . . - .- . ._. . . - - - .~ . - - .- - - --
i S Q >zw wnla6C i 3. u
-8ASE5 -
APPLICA8LE 50M. An idl/ RCP cannot, therefo/e, produce a returr/to '
3 .ol ,
SAFETY ANALYSES power from the hot standby condifion. /
(continued) .
j SOM satisfies Criterion 2 of kh/NRC Pol)(y Stattdent hO c.F K S o.W 4.WI a LCO The MSLB (Ref. 2) and'the boron dilution (Ref. 3)
, are the most limiting analyses that establish th DM l (of jrhe kCO) For MSLB accidents, if the LCO is violated l there is a potential to exceed the ONBR limit and to exceed 1 10 CFR 100, " Reactor Site Criteria," limits (Ref. 4). For the boron dilution accident, if the LC0 is violated, then i
the minimum required time assumed for operator action to terminate dilution.may no longer be applicable.
1 G+<.i eed)
SOM is a cor pnysics design condition that can b nsured I
through through ositioning (regulating and shutdown e soluble boron concentration.
and J
h <
i APPLICA81LITY In MODES 3 (i!!b the SOM requirements are applicable to --
i provide sufficient negative reactivity to meet the i assumptions of the safety analyses discussed above. In j l MODESIand2,SOMisensuredbycomplyinnwithLCO3.1.M
{ "Shutdowh contral Element / Assembly KEA) nserA. ion Limiti.'l (
p' # ,U ! nb ~
and LC0 3.1.@MJ1f the sortion limits ' LC0 3.1.6 or
!' LCO 3. 7 are not bein complied with, 5 is not WF- 1%j h autom ically violate . The SOM reust b calculated by
.perf ing a reactiv' y balance calcul ion (consider ng the
,' jlis d reactivity e lects in Bases Se ion SR 3.1.1. ). '
IN I]n MO 5 SOM is add essed by LCO 3.1. ~ "5 HUT 00WN
'( )-T , 5 200' .7 in M002 6, the shutdown reactivity
.' requirements are given in LCO 3.9.1, ' Boron Concentration."
ACTI0lts L.1 i If the SOM requirements are not met, boration must be initiated promptly. A Completion Time of 15 minutes is
, adequate for an operator to correctly align and start the
- required systems and components. It is assumed that i boration will be continued until the SOM requirements are met.
(continued) i CEOG STS 8 3.1-4 Rev 1, 04/07/95 4
3-q l
B 3.1.1
_ 8ASES- -
1_ '
] ACTIONS L1 (continued) i In the determination of the required combination of boration
, flow rate and boron concentration, there is no unique requirement that must be satisfied. Since it is imperative
]
- ^ ",, ,J to raise the
~
posilbTeT'f boron1constntrafiom he~ boron concentration shouldof bethe $5 a highly as soon 'h as b 4 concentrated solution, such as that normally found in the
@h_ee,b.4,bboric acid storage tan (R the' t, orated Avater sto(ace tank
{ The operator should borate oi w 6ne best source available
- for the plant conditions.
In determining the boration flow rate, ti:e time ore life {@
i must be considered. For insLince, the most difficult time 1 in core life to increase the%15 boron con. 9tration is at ,
2 the beginning of cycle, when the borna concentration may i approach or exceed 2000 ppm. Assuming that a value of 8
l j @ {1% atts must be recovered and a boration flow rate of55 j of thetBS b Toroh73rth'y100ppminapproximate1)@35 If a l@
ofl19 acm/ Dom is assumed, this e minutes.
1 g c-'i 4 mbination of I 2
_____/mlparameters will increase the SDM by 1% These boration l l parameters of C5} gpm and tidppm represent typical values I and are provided for the purpose of offering a specific example.
l i SURVEILLANCE SR 3.1.1.1
, REQUIREMENTS i SDM is verified by pe*4eem 4g a reactivity balance j calculation, considering the listed reactivity effects:
I -
a.@pSboronconcentration; l@
- b. M positions; {h j c.b@CSaveragetemperature; {
t 3 d. Fuel burnup based on gross thermal energy generation; i
- e. Xenon concentration; l l f. / Samarium foncentratiofh land (h
@p Isothermal [emperature p#oefficient(!TC).{h
}~ (continued) j CE0G STS B 3.1-5 Rev 1, 04/07/95 i
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SECTION 3.1 INSERT 1 Samarium is not considered in the reactivity analysis since the analysis assumes that the negative reactivity due to samarium is offset by the positive reactivity of plutonium build in.
RRl 3.l-0 l INSFRT 2 To maintain co istency with the assumptio s used in the MSLB analysis, vo values of SDM are specified i the surveillance requireme s. J SR 3.1.1.1 rec uires SDM to be 2 2% Ap. filiiTSDM value ensuTiirthe nsequences of an MSLB, tyi we I as the other events scribed in the Applicable Safety nalysis, will be accepta le whenever the plant is i MODE 3 with T,,.,2 525*F and f ur Primary Coolant Pumps (PCPs are in operation. Theref e, SR 3.1.1.1 is modified by a No which only requires this SRt e met in MODE 3 when mm 2 525 _F and four PCPs are i _peration. SR 3.1.1.2 reauires
'S D. to be 2 3.5% 60 IThis SDM value ensures the consje ueneds of an MSLBfli!Iwellgas th )
^
,_filTiiir eVeni L clescriD4DN13hcilbic 6fdAnaiy'si[fviITdTEc t3bl6iis~aTeiu'}fi'6fa cooldown of the PC5TihTcifiddi poTti7eTeatrsitfiriN presence o[f a nega
~
temnerature coefficieg. As such, the requirements of SR[3.1.1.2{must be met whenevet me plant is in MODE.3rwim 1,,, a 25 F with less than loufFCFs operating, MODE T tth 1,,, < 525 ~F, and N DES 4 and 5. T erefore, SR 3.1.1.2 is nyfdified by a Note which on) requires this SR lo b met in MODE _3 ith T_ > 525'F with le/s than four PCPs operating, ODE 3 with T,,, j
~ ~
(. 5'FgMODES 4 and 5.
B 3.1 6 3-s
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ATTACHMENT 6 I- JUSTIFICATION FOR DEVIATIONS SPECIFICATION 3.1.1, SHUTDOWN MARGIN. T,,, > 200'F i
j Change Discussion i
h 4
l
- 7. ISTS Change Traveler TSTF-136 combines ISTS 3.1.1 and ISTS 3.1.2 into a single specification in order to eliminate unnecessary and confusing duplication, and renumbers the remaining specifications in Section 3.1. The impetus for this change
/(M 3
i was the approval of TSTF-9 which allowed the nlues for shutdown margin to be la M*
1 moved to the COLR. As a result of TSTF-9, the LCO, Actions, and Surveillance I '
- Requirements of ISTS 3.1.1 and ISTS 3.1.2 were the same. APJ.4. i Palisades ptas has net relocated the shutdown margin values to the COLIC itaas acopted g
l 1 (E5nssidation SISTS 3.1.1 and ISTS 3.1.2 into a single specification. Proposed ITS X 371.1 address the plant conditions encompassed in MODES 3,4, and ggs a result of this consolidatiorgfa new Surveittance Requi ment nas been added (115 SK . l .1.2),
{
! Eand Note ~~s been included to SR 3.1.1.1 modify the performance of the 1 surveilla 'es. The format of the Surveill ce Requirements are consistent ith the I l 4
tstylea format presented in the Writer Guide (NUMARC 93 0D d 3
! 8. The Palisades plant was designed prior to issuance of the General Design Criteria
- (GDC) in 10 CFR 50. Therefore, reference to the GDCs is omitted and appropriately
- replaced by reference to " Palisades Nuclear Plant design criteria ." The Palisades Nuclear Plant design was compared to the GDCs as they appeared in 10 CFR 50 Appendix A on July 7,1971. It was this updated discussion, including the identified exemptions, which fctmed the original plant Licensing Basis for future compliance l with the GDCs.
- I-d
- 9. TSTF-9 permits relocation of the shutdown margin values specified in ISTS 3.1.1 and -
this option End has ulamtamed tne requirea snumown va10ps)in the ITS.
% e o-Pero /Hafr.
, hun %% See %r key Is provodJ in DOL. LA.) Ger .trS LI l .
- 10. Samarium is not considered inihe Palisades Nuclear Plant reactivity bahnce due to the ,
- fact the that Palisades Nuclear Plant fuel vendor does not account for Samarium in fuel !
design calculations. The vendor assumes that the negative reactivity defect due to l Samarium is offset by the positive reactivity of Plutonium build in. Plutonium build in ;
and Samarium are equally competing reactivity effects that are accounted for in fuel l design calculations. Therefore, including Samarium into the SDM calculation would j not be correct for the Palisades Nuclear Plant.
4 Palisades Nuclear Plant Page 2 of 3 01/20/98 f
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Reporting Requirements
, 5.6 5.6 Reporting Requirements 5.6.4 Monthlv Ooeratino Report Routine reports of operating statistics and shutdown experience shall be submitted on a monthly basis to the NRC no later than the fifteenth of each month following the calendar month covered by the report.
- 5.6.5 CORE OPERATING LIMITS REPORT (COLR) 4
! a. Core operating limits shall be established prior to each p reload cycle, or prior to any remaining portion of a reload aj,ol
- cycle, and shall be documented in the COLR for the following:
's.I.1 5k Aduo n inarbs y 3.126 Regulating Rnd Group Position Limits 4
3.2.1 3.2.2 Linear Heat Rate Limits Radial Peaking Factor Limits hx 3.2.4 ASI Limits
- b. The analytical methods used to determine the core operating limits shall be those approved by the NRC, specifically those described in the latest approved revision of the following documents:
i 1. XN-75-27(A),"ExxonNuclearNeutronicsDesignMethods N-ol for Pressurized Water Reactors," and Supplements 1(A),
4 2(A), 3(P)(A), 4(P)(A), and 5(P)(A); Exxon Nuclear j Company. (LCOs3 3.1.6, 3.2.1, 3.2.2, & 3.2.4) g 31.1
- 2. ANF-84-73(P)(A), " Advanced Nuclear Fuels Methodology for Pressurized Water Reactors: Analysis of Chapter 15 Events," and Appendix B(P)(A) and Supplements 1(P)(A), 2(P)(A); Advanced Nuclear Fuels Corporation.
(LCOs33.1.6, 3.2.1, 3.2.2, & 3.2.4) T sq
- 3. XN-NF-82-21'(P)(A), " App ' cation of Exxon Nuclear Company PWR Thermal Ma' 'n Methodology to Mixed Core Configurations," Exxon helear Company.
(LCOs 3.2.1, 3.2.2, & 3.2.4)
- 4. ANF-84-093(P)(A), "Steamline Break Methodology for PWRs," and Supplement 1(P)(A); Advanced Nuclear Fuels Corporation. (LCOs33.1.6, 3.2.1, 3.2.2, & 3.2.4) X M1
- 5. XN-75-32(P)(A), " Computational Procedure for Evaluating Fueland 2(P)(A), 3(P)(A), Rod Bowing,")and 4(P)(A Supplements
- Exxon Nuclear Company. 1(?)(A),
(LCOs3.1.6,3.2.1,3.2.2,&3.2.4)
Palisades Nuclear Plant 5.0-25 Amendment No. 01/20/98 3-u
.. . . . _ .. - - . - - . - . . = _ - - . -
'b i enhM i
3 6.0 40MtWf1TRATIVE CCNTROL1 i
j r, (, ,5 I.44 6 cera anaratina timits Raeert (COLR) g Core operating lietts shall be estabitshed prior to each reload j j a.
j cycle, or prior to any remaining portion of a reload cycle. and shall be documented in the COLR for the following: (f h y l
3.I t Q ^
\ SkAdaan (Quo r) j 11.9 A5! Listts. 1.2 _ _ ss.4'*
i 1.b le a Regulating r "namt - Limits j 3.1.1 y Linear Heat Rat. 3 Limits 3 .1, 't. g Radial Peaking Ft .it Lietts I l b. The analytical methods used to determine the core operating limits shall be those approved by the NRC. specifically those described in
' the latest approved revision of the following documents:
l 1. IN 75 27(A). ' Exxon Nuclear Neutrontes Design Methods for j Pressurtzed Water Reactors.' and Supplements 1(A). 2(A).
L j 2. ANF 84 73(P)(A). ' Advanced Nuclear Fuels Methodology for I Pressurized Water Reactors: Analysis of Chapter 15 Events.' I and Appendix $(P)(A) and Supplements 1(P)(A). 2 Advanced N g r Fuels Corporation. (LC03 @ (P) A - g,g
, ,g
- b % { s.t.l 3.t.M 1. s. O
- 3. IN NF 42 !!(P)(A). ' App 1tcation of Exxon Nuclear Cocoany PWR Thermal Margin Methodology to Mix C fig ns.' , j
' Exxon Nuclear Company. (LCOs . &
la 1. i 1.1.1. i 4, ANF 84 093(P)(A) 'Steneline treak Methodology for PWRs.' and l
1 P Ad Nc els Corporation.
)
- 5. IN 75 32(P){A). ' Computational Procedure for Evaluating Fuel Rod Bowine, and Supplements 1(P)(A). 2 p A) 3 P A). n 4P A' taxon Nuclear Company. (LCOs . . . .
) s. w 1.1.b 2.1.6
- 1. t. o
- 6. L as defined by:
a) IN NF 42 20(A)
- Exxon Nuclear Cogany Evaluation Model EXEWPWR ECCS Model Updates.' and supplements 1(P)(A).
- 2(P)(A). 3(P)(A), and 4(P)(A); Exxon Nuclear Company.
' b) IN NF St 07(P)(A). ' Exxon Nuclear Company ECCS Cladding l
Swelling and Rupture Modcl.' Exxon Nuclear Company.
I c) IN NF 81 54(A). 'R00EIt Fuel Red Thermal Mechanical Response Evaluation Model.' and Supplements 1(P)(A).
i 2(P)(A). 3(P)(A), and 4(P)(A); Exxon Nuclear Company, i
6 20 i
' Amendment No, M4. 174 October 31, 1994 4.
3-v
1 ATTACHMENT 3 DISCUSSION OF CHANGES j CHAPTER 5.0, ADMINISTRATIVE CONTROLS j A.5 CTS 6.4.1 requires that written procedures shall be established, implemented, and maintained for the activities listed. In this list, the CTS contains item b., " Refueling 1 operations, and item c., " Surveillance and test activities of safety-related activities."
- These items are included in the procedures recommended in Appendix "A" of. ,
j Regulatory Guide 1.33, Revision 2, February 1978 which is referenced in CTS 6.4.la '
and included in the proposed ITS 5.4.la. Therefore, since these procedures are j already required by the reference to Regulatory Guide 1.33, Revision 2,
- February 1978, they are not included in the proposed ITS. This change is an
- administrative change since no requirements have changed. This change maintains
! consistency with NUREG-1432. .
}
l A.6 CTS 6.4.1 requires that written procedures shall be established, implemented, and ,
1 maintained for the activities listed. In this list, the CTS contains item f., " Site '
Security Plan implementation" and item g ," Site Emergency Plan implementation.."
These items were recommended to be removed from the Technical Specifications in ,
- NRC Generic Letter 93 07 since they are duplicative of regulations contained in the !
j Code of Federal Regulations part 50 and 73. This change is considered to be an .
- administrative change since these requirements must still be met as required by the
- Code of Federal Regulations. This change maintains consistency with NUREG-1432.
A.7 CTS 6.5.7 is entitled " Inservice Inspection and Testing Program." In the proposed
- ITS 3.5.7, the title is changed to the " Inservice Testing Program." This change is ( !
considered to be an administrative change since the requirements of the program are \ .
unchanged. This change maintains consistency with NUREG-1432. p0!
,[ A.8 NM5b.1 It ts, among r erenced LCus, -3711)T.#That item i[uNe ssrynhl has been deleted. Neither S 3.10.1, nor its ITS replacement reference e COLR. l CTS 6.6.5 a. lists the cor operating limits that are established and doc ented in the 1 p) COLR prior to each cor reload. Specifically, these limits are: ASI L' its CTS 3.1.1), Regulatin Group Insertions Limits (CTS 3.10.5), Line Heat Rate l
l
[ljy ph{ (Limits (CTS 3.23.1), pd Radial P'eaking Factor Limits (CTS 3 . CTS 6.6.5 b. I
/ list the documents apfoved by the NRC that describe the analytica r. ethods used lto '
l
{ determine the core o erating limits. As part of this listing, cross eferences are made l ,
- to the LCOs pertai ng te the affected limit (e.g., ASI Limits, gulating Group l
, insertion Limits, c...). In error, CTS 6.6.5 b.1. lists CTS 3 0.1 (Shutdown l
- Margin Require nts) as an LCO related to a document that escribes analytical I methods used to determine the core operating limits. Since hutdown Margin is not a l l i
cycle dependen limit (the limit is contained in the technica specifications and not in l
. the COLR), re erencing CTS 3.10.1 in CTS 6.6.5 b.1 is i appropriate and has been 1 l
deleted. This change has been characterized as administr tive in nature since it does l p =y requirtment of the CTS, but simply corrects an administrative oversight. l
^ i Palisades Nuclear Plant - Page 2 of 7 10/10/98 i
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INSERT R A I 3.1 bl .
j.
CTS 6.6.5 a. lists the core operating limits that are established and documented in the COLR
- prior to each core reload. Specifically, these limits are: ASI Limits (GS 3.1.1), Regulating ,
Group Insertions Limits (CTS 3.10.5), Linear Heat Rate Limits (CTS 3.23.1), and Radial Peaking Factor I.imits (CTS 3.23.2). CTS 6.6.5 b. list the documents approved by the NRC that describe the analytical methods used to determine the core operating limits. As part of this
- listing, cross references are made to the LCOs pertaining to the affected limit (e.g., ASI j Limits, Regulating Group Insertion Limits, etc...). In error, CTS 6.6.5 b.1 lists CTS 3.10.1 (Shutdown Margin Requirements) as an LCO related to a document that describes analytical i methods used to determine the core operating limits. However, as part of the conversion to the -
Improved Technical Specifications, the values for Shutdown Margin were relocated from j CTS 3.10.1 to the COLR consistent with NUREG-1432 as modified by TSTF-9. As such, CTS 6.6.5 (proposed ITS 5.6.5) has been revised to include ITS LCO 3.1.1 " Shutdown
! Margin" as a limit that is established and maintained in the COLR. This change has been j characterized as administrative in nature since it does not alter any requirement of the GS. but j simply provides conforming information.
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I Reporting Require:ents j
5.6
-- 5.6 Reporting Requirements
,q 5.6.4 Monthly coeratina Reoorts (continued) l l
l power o,er ged relief valve /or pressurizer safety / valves,]lshall ao submitted on a monthly basis no later than the 15th of each @
month following the calendar month covered by the report.
3 l
(. .( 5.6.5 CORE OPERATING LIMITS REPORT (COLR1 pll si'o
! a. Core operating limits shall be established prior to each reload cycle, or prior to any remaining portion of a reload )(
cco M4T ph u.o 3. i. : rn$* cycle, and shall be documented in the COLR for the
- Pa d* ' 11owing:
Gy ..,+5 i
u
- LL \
M' y The indifidual specifications thpt address cor$ operating limits must be referdnced here. / /
f.
t.co t L t LLA 3 b'S b. The analytical methods used to determine the core operating
- l Y.'f, L limits shall be those previously reviewed and approved by .
l [r u *J the NRC, specifically those described in following tco u.s
- documents
- y m v, , y g j Identify e Topical Repo t(s) by number, itle,date,fand f u scRT / p NRC staff approval doc t, or identify he staff SafAty F
- Evaluati n Report for a lant specific thodologyby/NRC letter d date. / _
! c. The core operating limits shall be determined such that all i applicable limits (e.g., fuel thermal mechanical limits,
, core themal hydraulic limits, Emergency Core Cooling
- analysis limits, and accident analysis limits) of the safety analysis are met.
i
- d. The COLR, including any mid cycle revisions or supplements, 4 shall be provided upon issuance fer each reload cycle to the NRC.
5.6.6 ReactrirCdantSystem(RC PRESSURE @D TEM RATURE LIMITS i REPOR" (ITLin a.. 5 pressure and mperature limits f r heatup, cooldown, low temperature eration, critical , and hydrostatic
! (continued)
CEOG STS 5.0 21 Rev 1, 04/07/95 3-y
1
- SECTION 5.0 4
INSERT
- . pl j 1. XN-75-27(A), " Exxon Nuclear Nt utronics Design Methods for Pressurized 3.PD 4 4
Water Reactors," and Supplements 1(A),2(A),3(PXA),4(PXA), and 5(P)(A);
Exxon Nuclear Company. (LCOs .1.6, 3.2.1, 3.2.2, & 3.2.4) {
i 2. ANF-84-73(P)(A), " Advanced Nuclear Fuels Methodology for Pressurized Water Reactors: Analysis of Chapter 15 Events," and Appendix B(P)(A) and ,
j Supplements 1(P)(A),2(P)(A); Advanced Nuclear Fuels Corporation. y
- (LCOsn 3.1.6, 3.2.1, 3.2.2, & 3.2.4)
- 3.l.1
- 3. XN-NF-82-21(P)(A), " Application of Exxon Nuclear Company PWR Thermal l Margin Methodology to Mixed Core Configurations," Exxcn Nuclec Company. (LCOs 3.2.1, 3.2.2, & 3.2.4) i 4. ANF-84-093(P)(A), "Steamline Break Methodology for PWRs" and Supplement 1(P)(A); Advanced Nuclear Fuels Corporation. I l (LCOsg3.1.6, 3.2.1, 3.2.2, & 3.2.4) s.1 I
- 5. XN-75-32(P)(A), " Computational Procedure for Evaluating Fuel Rod Bowing,"
and Supplements 1(PXA), 2(PXA), 3(P)(A), and 4(P)(A); Exxon Nuclear
, Company. (LCOs 3.1.6, 3.2.1, 3.2.2, & 3.2.4) l
(LCOs 3.1.6, 3.2.1, & 3.2.2) i a) XN-NF-82-20(A), " Exxon Nuclear Company Evaluation Model EXEM/PWR ECCS Model Updates," and Supplements 1(P)(A),2(P)(A),
3(P)(A), and 4(P)(A); Exxon Nuclear Company.
b) XN-NF-82-07(P)(A), " Exxon Nuclear Company ECCS Cladding Swelling and Rupture Model," Exxon Nuclear Company.
l i
I 5.0-21
- 4 3-z
CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO TNE DECEMBER 09, 1998 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.1, REACTIVITY CONTROL SYSTEM NRC REQUEST:
3.1-02 ITS 3.1.3 [STS 3.1.4] Moderator Temperature Coefficient (MTC)
ITS SR 3.1.3.1 and STS SR 3.1.4.1 Note JFD 11 The STS SR 3.1.4.1 includes a note that the SR need not be performed prior to l entry into Mode 2. This note has been excluded in the ITS because the frequency specifies prior to 2% RTP.
Cotenent: The SR frequency does not negate the applicability of SR 3.0.4; that SRs must be met prior to entry into modes of applicability. In any case, including the note avoids misinterpretation. Recommend including the note.
Consumers Enerav Resoonse:
The Note which modifies ISTS SR 3.1.4.1 is intended to avoid a potential SR 3.0.4 conflict. However, the inclusion of this Note in the ISTS is redundant since the Frequency specifies the precise requirement for performing
- the surveillance. NUMARC 93-03 " Writer's Guide for the Restructured Technical Specification" Section 4.1.7 (Chapter 3 Surveillance Requirements Contents) !
item "f" states; "To specify the precise requirements for performance of a Surveillance, such that exceptions to SR 3.0.4 would not be necessary, the Frequency may be specified such that it is not [due] until the specific conditions are met. Alternately, the surveillance may be stated as not required [to be met or performed] until a particular event, condition, or time l l
has been reached." The Frequency of proposed ITS SR 3.1.3.1 is specified as a
" condition" versus a " Mode". Therefore, a corresponding Note in the SR would have to be stated as a " condition" (i.e., Not required to be performed prior
! to 2% RTP) to avoid an SR 3.0.4 conflict between the entry conditions of
- Mode 2 (keff 2 0.99) and 2% RTP. Since a Note containing this information would be redundant to the Frequency, it was not included in the ITS.
~
Affected Submittal Paaes:
None l
l 4 l I
i
CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO THE DECEMBER 04, 1998 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.1, REACTIVITY CONTROL SYSTEM NRC REQUEST:
3.1-03 ITS 3.1.3 [STS 3.1.4] Moderator Temperature Coefficient (MTC)
STS SR 3.1.4.2 JFD 6 The STS SR 4.1.4.2 is not included because it is not in the CTS and "... the most negative limit is also assured of being met by design."
Comnent: Once the initial MTC measurement is met is it always true that End-Of-Cycle (E0C) measurements will be met for all core loadings? Is this a plant unique feature?
Consumers Enerav Response:
Yes, once the value of MTC is verified to be less positive than the technical specification limit at the beginning of core life, the value of MTC will always be less than the technical specification limit at the End-of-Cycle based on current core loading design methodologies. It is believed this feature is not unique to the Palisades plant.
In regards to the change in MTC over core life, ISTS SR 3.1.4.2 requires a verification that MTC is within the lower limit assumed in the safety analysis after reaching 40 EFPD of core burnup, and within 7 EFPD of reaching two-thirds of the expected core burnup. As discussed in JFD 6, the CTS does not contain a requirement to verify MTC is within the lower limit assumed in the safety analysis since this value is assured by core design. That is, the ,
measured velue of MTC can be extrapolated using core modeling techniques to I determine the value that will exist at the end of core life. The predicted I value of MTC is verified to be less negative than the value previously assumed in the safety analysis. Inherent to this process is the assumption that the core continues to behave as designed. This assumption is verified by performing proposed ITS SR 3.1.2.1 which verifies the overall core reactivity .
balance is within plus or minus 1% of the predicted values every 31 EFPD. l Should an anom31y greater than 1% develop between the measured and predicted I core reactivity values, an evaluation of the core design and the effects on !
the safety analyses must be performed.
Affected Submittal Paaes:
None 1
s
CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO THE DECEMBER 04, 1998 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.1, REACTIVITY CONTROL SYSTEM i NRC RE00EST:
3.1-04 ITS 3.1.4 [STS 3.1.5] Control Rod Alignment ITS 3.1.4 Required Action D Bases LC0 section (page B 3.1-26) and Bases ACTIONS (page B 3.1-30)
DOC M.6
! JFD 10 and JFD 17 I The ITS has added a Required Action D that an immovable but trippable control rod
- shall be returned to operable status prior to entering Mode 2.
Consent H: The completion time for the Required Action is prior to entering the
. LC0's applicability, which is illogical; the condition is not needed.
Consumers Enerav Resonant:
ITS Condition "D" addresses the situation when one full-length control rod is immovable but trippable. As described in DOC M.6, the CTS does not contain an explicit LC0 for control rod Operability. Thus, the plant is allowed unrestricted operation when one control rod is inoperable. Since proposed ITS 3.1.4 requires all control rods to be Operable, declaring an immovable but i trippable control rod inoperable without a corresponding Required Action, would require entry into Specification 3.0.3. As such, ITS Condition "D" has been incorporated to preclude an unnecessary plant shutdown due to an immovable control rod. Since unlimited continued operation with an inoperable, but trippable, rod is allowed, LC0 3.0.4 would not prohibit MODE changes while in
- Condition "D." The proposed Completion Time was specified to assure repairs were 3
made prior to the next reactor start-up.
Affected Submittal Paaes:
j Att 1 ITS 3.1.4, page 3.1.4-2 Att 2 ITS 3.1.1, page B 3.1.1-3 Att 2 ITS 3.1.4, page B 3.1.4-1 Att 2 ITS 3.1.4, page B 3.1.4-4 Att 2 ITS 3.1.4, page B 3.1.4-8
. Att 2 ITS 3.1.4, page B 3.1.4-10 Att 2 ITS 3.1.4, page B 3.1.4-11 ,
Att 2 ITS 3.1.4, page B 3.1.5-1 Att 2 ITS 3.1.4, page G 3.1.5-4 )
Att 2 ITS 3.1.4, page B 3.1.6-1 Att 2 ITS 3.1.4, page B 3.1.6-4 Att 5 NUREG, page B 3.1-34 Att 5 NUREG, B 3.1-36 Insert 6
I
Control Rod Alignment 3.1.4 ACTIONS CONDITION REQUIRED ACTION COMPLETION ' TIME - RM fvtL lu%h (
D. One^ control rod D.1 Restore control rod Prior to ininovable, but to OPERABLE status. entering MODE 2 trippable. from MODE 3 E. Required Action and E.1. Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time not met.
E One or more control rods inoperable for reasons other than Condition D.
M Two or more control rods misaligned by
> 8 inches. J E
Both rod position indication channels inoperable for one or ;
- more control rods.
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Palisades Nuclear Plant 3.1.4-2 Amendment No. 01/20/98 i=
6-a
SDM B 3.1.1 BASES APPLICABLE In addition to the limiting MSLB transient, the SDM SAFETY ANALYSES requirement for MODES 3 and 4 must also protect against an (continued) inadvertent boron dilution; (Ref. 3) and an uncontrolled control rod bank withdrawal from subcritical conditions (Ref. 5).
Each of these events is discussed below.
In the boron dilution analysis, the required SDM defines the reactivity difference between an initial subcritical boron concentration and the corresponding critical boron concentration. These values, in conjunction with the configuration of the PCS and the assumed dilution flow rate, directly affect the results of the analysis. This event is most limiting at the beginning of core life when critical boron concentrations are highest.
The withdrawal of a control rod bank from subcritical conditions aJds reactivity to the reactor core, causing both the core power level and heat flux to increase with corresponding increases in reactor coolant temperatures and pressure. The withdrawal of control rod banks also produce ,
a time dependent redistribution of core power. j Depending on the system initial conditions and reactivity insertion rate, the uncontrolled control rod banks i withdrawal transient is terminated by either a high power !
trip or a high pressurizer pressure trip. In all cases. l power level, PCS pressure, linear heat rate, and the DNBR do l not exceed allowable limits.
SDM satisfies Criterion 2 of 10 CFR 50.36(c)(2).
LC0 k.
The MSLB (Ref. 2) and the boron dilution (Ref. 3) accide.ots..- n are the most limiting analyses that establish the 4DM kalue 3 OL tM LCO. For MSLB accidents, if the LC0 is violated ?
(
there is a potential to exceed the DNBR limit and to exceed 10 CFR 100, " Reactor Site Criteria," limits (Ref. 4). For the boron dilution accident, if the LC0 is violated, then the minimum required time assumed for operator action to terminate dilution may no longer be applicable.
SDM is a core physics design condition that can be ensured Euu. lugte s throughxontrol rod positioning (regulating and shutdown h0 rods) and through the soluble boron concentration.
Palisades Nuclear Plant B 3.1.1-3 01/20/98 6-b .
. 1
{ Control Rod Alignment 4- B 3.1.4 h B 3'.1 REACTIVITY CONTROL SYSTEMS
. B 3.1.4 Control Rod Alignment l 4
BASES
, BACKGROUND The OPERABILITY (e.g., 'trippability) of the shutdown and pk.d/ i i regulating rods is an initial assumption in all' safety: k :
- analyses that assume gontrol rod insertion upon reactor )(
j g,f7 trip. Maximum control rod misalignment is an initial
- assumption in the safety analysis that directly affects' '
! core power distributions and assumptions'of available SDM. l 1
l' The Palisades Nuclear Plant. design criteria contain .the' q
3pplicable distributioncriteria designfor these reactivity)and requirements (Ref.1. power i l
Mechanical or electrical failures may 'cause a control rod ..
! to become inoperable or to become misaligned from its -l group. Control rod iaw91'it r misalignment may cause y( d - '
i increased power peaking, due to the asymmetric reactivity-distribution, and a reduction in the. total available ,
control rod worth for reactor shutdown.. Therefore, control I rod alignment and OPERABILITY are related to core operation ;
in design power peaking limits and the core design-
- requirement of a minimum SDM.
I Limits on control rod alignment and OPERABILITY have been
! established, and all control rod positions are monitored and controlled during power operation to ensure that the power distribution and reactivity limits defined by the design power peaking and SDM limits are preserved.
Control rods are moved by their Control Rod Drive I~
j Mechanisms (CRDMs). Each CRDM moves its rod at a fixed rate of approximately 46 inches per minute. Although the (t[
- ability to move a ontrol rod by its drive mechanism is not X i an initial assumpt used in the safety analyses, it-is reqC.ed to suppor PEkABILITY.. As such, the inability to
- move a4 control rod results in that. control rod being y y inoperable.
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Palisades. Nuclear Plant B 3.1.4-1 01/20/98
)
6-c.
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J Control Rod Alignment-B 3.1.4 '
BASES
' APPLICABLE The most limiting static misalignment occurs-when Bank .4 is.
SAFETY ANALYSES fully inserted with one rod fully withdrawn _ (Bank 4 is -
l (continued) 99 inches out of alignment with the rated Power Dependent
[ Insertion Limit (PDIL). This event was bounded by the dropped full-length control rod event (Ref. 4).
,i
! Since the control rod = drop incidentt result in the most rapid approach to SAFDLs caused by a control rod 4 misoperation, the accident analysis anilyzed a sir e i
full-length control rod drop. The most rapid appr ach to j the DNBR SAFDL may be caused by a single full-length control rod drop.
1 The above control rod misoperations may nr may not result i
j in an automatic reactor trip. In the. case of.the full-length rod drop, a prompt decrease in core average q i power and a distortion in radial power are initially
! produced, which, when conservatively coupled, result in a local power and heat flux increase, and a decrease in DNBR i parameters.
The results of the control rod misoperation analysis show l l- that during the most limiting misoperation events, no violations of the SAFDLs, fuel centerline temperature, or l
PCS pressure occur.
Control rod alignment satisfies Criteria 2 and 3 of 10 CFR 50.36(c)(2).
LCO The limits on shutdown, regulating, and part-length rod 9' .
alignments ensure that the assumptions in the safety l analysts will remain valid. The requirements o QJ loy* - N OPERABILITY ensure that upon reactor trip, the control rods will be availabi'e and will be inserted to provide enough negative reactivity to shut down the reactor. The OPERABILITY requirements also ensure that the control rod banks maintain the correct perer di:trihtier : d :: tr:1 k eed alignment'and that each(control rod is capable of beine d 1 moved by its CRDM. The OPERAt31 Lilt requirement for the
_part-length rods is that they are fully withdrawn.: d 3 : 4 ;
cepab'e of be45 - ed by their C"S .
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Palisades Nuclear Plant B 3.1.4 01/20/98 6-d
I Control' Rod Alignment P 3.1.4 .;
- BASES
- ACTIONS (continued) hl
' k*\d Condition D is entered whenever it is discovered that 'a X singleacontrol rod can not be moved by its operator yet tha ,
contro' rod is still capable of being tripped. Although g 1 Qd- 4 3 the aD111ty to move a& control rod is not an initial' 4
4 assumption used in the safety analyses,.it does relate to ,k X
,_tcontrol rod OPERABILITY. The inability to move a+ control '
- roa oy its operator may De ind1Cative of a systemic failure (other than trippability) which could potentially affect -
i other rods. Thus. declaring aw ontrol rod inoperable in 1('
this instance is conservative since it limits the number of .
l_3 control rods which can not be moved by their operators to
- only one. .The Completion Time to restore an inoperable l
control rod to OPERABLE status is stated as prior to i entering MODE 2 from MODE 3. This Completion Time allows unrestricted operation in MODES 1 and 2 while i conservatively preventing a reactor startup with an - g~
! ininovable+ control rod.
L.1 If the Required Action or associated Completion Time of Condition A, Condition B, Condition C, or Condition D is
[
4 not met; one or more control rods are inoperable for reasons other than Condition D; or two or more control rods l
! are misaligned by > 8 inches, or two channels of control rod position indication are inoperable for one or more control rods, the plant is required to be brought to
' MODE 3. By being brought to MODE 3, the plant is brought outside its MODE of applicability. Continued operation.is
- not allowed in the case of more than one control rod misaligned from any other rod in its group by > 8 inches, or two or more rods. inoperable. This is because these
- cases may be indicative of a loss of SDM and power i
re-distribution, and a loss of safety function, respectively.
- Also, if no rod position indication exists for one or more
- control rods, continued operation is not allowed because the safety analysis assumptions of rod position cannot be
< ensured. ,
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- Palisades Nuclear P1 ant B 3.1.4-8 01/20/98~
6-e ,
Control Rod Alignment B 3.1.4 BASES SURVEILLANCE SR 3.1.4.3 REQUIREMENTS (continued) Demonstrating the rod position deviation alarm is OPERABLE verifies the alarm is functional. The 92 day Frequency takes into account other information continuously available to the operator in the control room. so that during control rod movement, deviations can be detected.
SR 3.1.4.4 Verifying each full-length control rod is trippable would reouire that eachecontrol rod be tripped. In MODES 1 Y pag. g
\ and 2. triopina each<ontrol rod would result in radial or y axial power tilts, or oscillations. Therefore, individual full-length control rods are exercised every 92 days to provide increased confidence that all+ control rods continue y 4
to be trippable, even if they are not regularly tripped. A i movement of 6 inches is adequate to demonstrate motion without exceeding the alignment limit when only one control
' rod is being moved. The 92 day Frequency takes into consideration other information available to the operator -
in the control room and other surveillances being performed more frequently, which add to the determination of OPERABILITY of the control rods. At any time, if a control rod (s) is inoperable, a determination of the trippability
- of the control rod (s) must be made, and appropriate action taken.
l Palisades Nuclear Plant. B 3.1.4-10 01/20/98 6-f
Control' Rod ~ Alignment ' ]
B.3.1.4 BASES SURVEILLANCE-REQUIREMENTS SR 3.1.4.5 CM Q
(continued) Perfcrmance of.a CHANNEL CALIBRATION of eacherod position X i indication channel ensures the channel.is OPERABLE and I
. capable of indicating control rod position over the entire length of the control rod's travel with the exception ofJ the secondary. rod position indicating channel dead band-near the bottom of travel. This dead band exists because the control rod drive mechanism housing seismic support prevents operation of the reed switches. Since this Surveillance must be performed when the reactor is shut down, an 18 month Frequency to be coincident with refueling outage was selected. Operating experience has shown that' these components usually pass this Surveillance when performed at a Frequency of once every 18 months.
Furthermore, the Frequency takes into account other surveillances being performed at shorter Frequencies,'which determine the OPERABILITY of the control rod position indicating systems.
SR 3.1.4.6 j Verification of full-length control rod drop times q determines that the maximum control rod drop time is consistent with the assumed drop time used in that safety analysis (Ref. 2). The 2.5 second acceptance criteria is measured from the time the CRDM clutch is deenergized by the reactor protection system or test switch to 90%
{ insertion. This time is bounded by that assumed in the safety analysis (Ref.2). Measuring drop times prior to .
reactor criticality, after reactor vessel head reinstallation, ensures that reactor internals and CRDMs hgl i i
fd. will not interfere withgontrol rod motion or drop time and 4 '
I that no degradation in these systems has occurred that t
, ( would adversely affect
- control rod motion or drop time. g i
Inalviaualwontrol rods whose drop times are greater than i safety analysis assumptions are not OPERABLE. This SR is l performed prior to criticality, based on the need to j perform.this Surveillance under the conditions that apply '
- during a plant outage and because of the potential for an ;
unplanned plant transient if the Surveillance were -l performed with the reactor at power.
)
- Palisades Nuclear P1 ant B 3.1.4-11 01/20/98
. 6-g-
Shutdown and Part-Length Rod Group Insertion Limits B 3.1.5 8 3.1 REACTIVITY CONTROL SYSTEMS B 3.1.5 Shutdown and Part-Length Rod Group Insertion Limits BASES g BACKGROUND The insertion limits of the shutdown assumptions in all safety analyses that assum(control rod rods are(~
insertion upon reactor trip. The insertion limits directly affect core power distributions and assumptions of available SDM, ejected rod worth, and initial reactivity insertion rate.
The Palisades Nuclear Plant design criteria (Ref.1) and 10 CFR 50.46, " Acceptance Criteria for Emergency Core Cooling Systems for Light Water Nuclear Power Reactors,"
contain the applicable criteria for these reactivity and power distribution design requirements. Limits on shutdown rod insertion have been established, and all rod positions are monitored and controlled during power operation to ensure that the reactivity limits, ejected rod worth, and SDM limits are preserved.
The shutdown rods are arranged into groups that are radially symmetric. Therefore, movement of the shutdown l rod groups does not introduce radial asyumetries in the core power distribution. The shutdown and regulating rod groups provide the required reactivity worth for immediate reactor shutdown upon a reactor trip.
The Palisades Nuclear Plant has four part-length control rods installed. The part-length rods are required to
, remain completely withdrawn during power operation except during rod exercising performed in conjunction with SR l 3.1.4.4. The part-length rods do not insert on a reactor '
trip.
The design calculations are performed with the assumption that the shutdown rod groups are withdrawn prior to the regulating rod groups. The shutdown rods can be fully withdrawn without the core going critical. This provides (el available negative r ity for SDM in the event of ~
ALL cascoL _ boratiAn errors. 4 e dat nwn rod groups are controlled X manually by the control room operator. During normal plant operation, the shutdown rod groups are fully withdrawn.
The shutdown rod groups must be completely withdrawn from the core prior to withdrawing any regulating rods during an approach to criticality. The shutdown rod groups are then left in this position until the reactor is shut down.
Palisades Nuclear Plant B 3.1.5-1 01/20/98 6-h 4
Shutdocn and Part-Length Rod Group Insertion Limits B 3.1.5 BASES LCO Maintaining the shutdown rod groups within their insertion (continued) limits ensures that a sufficient amount of negative reactivity is available to shut down the reactor and !
2 maintain the required SDM following a reactor trip. l Maintaining the part-length rod group within its insertion l limit ensures that the power distribution envelope is I maintained.
APPLICABILITY The shutdown and part-length rod groups must be within i their insertion limits, with the reactor in MODES 1 and 2.
In MODE 2 the Applicability begins anytime any regulating rod is withdrawn above 5 inches. This ensures that a-sufficient amount of negative reactivity is available to .
shut down the reactor and maintain the required SDM l following a reactor trip. In MODE 4, 5, or 6, the shutdown l rod groups are inserted in the core to at least the lower i electrical limit and contribute to the SDM. In MODE 3 the '
shutdown rod groups may be withdrawn in preparation of a !
reactor startup. Refer to LC0 3.1.1, " SHUTDOWN MARGIN (SDM)," for SDM requirements in MODES 3, 4, and 5.
LC0 3.9.1, " Boron Concentration," ensures adequate SDM in MODE 6.
The Applicability has been modified by a Note indicating the LC0 requirement is suspended during SR 3.1.4.4 (rod exercise test). Control rod exercising verifies the M)g freedom of the rods to move, and requires the individual %
shutdown (dF4mrt Tif6Eh rods to move below the LC0 limits X for their group. Only the full-length rods are required to be tested by SR 3.1.4._4. The part-length rods may alspL yneedtobe eriodicaHy exercised to aintain mechanical
/ seal inte ity. Therefore, though t required part of R y 3.1.4.4, he part-length control r s may be exercised under th control' led conditions o SR 3.1.4.4. ,
Positioning of an individual control . rod within its group is addressed by LC0 3.1.4, " Control Rod Alignment."
IS mwed b(.lW 6 rnovr) h6me.wf) 14 c hM knph red h Mmd a fIrw umlahd I.co; h kct vin.d MC,iwd ch 0ndd% b much be. WF6
. 1 Palisades Nuclear Plant B 3.1.5-4 01/20/98 1 6-1
Regulating Rod Group Position Limits i B 3 1.6 !
B 3.1 REACTIVITY CONTROL SYSTEMS B 3.1.6 Regulating Rod Group Position Limits 4
BASES d l grovv )( )(
< BACKGROUND The insertion limits of the regulating roha%d re initial bytk assumptionsinallsafetyanalysesthatassume/ rod insertion upon reactor trip. The insertion limits directly bg affect core power distributions, assumptions of available 1 SDM, and initial reactivity insertion rate. The applicable criteria for these reactivity and power distribution design requirements are contained in the Palisades Nuclear Plant design criteria (Ref.1), and 10 CFR 50.46, " Acceptance
] Criteria for Emergency Core Cooling Systems for Light Water
- Nuclear Power Reactors" (Ref. 2).
Limits on regulating rod group insertion have been established, and all regulating rod group positions are 4
monitored and controlled during power operation to ensure that the power distribution and reactivity limits defined by the design power peaking, ejected rod worth, reactivity insertion rate, and SDM limits are preserved.
The regulating rod groups operate with a predetermined amount of position overlap, in order to approximate a linear relation between rod worth and rod position (integral rod worth). The regulating rod groups are withdrawn and operate in a predetermined sequence. The group sequence and overlap limits are specified in the COLR.
The regulating rods are used for precise reactivity control of the reactor. The positions of the regulating rods are manually controlled. They are capable of addin very quickly (compared to borating or diluting)g reactivity The power density at any point in the core must be limited to maintain specified acceptable fuel design limits, including limits that preserve the criteria specified in 10 CFR 50.46 (Ref. 2). Together, LCO 3.1.6; LCO 3.2.3,
" QUADRANT POWER TILT (T.)"; and LCO 3.2.4, " AXIAL SHAPE INDEX (ASI)," provide limits on control component operation and on monitored process variables to ensure the core operates within the linear heat rate (LCO 3.2.1, " Linear Heat Rate (LHR)") and radial peaking factor F/ and F/
(LCO 3.2.2, " Radial Peaking Factors) limits in the COLR.
4 Palisades Nuclear Plant B 3.1.6-1 01/20/98 6-j
Regulating Rod Group Position Limits .
B 3.1.6 BASES l 4' Fuel cladding damage does not occur when the core is APPLICABLE SAFETY ANALYSES operated outside these LCOs during normal operation.
(continued) However, fuel cladding damage could result, should an l accident occur with simultaneous violation of one or more i of these LCOs. Changes in the power distribution can cause j
- increased power peaking and corresponding increased local l a LHRs. l 1
The SDM requirement is ensured by limiting the regulating ,
and shutdown rod group insertion limits, so that the l allowable inserted worth of the rods is such that 1
sufficient reactivity is available to shut down-the reactor i to hot zero power. SDM assumes the maximum worth rod remains fully withdrawn upon trip (Ref. 4).
The most limiting SDM requirements for Mode 1 and 2 conditions at Beginning of Cycle (80C) are determined by the requirements of several transients, e.g., Loss of Flow, i etc. However, the most limiting 50M requirements for MODES 1 and 2 at End of Cycle (E0C) come from just one transient, Main Steam Line Break (MSLB). The requirements of the MSLB event at E0C for the full power and no load conditions are significantly larger than those of any other l event at that time in cycle and, also, considerably larger than the most limiting requirements at BOC. g i N Although the most limiting SDM requirements at EOC are much
! n larger than those at BOC, the available SDMs obtained via )( l tuff hp _trionina the3 control rods are substantially larger due to '
the much lower boron concentration at E0C. To verify that adequate SDMs are available throughout the cycle to satisfy the changing requirements, calculations are performed at both B0C and E0C. It has been determined that calculations ;
i at these two times in cycle are sufficient since the ,
l difference between available SDMs and the limiting SDM gy requirements are'the smallest at these times in cycle. The
[vd e$h rTneasurement okcontrol the Startup Testing rod Program banktf worththat demonstrates performed as part of #A)g the core has the expected shutdown capability. Consequently, adherence 1.Poy to LCO 3.1.5, " Shutdown and Part-Length Rod Group Insertion j
- Limits," and LC0 3.1.6 provides assurance that the di available at any time in cycle will exceed the limiting (
SDM requirements at that time in cycle.
. Sbm Palisades Nuclear Plant B 3.1.6-4 01/20/98 ,
'6-k
Shutdowng!nsertion Li:::its (Asfalsh) land fLN_e-48.4 kpl B 3.1 REACTIVITY CONTROL SYSTEMS b B 3.1 Shutdown lConp ol Ele #nt Assemb)'y (CEA)l Insertion 9 Limits [Ap
- ud f +-L<q4L.R.4 p}Q f(M BASES 4
BACKGROUND The insertion limits of the shutdown assumptions in all safety analyses that assume g nsertion upon reactor trip. The insertion limits directly affect are init h "*I"
('
e core pow r distributions and assumptions of available 50M, W ejected rth, and initial reactivity insertion rate. l bS f_fij $ 6/{ T" / % ufhe applicable criteria for these reactivity and power distribution design requirements.lare 10 fR 50, Appepdix A, (
- GDC 10 ' Reactor Design," and GDC 26, " eactivity (mi ts'
! (Ref. ), and 10 CFR 50. 6, "Acceptan Criteria r Emer ency, cera coolina vstems for L ht Water N clear Power Ra tors'i(RefS2). i.imits on shutdowr(23 insertion have M been established, and a11Tdl5 positions'are monitored and h controlled during power operation to ensure that the reactivity limits, ejected @ worth, and SDN limits are {h l
preserved.
Q
! The shutdown are arranged into groups that are radially l@ I
? synsnetri c. T erefore, movement of the shutdown C B % dent not introduce radial asynsnetries in the co wer *d W C b, distribution. The shutdown and regulating rovide the required reactivity worth for immediate reactor shutdown- !
upon a reactor trip. -4 ,,. p).
}
DoSERT 2. -a r (4 Y -r)
The design calculation are performed with the assumption ,
,,a th t the shutdown re withdrawn prior to the regulating l <.-ri . The shutdown can be fully withdrawn without the .
- core going critical. his provides available negativ. i reactivity for SOM in the event of boration errors. Uk t ,
%w"J.-of(snytqsWDWA
% Jy are controlled manuallylor/automatMallyTb the tontrol room operator. During nonnal GiaD operation,w @
ine shutcown M are fully withdrawn. The shutdown @Ql ,
must be completely withdrawn from the core prior to I withdrawing any regulating CIAn during an approach to f 3 I
__ g criticality. The shutdown position until the reactor is cia shut_1) are They then affect left incore this lL*J V*-* O down.
i power, burnup distribution, and add negative reactivity to '
shut down the reactor upon receipt of a reactor trip signal.
J (continued)
CE0G STS B 1.1-34 Rev 1, 04/07/95 o
. 6-1
\
a
- SECTION 3.1 4
! INSERT 6 hI I
$l'O The Applicabilty has been modified by a Note indicating the LCO requirement is suspended during SR 3.1.4.4 (rod exercise test). Control rod exercising verifies the freedom of the rods 4
to move, and requires the individual shutdownbr r2R-lenemirods to move below the LCO limits for their group. Only the full length rods are required to be tested by SR 3.1.4.4. The part-length rods may alsolneed to be peri ally exercised to maintai mecharucal seal integrity. They4 fore, though not require part of SR 3.1.4.4, the pa -length control rods ma
,be exercisedgnder the controlled condi ons of SR 3.1.4.4. __
Positioning of an individual control rod within its group is addressed by LCO 3.1.4, " Control Rod Alignment."
b C, M oucd b o w rJJc/j /hOL f0.O 41rk /d h m W
! bbtasSutofacvaraec.ic.k/cco, the dovde ,
z i 0 klbA inV$$ fQ i
't ge' B 3.1-36 6-m i
i l
CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO THE DECEMBER 04, 1998 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.1, REACTIVITY CONTROL SYSTEM NRC REQUEST:
3.1-04 ITS 3.1.4 [STS 3.1.5] Control Rod Alignment ITS 3.1.4 Required Action D Bases LCO section (page B 3.1-26) and Bases ACTIONS (page B 3.1-30)
! DOC M.6 JFD 10 and JFD 17
- The ITS has added a Required Action D that an imovable but trippable control rod
- shall be returned to operable status prior to entering Mode 2.
Comnent #2: The definition for operable control rod is at variance with the STS j definition. In the STS control rod operability is equated with trippability, not 1 movability. In the ITS the control rods must be trippable and movable to be ;
operable; for plants converting to the STS this is a plant unique definition, why? Recommend deleting Required Action D.
- Consumers Enerav Resconse:
1 i )
While it is acknowledged the ISTS equates control rod Operability with 1 trippability and not movability, ITS 3.1.4 has retained the CTS requirement that an immovable control rod is inoperable. This was done, in part, to preserve the operational flexibility in the CTS which precludes a forced plant shutdown in the event a single control rod becomes inoperable (immovable). For example; in the l ISTS an immovable (but trippable) control rod is considered Operable.
Correspondingly, the Bases for ISTS SR 3.1.5.5 explains that an immovable control rod is considered Operable if discovery is made between required performances of
- SR 3.1.5.5 (an SR 3.0.1 exemption). However, at the time the control rod fails l to meet the acceptance criteria for the freedom of movement test (ISTS SR 3.1.5.5), the control rod is declared inoperable and a shutdown to .
1
! Mode 3 is required. In comparison, an immovable control rod in the ITS is 4 declared inoperable and entry is made into the appropriate Required Actions which allow continuous operation. Since surveillances do not have to be performed on I inoperable equipment (SR 3.0.1), restoration of the inoperable control rod is not required until the plant enters Mode 2 from Mode 3.
Affected Subnittal Paaes:
See NRC Request number 3.1-04 comment #1 for Affected Submittal Pages.
i
' 7
l l
CONVERSION TO IMPROVED TECNNICAL SPECIFICATIONS l RESPONSE TO THE DECEMBER 04, 1998 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.1, REACTIVITY CONTROL SYSTEM HRC REQUEST:
3.1-04 ITS 3.1.4 [STS 3.1.5] Control Rod Alignment ITS 3.1.4 Required Action D Bases LC0 section (page B 3.1-26) and Bases ACTIONS (page B 3.1-30) ;
DOC M.6 JFD 10 and JFD 17 The ITS has added a Required Action D that an immovable but trippable control rod shall be returned to operable status prior to entering Mode 2.
Comment #3: The only element to Part Length Control Rod Operability is that they be fully withdrawn; they do not need to be either Trippable or Moveable.
Consumers Enerav Resconse:
Agree. Since the only element to Part Length control rod Operability is that they must be fully withdrawn, the ITS and ITS Bases have been modified as appropriate to identify the Conditions, Required Actions, and Surveillance Requirements that apply specifically to the full length control rods. This change should help !
clarify the Operability requirements associated with the Part Length control rods.
The change in wording, between CTS " control rod" and (revised) ITS " full length control rod," was necessitated by the ITS omission of the CTS definition of
" Control Rod" which states " CONTROL RODS shall be all full-length shutdown and regulating rods." The words " shutdown and regulating" need not be retained, because there are no other full length control rod types in the Palisades design.
Affected Submittal Paaes:
See NRC Request number 3.1-04 comment #1 for Affected Submittal Pages.
8
CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO THE DECEMBER 04, 1998 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.1, REACTIVITY CONTROL SYSTEM l NRC RE0UEST:
3.1-05 ITS 3.1.4 [STS 3.1.5] Control Rod Alignment
- ITS 3.1.4 Required Actions A and B, Completion Times ITS SR 3.1.4.1 and SR 3.1.4.2
- DOC A.4, DOC M.3 and JFD 19 The ITS adds new Required Actions to perform a rod position verification (SR 3.1.4.1) 15 minutes after control rod movement when either a channel of rod j position indication is inoperable or when the rod position deviation alarm is inoperable.
Cormsent U: The completion times should include a 15 minute requirement for when the inoperability is first discovered (i.e., "15 minutes MQ Once within...").
Consumers Enerav Resconse:
An initial performance of rod position verification (ITS SR 3.1.4.1) upon discovery that one channel of rod position indication is inoperable is not warranted based on the following: 1) Operability of the remaining indication channel, 2) knowledge of rod position prior to the loss of the indication channel, and 3) the routine performance of rod position verification every l 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. The proposed Completion Time is conservatively appropriate since failure of one of the two rod position channels simply represents a loss of :
- redundancy. In addition, since rod motion is performed manually (i.e., automatic j rod control is not used), the remaining indication channel is verified to ;
4 function as expected each time the affected control rods are moved.
1 Affected Submittal Paaes:
None l
l 1
I l
9
CONVERSION TO IMPROVED TECNNICAL SPECIFICATIONS ,
RESPONSE TO THE DECEMBER 04, 1998 REQUEST FOR ADDITIONAL INFORMATION l SECTION 3.1, REACTIVITY CONTROL SYSTEM l
l NRC REQUEST-3.1-05 ITS 3.1.4 [STS 3.1.5] Control Rod Alignment l ITS 3.1.4 Required Actions A and B, Completion Times l ITS SR 3.1.4.1 and SR 3.1.4.2 00C A.4, DOC M.3 and JFD 19 The ITS adds new Required Actions to perform a rod position verification (SR 3.1.4.1) 15 minutes after control rod movement when either a channel of rod position indication is inoperable or when the rod position deviation alarm is inoperable.
Comnent #2: Discuss how a rod position verification and a channel check differ.
Consumers Enerav Resoonse:
A rod position verification is a verification that the control rods are positioned and aligned as assumed in the safety analysis. For Palisades, this means that each control rod is aligned within 8 inches of all other control rods in its group. Verification of rod position can be obtained from either the primary or secondary rod position indicating channels.
A Channel Check is a assessment of channel behavior and is generally achieved by comparing the output of the primary rod position indication instruments to the output of the secondary rod position indication instruments. Thus, a Channel Check assures the instrumentation used to monitor control rod position is functioning properly.
Affected Submittal Paaes:
None 10
CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO THE DECEMBER 09, 1998 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.1, REACTIVITY CONTROL SYSTEM NRC RE0UEST:
3.1-06 ITS 3.1.5 [STS 3.1.6] Shutdown and Part Length Rod Group Insertion Limits ITS 3.1.5 Applicability JFD 6 and DOC A.6 The ITS applicability differs from both the STS and CTS by equating control rods withdrawn less than 5 inches with fully inserted control rods.
Comment: During startup, are the Regulating Control Rods " bumped" off the bottom < 5" before the Shutdown and Part length Control Rods are fully withdrawn?
Consumers Enerav ResD0nse:
During a plant startup the regulating rods may be <5 inches from the bottom of their travel before the shutdown and part length rods are fully withdrawn. This could result from " bumping" the regulating rods prior to an initial startup after a refueling outage or following a reactor trip or, based on the "as left" position of the regulating rods following a mid-cycle shutdown. The control rod drive system is designed with " lower electrical limit switches" which prevent individual control rods from being inserted beyond 3 inches (plus or minus limit switch uncertainties) from the bottom of their mechanical travel. Thus, when the regulating rods are manually inserted using their drive motor (versus from a j reactor trip signal), insertion is electrically interrupted approximately 3 i inches from the bottom of full (mechanical) rod travel. For an initial startup I after a refueling outage or following a reactor trip, the regulating rods may be l bumped off their lower mechanical stops (but less the 5 inches) to prevent thermal binding in the control rod drive piston guide tube.
Affected Submittal Pages:
None l
l l
1 11
CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO THE DECEMBER 04, 1998 REQUEST FOR ADDITIONA' INFORMATION SECTION 3.1, REACTIVITY CONTROL SYSTEM NRC REQUEST 3.1-07 ITS 3.1.5 [STS 3.1.6] Shutdown and Part Length Rod Group Insertion Limits Bases ITS 3.1.5 LC0 Section (page B 3.1-36) Insert 2 JFD 8
. The ITS 3.1.5 Bases LC0 paragraph includes clarifying inform 6 tion provided as Insert 2.
Coment: This information adds clarity and conservatism. Request that a TSTF be provided to incorporate this information into the STS.
Consumers Enerav Resoonse:
Upon further review of the information proposed in the Bases of ITS 3.1.5, it was determined that the addition of this information created the potential for a misapplication of the ITS. That is, anytime it is discovered that a control rod 4
can not be moved by its operator the Conditions of ITS 3.1.4 must be entered.
Sin'.e movement of the shutdown rods is typically limited to the control rod exercise test, the inability to restore a shutdown rod to within the limits of the LC0 would be indicative of an inoperable (i.e.. immovable) control rod.
Initially, ITS 3.1.5 Required Action A.1 allowed 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> to restore the shutdown or part-length to within the group limit. This was NOT intended to provide an additional 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> delay into the Conditions of ITS 3.1.4 for an inoperable control rod. Furthermore, the Bases stated " declaring a rod which is below its insertion limit, but within 8 inches of all other rods in its group, to be misaligned is acceptable." This statement was ambiguous since the only reason a rod would be misaligned is because it could not be realigned by its motor operator. Therefore, to eliminate potential confusion, ITS 3.1.5 Required Action A.1 has been revised to declare the affected control rod inoperable, and
- to enter the Conditions and Required Actions of LC0 3.1.4 immediately.
While it is analytically conservative for Palisades to declare a single control rod that is not within its insertion limits inoperable, it is not known whether
< this interpretation is appropriate for all CE designs. As such, this change has been proposed as a plant specific change.
12
CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO THE DECEMBER 04, 1998 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.1, REACTIVITY CONTROL SYSTEM Affected Submittal Paces:
Att 1 ITS 3.1.5, page 3.1.5-1 Att 2 ITS 3.1.5, page B 3.1.5-3 Att 2 ITS 3.1.5, page B 3.1.5-5 Att 3 CTS, page 3-53 (ITS 3.1.5, page 1 of 3)
Att 3 DOC 3.1.5, page 3 of 5 Att 3 DOC 3.1.5, page 4 of 5 Att 5 NUREG, page 3.1-13 Att 5 NUREG, page B 3.1-36 Insert Att 5 NUREG, page B 3.1-37 Att 5 NUREG, page B 3.1-37 Insert Att 6 JFD 3.1.6, page 4 of 4 I
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l 13
Shutdown and Part-Length Rod Group I".sertion Limits 3.1.5 3.1 REACTIVITY CONTROL. SYSTEMS b al 'I 3.1.5 Shutdown and Part-Length ^ Rod Group Inserti<'1. Limits LCO 3.1.5 All shutdown and part-length rod groups shall be withdrawn to 2 128 inches.
' APPLICABILITY: MODE 1, MODE 2 with any regulating rod withdrawn above'S inches.
NOTE----------------------------
This LCO is not applicable while performing SR 3.1.4.4 (rodexercisetest).
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME p\
w-m NI A. One or more shutdown A.1 ~ Restore hutdo n'anii 440ses or part-length rods part-1 gth r d 1 m g4%J an+,,t
-geespe not within group to wi hin red 6) msNdle a ad -
limit. limi . -
eMk,e % a fthAL. dond.hea
( bd kivi<*d Act% o &
Lc.o S. I.4.
B. Required Action and 8.1 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time n?t met.
SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.5.1 Verify each shutdown and part-length rod 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> group is withdrawn 2 128 inches.
Palisades Nuclear Plant 3.1.5-1 Amendment No. 01/20/98 13-a
J Shutdown and Part-Length Rod Group Insertion Limits B 3.1.5 l I
BASES l 1
APPLICABLE The acceptance criteria for addressing shutdown rods as !
SAFETY ANALYSES well as regulating rod insertion limits and inoperability !
(continued) or misalignment are that: !
- a. There be no violation of:
- 1. Specified acceptable fuel design limits, or
- 2. Primary Coolant System pressure boundary damage; -l and
- b. The core remains subcritical after accident transients. 1 C0' i
As such, the shutdown and part-length rod group insertion l limits affect safety analyses involving core reactivit p d ' y ejected rod worth, and SDM (Ref. 2). The part-length rods ,
have the potential to cause power distribution envelopes to j be exceeded if inserted while the reactor is critical.
Therefore, they must remain withdrawn in accordance with l the limits of the LC0 (Ref. 3). l The shutdown and part-length rod group insertion limits satisfy Criterion 2 of 10 CFR 50.36(c)(2).
The shutdown and part-length rod groups must be within SkD}
N l LC0 A
their insertion approaching limits For criticality. anyatime therod control reactor is be group to critical or {'h,w consideicd above its insertion limit, all rods in that I group Lother than miem14 p d .cde ;dd ::;;d y LCO ^,,1, ,
"C:r.tr:1 Y 71iv,,a.,;"i must be above the insertion limit. l If only on rod in a g oup is below the insertion limit, t the group ay be cons ered to be ab ve the limit if th rod is c sidered to e misaligned, and the appropriate !
conditio of LC0~3.1 4 is entered. Since LC0 3.1.4 wo ld I not all continued operation wit more than one rod !
misali ed, declar g a rod whic is below its group' l inser on limit, b t within 8 in hes of all other ro s in i t its oup, to be isaligned is cceptable. This ac ion may on1 be taken if 11 other con ol rods are proper y ali ned, l l
i l
Palisades Nuclear Plant B 3.1.5-3 01/20/98 l
13-b ,
1 i
Shutdocn and Part-Length Rod Group Insertion Limits B 3.1.5 BASES ACTIONS L1 Prior to entering this condition, the shutdown and part-length rod groups were fully withdrawn. If a shutdown rod group is then inserted into the core, its potential ,
negative reactivity is added to the core as it is inserted.
If the shutdo
~
or part-length rod roups are not wit n fWl'0
- limits, then hours is allowed fo restoring the ro
.L. M bI groups to thin limits. The 2 ur total Completio Time [
allows th operator adequate ti to adjust the rod groups
/ // in an or rly manner and is co istent with the re ired
- Completion Times in LC0 3.1. __
i fL1 When Required Action A.1 cannot be met or completed within the required Completion Time, a controlled shutdown should be commenced. The allowed Completion Time of 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> is reasonable, based on operating experience, for reaching MODE 3 from full power conditions in an orderly manner and without challenging plant systems.
)
SURVEILLANCE SR 3.1.5.1 REQUIREMENTS Verification that the shutdown and part-length rod groups are within their insertion limits prior to an approach to criticality ensures that when the reactor is critical, or i being taken critical, the shutdown rods will be available ,
to shut down the reactor, and the required SDM will be !
maintained following a reactor trip. Verification that the part-length rod groups are within their insertion limits ensures that they do not adversely affect power l distribution requirements. This SR and Frequency ensure t
that the shutdown and part-length rod groups are withdrawn before the regulating rods are withdrawn during a plant I startup. I Palisades Nuclear Plant B 3.1.5-5 01/20/98 ;
.13-c
l SECTION 3.1 INSERT 1 I ' 3-If one or more shutdown or part-length rods are not within limits, the affected rod (s) must be declared inoperable and the applicable Conditions and Required Actions of LCO 3.1.4 entered immediately. This Required Action is based on the recognition that the shutdown and part-length rods are normally withdrawn beyond their insertion limits and are capable of being moved by -
their control rod drive mechanism. Although the requirements of this LCO are not applicable during performance of the control rod exercise test, the inability to restore a control rod to within !
the limits of the LCO following rod exercising would be indicative of a problem affecting the OPERABILITY of the control rod. Therefore, entering the applicable Conditions and Required Actions of LCO 3.1.4 is appropriate since they provide the applicable compensatory measures commensurate with the moperability of the control rod.
t e
l 13-d
$ p e s : Sc. A'
- e T. l. 6 CONTadtR00ANDpdWROfSTRIRdIONLIMITS 3.@ Qce.c.t.' 4 C ,mh .) ,
[3M Shutdownhedfimits l' " ' ' ' ^$b "f r is n s < A .n) gr:edl.htCO l fp. All sh~utdown rocs snall be withdrawn before any regulating ,h' I rods are withdrawn f y g ., % -
- b. The shutdown rods shall not be withdrawn until normal water s l l level is established in the pressurizer.
)
aIfut,,Lll4 [cco [ The shutdown rods shall not be inserted below their exercise T limit until all re ulating rods are inserted. ~
S il N.10.7 j Low power A vsics Teshi~ D f l
< Sections (3.l(1m- V10.1.bl3.10.3. d.d9M.b, 3.f0.dand 3.10.6 N may be deviated from during in.m comer anyz xs tsuinn a m cpDM i
- Sec 3.17 Mi exercisesfif necessary to perform a test Dut only for .ie timp
- l (necessary to perform the test m ,
l A .8 i
, um :
Sufficient CONTR04. R shall be withdrawn at all time to assure that ,
- the reactivity decre e free a reactor trip provides adequate shutdown ;
j margin. The available worth of withdrawn rods must t'nclude the ,
- reactivity defect 'f power and the failure of the ipthdrawn rod of !
- highest worth to nsert. The requirement for a s tdown margin of 2.0% 1
! in reactivity w) h 4 pump operation, and of 3.75 in reactivity with !
, less than 4 pupp operation, is consistent with e assumptions used in i i the analysis af acciden conditions (including steam line break) as , !
! reported infleference 1 and additional analys s. Requiring the boron ncentration at less than g' 7,j concen*. rat $n to be at cold shutdown boron '
j hot shtitdown assures adequate shutdown ma n exists to ensure a return !
! to power does not occur if an unanticipa cooldown accident occurs. I i
- This r utrement applies to normal opera ng situations and not during l emerg cy conditions where it is necess y to perfons operations to
{ mitigate the consequences of an acciderft. By imposing a minimum ,
- shu down cooling pump flow rate of 2 0 gps, sufficient time is provided '
- f the operator to terminate a bor dilution under asynenetric !
, nditions. For operation with no rimary coolant pumps operating and a I
! ocirculating flow rate less than 810 goe the increased shutdown margin l and controls on charging pump op, rability or alternately the surveillance of the charging pps will ensure that the acceptance h- l i criteria,"for an iriidvertent ron dilution event will not be }.\- O l 4 l violated. ' The :hange in in artion limit with reactor power insures '
l
! that the shutd'sen requirene s for 4 pump operation is met at all power levels. The 2,5 second de o time specified for the CONTROL R005 is the 1 i
dror time sed in the tr sient analysis.'" ;
Db b { b*
knendment No. H, 64, p, 64. H4, W. W.169 July 26, 1995 3 53 i
13-e it.g\b
ATTACIBiENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.1.5, SHUTDOWN AND PART-LENGTH ROD GROUP INSERTION LIMITS A7 CTS 3.10.6a states "All shutdown rods shall be withdrawn before any regulating rods are withdrawn." CTS 3.10.6c states "The shutdown rods shall not be inserted below their exercise limit until all regulating rods are inserted." The proposed ITS 3.1.5 LCO states "All shutdown and part/ length rod groups shall be withdrawn to a 128 inches." The Applicability for LCO 3.1.5 is MODE 1, MODE 2 with any regulating rod withdrawn above 5 inches. The proposed ITS wording for the LCO and Applicability is equivalent to the CTS wording in 3.10.6b. In the ITS, the shutdown rods must be withdrawn 2128 inches by the LCO before the regulating rods are withdrawn above 5 inches (see DOC A.6 for discussion on 5 inches criteria). In addition, the CTS 3.10.6c requirement that the shutdown rods cannot be inserted below their exercise limit is also maintained in the ITS. This is because the shutdown rods cannot be inserted, except for rod exercising allowed by Applicability note, until out of the MODE of Applicability which required the regulating rods to be s 5 inches withdrawn. Therefore, the CTS and the proposed ITS are equivalent.
A.8 CTS 3.10.7 includes an exception which allows a deviation from the requirement for
. shutdown rod limits during performance of CRDM exercises. The exception contains a ;
qualifying statement which reads "if necessary to perform a test but only for the time !
necessary to perform the test." The Applicability Note for proposed ITS 3.1.5 which ;
also provides an exception from the requirement for shutdown rod limits during )
performance of CRDM exercise does not contain this same qualifier since these type j details are governed by the usage rules for the ITS. Therefore, deletion of this 1
- information is considered administrative in nature. This change is consistent with NUREG-1432.
l Adh) 4 l AS Se wr# Ap rem coc m.1) (\ v ;
Palisades Nuclear Plant Page 3 of 5 01/20/98 13-f
i r#b 5 @l &07 ATTACHMENT 3 DISCUSSION OF CHANGES
% SPECIFICATION 3.1.5, SHUTDOWN AND PART-LENGTH ROD GROUP
- INSERTION LIMITS MORE RESTRICTIVE CHANGES (M) l, Thm wm no "inae km.w. , chp cuuaw a #u SM&oun.
M.9 CTS 3.10.3 and CTS 3.10.6 stipulate the requirement for rod position on an individual y 4 rod basis (i.e., all shutdown and part-length rod must be fully withdrawn). In addition, CTS 3.4.10.4a requires that a control rod must be aligned within 8 inches from the remainder of the bank. The CTS does not specify rod positions on a group basis, and does not contain actions when controls rods are misaligned from their groups by less i than 8 inches. Proposed ITS 3.1.5 establishes insertion limits for the shutdown and 1 part-length rod groups by requiring them to be withdrawn a 128 inches. Required g Action A.1 of ITS 3.1.5 requires that any shutdown or part-length rod group that is not hMk - Wwidiin im..; Wik ' " If the 4 within its group insertion limit by- y
' rd h e of equired Action and associacea to pletion Time are not met, Required Action B.1
- If requires the plant to be in Mode ithin 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. To ensure compliance with the z
4 20 requirements of LCO 3.1.5, for a control rod group to be considered above its ImdaW insertion limit, all rods in that group Ses inn *HW rue addwd by x )
I LCO 3.1.4, "Cor.au; ".~4 A"r:::") must be above the insertion limit. )If only one )
, "od in a grou is below e insertion I , the group may be considered be above l
.he limit if t at rod is onsidered to be isaligned, and the appropriate ndition of l LCO 3.1.4 s enter . Since LCO 3. 4 would not allow continued o ration with note tha one r misaligned, decla ng a rod which is below its gr p's insertion l imit, b withi inches of all oth rods in its group, to be misali ed is acceptable. , !
The R uired ctions of the ITS e more restrictive than the CT since the ITS limits /
the n mber control rods with an be misaligned from their ero n by less than >
Bi .
es t only one rod [Therefnre $e additioq ofITS Required Actions A.1 and i 0-B.1 is characterized Co.n aat; =c=d:rWinddWr.- 2 i 9/ changet. cant.L fu & fun kl
.s W . n fo.it J q is ruf e=J Hs inocNior) .kmof /3 066*M 88M N l LESS hhSTiblV Cb'db NhN DETAILS TO LICENSEE .
CONTROL, LED DOCUMENTS (LA)
LA.1 CTS 3.10.6b states "The shutdown rods shall not be withdrawn until normal water level is established in the pressurizer." This requirement was included to help assure an inadvertent criticality will not occur with the PCS water s4id. This statement is more appropriate for being addressed in plant procedures and is not included in the !
proposed ITS. Changes to plant procedures are made in accordance with the plant )
procedure change process. This change maintains consistency with NUREG-1432.
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Palisades Nuclear Plant Page 4 of 5 01/20/98 l 1
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757F.IM/g Shutdocn Insertion Liaits gy i
.. @ .4 P & LengA R.J C,ry 5 3.1 REACTIVITY CONTROL SYSTEMS 3.1[ Shutdown EnWol Elemery( Assembly /CEAjj I,nsertion Limits [Walp%h h and Ec t lead ( ke $cowp} h 3.1 @g
~
LCO All shutdown @ shallbe withdrawn to 2 M finches, cd end.le @ cod 3 < *- P3
' ' ^ ' ' ^
Q APPLICABILITY: MODE 1, N00E 2 with any regulating ICEA nyt fully Wsertedl2
............................N0TE--- -.------------- - .....
This LCO is not applicable while perforsing SR 3.1 .
l(
...................................................y.......
(<.a c.- c;<e test)
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more shutdown A.I.1 Ve ify SDM 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> g not within limit. 2 ( 4. 5 ]f, A k/ k. l h lor p.et-le,,+k e d y*=t5] g ,
A.I.2 Initiate boration to I hour re e SDM to w thin h7f l>l }
b .,
A.
AS Rest e shutdow f I to withi A it. ,
j Q a s:t - \e + r of_k -$2-e B. Required Action and B.! Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion l Time not est.
D u.k.r<. G.R h.ch d C w 4 rol
( f'od d ,O o p (d.blt Grd
/? enke tkapphccJok Osnch+wnsar.aIfcqu,ru)
/htnrJ of (,,(,o 3./M, CEOG STS 3.1 13 Rev 1, 04/07/95 13-h
SECTION 3.1 INSERT 1 1
The part-length rods have the potential to cause power distribution envelopes to be exceeded if insened while the reactor is critical. Therefore, they must remain withdrawn in accordance with the limits of the LCO (Ref. 3).
EN D For a control rod group to be considered above its insenion limit, all rods in that group (cr.;
ILs ranc'i;;rd ^t J p.dt==d i,y LCO 3.1.4, "Co;re! P~i A!!;-*"4 must be abovelhe insertion limit.J If o y one rod in a group is below insenion limit, the group may } 4 ove the limit if that rod is consi red to be misaligned, and the app priate 1
) considered to be condition of LC 3.1.4 is entered. Since LCO 3 .4 would not allow continued ope tion with more th one rod misaligned, declaring a od which is below its group's inse ion limit, l but within 8 hes of all other rods in its gro p, to be misaligned is acceptable. his action i may only taken if all other control rods e properly aligned. l Maintaining the shutdown rod groups within their insenion limits... i l
l INSERT 3 j l
Maintaining the pan length rod group within its insertion limit ensures that the power distribution envelope is maintained. l l
INSERT 4 In MODE 2, the Applicability begins anytime any regulating rod is withdrawn above 5 inches.
INSERT 5
...to at least the lower electrical limit, and contribute to the SDM, In MODE 3, the shutdown rod groups may be withdrawn in preparation for a reactor stanup.
B 3.1-36 i
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Shutdown,CEAInsertionLicitshnpogD BASES Cs IUt .qw I Ad Grwp 7 ACTIONS h A .1LkK2/wd API) (continued) core as it is insertad. lif bor concentration is t 7
fchanged at t is time, 50M shou not change. This, however, ;
gt i F,M u [s.,h is verified ithin I hour, or oration is initiate to bring l the Son to itnin i$mit, if 1 e Cg4(,) 3, not ,es ,,o to within lie s crior to this ime.
~
f -
.lf the>M d,,n,,ot festorac m within limitsMthtfi VPdur) 1 _
- Sh aw s o r ,
%Ft- s,np land the 5DM if within IPnMf, then (Ti additionaDgnourt g I rh\ red grdP J ;
' allowed for restoring 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> total Comple CEA(s) to i thin limits. TP ime allows the operator ade Jate
-\ \, t> in an orderly manner and time to adjust thea s A h ,I consistent with th required Completion Times in
, @optfolM1 einn t Is semb I y (uA H MTc# enc) CO3.lh.)]
Sm i u When Required Action A.lG7~ED eannot be met or completed 3 within the required Completion Time, a controlled shutdown should be comenced. The allowed Completion Time of 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> ,
is reasonable, based on operating experience, for reaching l MODE 3 from full power conditions in an orderly manner and '
without challenging plant systems.
l SU'tVEILLANCE SR 3.1.6.1 ad pmly red gewef REQUIREMENTS Verification that the shutdown are within their insertion limits prior to an approach to criticality ensures that when the ctor is critical, or being taken critical, O/% ridt_theshutdown will be available to shut down the c W
m reactor, and e r6 quired SOM will be maintained following a reactor trip.s This SR and Frequency ensure that the
~f)ftRT()
v snutcown, Urn are withdrawn before the regulating (t s are
} withdrawn during a JULTD*startup.
\'
p 't
/) %@d Va.c+ Jk[]f Since the shutdowif@ are positioned manually b MD b fed {nyr . the i
control room operator, verification of shutdow ,( position at a Fre c of 12 nours is acequate 7,o ensure that the shutdowF are within their insertion limits. Also, the ,
12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Frequency takes into account other information '
available to the operator in the control room for he purpose of monitoring the status of the shutdown (continued)
CEOG 3TS 8 3.1-37 Rev 1, 04/07/95 13-j
J -: - -
SECTION 3.1 l INSERT 1 &$ b. ~ ;
l l
If one or more shutdown or part-length rods are not within limits, the affected rod (s) must be declared inoperable and the applicable Conditions and Required Actions of LCO 3.1.4 entered )
immediat0v. This Required Action is based on the recognition that the shutdown and part-length 1 rods are normally withdrawn beyond their insertion limits and are capable of being moved by their control rod d.ive mechanism. Although the requirements of this LCO are not applicable i during performance of the control rod exercise test, the inability to restore a control rod to within - l the limits of the LCO following rod exercising would be indicative of a problem affecting the ]
OPERABILITY of the control rod. Therefore, entering the applicable Conditions and Required Actions of LCO 3.1.4 is appropriate since they provide the applicable compensatory measures )
commensurate with the inoperability of the control rod.
1 i
INSERT 2 l l
Verification that the part-length rod groups are within their insertion limits ensures that they do l not adversely affect power distribution requirements.
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ATTACHMENT 6 JUSTIFICATION FOR DEVIATIONS SPECIFICATION 3.1.6, SHUTDOWN CEA INSERTION LIMITS Chance Discuccion
- 14. The NUREG-1432 Bases in the Applicability section states "In MODE 3,4,5, or 6, the shutdown CEAs are fully insened in the core and contribute to the SDM. In the proposed ITS, MODE 3 was deleted from this sentence end another sentence added to state "In MODE 3, the shutdown rod groups are not always fully inserted. In addition, the term " fully inserted" is changed .i the proposed ITS to state "to at least the lower electrical limit." This change is made ') remove confusion with respect to what constitutes " full inserted." For the Palisades control rod design, the lower electrical limit corresponds to the point where electrical rod insertion ceases, and is about 3 inches from the bottom of full rod travel. The reactivity level in this region is negligible. These changes are plant specific changes to provide clarification of the requirements for shutdown rod groups.
- 15. To reflect the incorporation of TSTF-136 which consolidates ISTS 3.1.1 and ISTS 3.1.2, the specification number for ISTS 3.1.6, " Shutdown CEA Insertion Limits," has been changed to ITS 3.1.5 and conforming changes have been made to the Bases. These changes are consistent with NUREG-1432 as modified by TSTF-136.
- 16. The definition of Shutdown Margin was revised in NUREG-1432 to clarify that changes in fuel and moderator temperature are included in the determination of the Control Element Assembly Power Dependent Insertion Limits which are used to ensure adequate Shutdown Margin in MODES 1 and 2. As a result of this change, ISTS 3.1.6 Required Action A.1.1 (verify SDM) and Required Action A.1.2 (initiate boration) have been deleted since they are no longer necessary to ensure adequate Shutdown Margin. Therefore, these Required Actions and associated Bases discussions are not i
included ir. prep <v'.a ITS 3.1.5. This change is consistent with NUREG-1432 as modified by TSTF-67.
RQ) 3 l- O] l
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Palisades Nuclear N - Page 4 of 4 01/20/98 l l
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- 17. ISTS 3.1.6 Required Acticn A.1 (as modified by TSTF-67) allows 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> to restore out-of-limit shutdown rods to within the limit of the LCO. Proposed ITS 3.1.5 Required' Action A.1 requires out-of limit shutdown (and part-length) rods to be declared j _ inoperable and the Conditions and Required Actions ofITS 3.1.4 entered immediately.-
Anytime it is discovered that a control rod can not be moved by its operator the control : .
j rod must be considered inoperable. Since movement of the shutdown rods is typically.
limited to the control rod exercise test, the inability to restore a shutdown rod to within
! the limits of the LCO would be indicative of an inoperable (i.e., immovable) control rod.
Therefore, the Required Actions for a shutdown rod outside its specified limit has been l
changed to be consistent with the Required Actions for an inoperable control rod.
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CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO THE DECEMBER 09, 1998 REQUEST FOR A.DDITIONAL INFORMATION i
SECTION 3.1, REACTIVITY CONTROL SYSTEM l
NRC REQUEST l l
3.1-08 ITS 3.1.6 [STS 3.1.7] Regulating Rod Group Position Limits ITS 3.1.6 Required Action B, Completion Time ITS SR 3.1.6.1 DOC A.4 and JFD 5 The ITS adds a new Required Action to perform a rod position verification (SR 3.1.6.1) 15 minutes after control rod movement when either the PDIL Alarm Circuit or the CR005 Alarm Circuit are inoperable.
Comnent: The completion times should include a 15 minute requirement for when the inoperability is first discovered (i.e., "15 minutes MQ Once within...").
Consumers Enerav Resoonse:
An initial performance of group position verification (ITS SR 3.1.6.1) upon discovery that the PDIL or CR00S alarm circuit is inoperable is not warranted based on the following; 1) violation of the power dependent insertion limit or the mis-sequence control rod groups can only occurs as a result of control rod movement, 2) knowledge of rod group position prior to the loss of the indication
- channel, and 3) the routine performance of rod group position verification every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. The proposed Completion Time is appropriate since rod positioning is performed manually (i.e., automatic rod control is not used), and verification of rod group position is performed within 15 minutes following rod motion. This Completion Time is also consistent with the CTS.
Affected Submittal Paaes:
None 4
14
CONVERSION TO IMPROVED TECNNICAL SPECIFICATIONS RESPONSE TO THE DECEMBER 04, 1998 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.1, REACTIVITY CONTROL SYSTEM NRC REQUEST 3.1-09 !TS 3.1.6 [STS 3.1.7) Regulating Rod Group Position Limits ITS 2.1.6 LC0 and Required Action B q ITS SR 3.1.6.1 DOC M.1 and JFD 10 The ITS includes explicit sequence and overlap requirements in the LCO, Required Actions and in SR 3.1.6.1.
Comment: This information adds clarity and conservatism. Request that a TSTF
- be provided to incorporate this information into the STS.
Consumers Enerav Resoonsit:
Neither NUREG-1432 (ISTS for CE Plants) nor NUREG-0212 (STS for CE Plants) contain a requirement for control rod group " overlap." During the development of NUREG-1432 the subject of an overlap requirement was discussed with the ,
participating CE plants. At that time it was felt that an overlap requirement l i
was not needed.
Palisades will propose a generic change to NUREG-1432 at the next meeting of the CE Owners Group Licensing Subcommittee to include an explicit rod group overlap requirement in the LC0 for ISTS 3.1.7.
I Affected Submittal Paaes: ,
None 15
CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO THE DECEMBCR 04, 1998 REQUEST FOR ADDITIONAL INFORMATION l SECTION 3.1, REACTIVITY CONTROL SYSTEM FRC REMEH 3.1-10 ITS 3.1.6 [STS 3.1.7] Regulating Rod Group Position Limits I Bases ITS 3.1.6 LC0 Section (page B 3.1-42) Insert 2 I JFD 13 l
The ITS 3.1.6 Bases LC0 paragraph includes clarifying information provided as :
Insert 2.
Coment: This information adds clarity and conservatism. Request that a TSTF be provided to incorporate this information into the STS.
Consumers Enerav Resoonse:
- \
Consistent with the response to NRC Comment 3.1-07, the Bases of ITS 3.1.6 has been revised to eliminate information that was found to be ambiguous. The )
revised Bases still clarifies that all rods in a given group must be above the ]
insertion limits in order for the group to be considered within its insertion i limits. Unlike the shutdown rods discussed in ITS 3.1.5, the regulating rods are moved as a group in response to changing plant conditions. As such, violation of the insertion limits on a group basis is possible. Thus, maintaining a 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> restoration period (consistent with the CTS and ISTS) is appropriate.
. i While it is analytically conservative for Palisades to declare a single control l rod that is not within its insertion limits inoperable, it is not know whether this interpretation is appropriate for all CE designs. As such, this change has been proposed as a plant specific change. l l
Affected Submittal Poaes:
Att 2 ITS 3.1.6, page B 3.1.6-5
, Att 5 NUREG, page B 3.1-42 Insert Att 6 JFD 3.1.7, page 4 of 5 1
1 l
16
t Regulating Rod Group Position Limits B 3.1.6 '
4 BASES APPLICABLE Operation at the insertion limits or ASI limits may \
SAFETY ANALYSES approach the maximum allowable linear heat generation rate (continued) or peaking factor, with the allowed T, present. Operation at the insertion limit may also indicate the maximum ejected rod worth could be equal to the limiting value in .
fuel cycles that have sufficiently high ejected rod worth.
The regulating and shutdown rod insertion limits ensure that safety analyses assumptions for reactivity insertion rate, SDM, ejected rod worth, and power distribution peaking factors are preserved.
The regulating rod group position limits satisfy Criterion 2 of 10 CFR 50.36(c)(2). l LCO The limits on regulating rod group seyden.e overlap, and physical insertion, as defined in the COLR, must be maintained because they serve the function of preserving power distribution, ensuring that the SOM is maintained, ensuring that ejected rod worth is maintained, and ensuring adequate negative reactivity insertion on trip, The overlap between regulating rod groups provides more uniform rates of reactivity insertion and withdrawal and is imposed .
to maintain acceptable power peaking during regulating rod group motion.p For a control rod group to be considered above its 0 insertion ... nit, all rods in that group (ether the -
miulinned ende addrace d by Len 2.1.1, "C;r, 7;l Lj J.'i;; .=t"} must be above_ the insertion ' imit. IIf only F rod in a grou is below the insertion 1 iit, the group y ,
be considere to be above the limit if hat rod is considered o be misaligned, and the propriate condi ion of LC0 3.1 4 is entered. Since LCO .1.4 would not a ow continued peration with more than e rod misaligne ,
declarin a rod which is below its roup's insertion, limit, but with n 8 inches of all other ds in its group, to be misalig ed is acceptable. This tion may only be aken if All_ ot r control rods are prope ly aligned.
Palisades Nuclear Plant B 3.1.6-5 01/20/98 16-a
SECTION 3.1 INSERT 1 The most limiting SDM requirements for Mode 1 and 2 conditions at (Beginning of Cycle (BOC) are determined by the requirements of several transients, e.g., Loss of Flow, etc.
However, the most limiting SDM requirements for Modes 1 and 2 at End of Cycle (EOC) come fromjust one transient, Main Steam Line Break (MSLB). The requirements of the MSLB event at EOC for the full power and no load conditions are significantly larger than those of any other event at that time in cycle and, also, considerably larger than the most limiting requirements at BOC.
- Although the most limiting SDM requirements at EOC are much larger than those at BOC, the available SDMs obtained via tripping the control rods are substantially larger due to the much lower boron concentration at EOC. To verify that adequate SDMs are available throughout the cycle to satisfy the changing requirements, calculations are performed at both BOC and EOC. It has been determined that calculations at these two times in cycle are sufficient since i the difference between nvailable SDMs and the limiting SDM requirements are the smallest at '
these times in cycle. The measurement of control rod banyworth performed as pan of the Stanup Testing Program demonstrates that the core has the expected shutdown capability.
k Consequently, adherence to LCO 3.1.5, " Shutdown and Part-Length Rod Group Insertion b
Limits," and LCO 3.1.6 provides assurance that the available SDgat any time in cycle will exceed the limiting SDM requirements at that time in cycle. M
[b1 INSERT 2 sP10 For a control rod group to be considered above its insertion limit, all rods in that group 4othee-t h - :- H y* -e
- W --s* hy ! N ' ' 1 * ' F -J #- - - - ; must be above the f' l insertion limit. IIf only e rod in a group is below the inscruon I it, the group rr '
considered to be above e limit if that rod is considered to be m' aligned, and the agpropriate condition of LCO 3.1 is entered. Since LCO 3.1.4 would no allow continued operation with more than one od misaligned, declaring a rod whi:h is low its group's insertion limit, but within 8 inche of all other rods in its group, to be misal' ned is acceptable. This action ,
may only be tak if all other control rods are properly ali ed. ,
l B 3.1-42 16-b
ATTACHMENT 6 JUSTIFICATION FOR DEVIATIONS SPECIFICATION 3.1.7, REGULATING CEA INSERTION LIMITS Change Discussion
- 12. The Palisades Nuclear Plant analysis does not model separate insertion limits for transient and steady state conditions as specified in Conditions A, B and C of NUREG-1432. The Palisades Nuclear Plant PDIL limits specify the regulating rod group position limits which account for anticipated power maneuvers and transient mitigation. Therefore, the proposed Palisades ITS removes the steady state and transient insertion limit discussion, where appropriate, and provides a discussion of the Palisades Nuclear Plant insertion limits. This is a plant specific change to reflect the Palisades CTS and analysis. g S), M
- 13. A discussion has been added in the Bases under the LCO section to clarify thatjif a ..
' individual regulat g rod does not meet the alignme t requirements of LCO 3.1.4,
" Control Rod A gnment," then LCO 3.1.4 may entered as long as the remai er oi
. the group is a ve its insertion limits. This disc ssion was added to help avoid confusion sin LCO 3.1.6 is written to addres regulating rods on a group bas' and i I LCO 3.1.4 dresses individual rod misalig nts. lThis is a plant specific change to reflect the Palisades control rod design and CTS requirements.
14 To reflect the incorporation of TSTF-136 which consolidates ISTS 3.1.1 and ISTS 3.1.2, the specification number for ISTS 3.1.7, " Shutdown CEA Insertion Limits," has been changed to ITS 3.1.6 and conforming changes have been made to the Bases. These changes are consistent with NUREG-1432 as modified by TSTF-136.
- 15. The definition of Shutdown Margin was revised in NUREG-1432 to clarify that changes in fuel and moderator temperature are included in the determination of the Control Element Assembly Power Dependent Insertion Limits which are used to ensure adequate Shutdown Margin in MODES 1 and 2. As a result of this change, ISTS 3.1.7 )
Required Action A.1.1 (verify SDM) and Required Action A.1.2 (initiate boration) have been deleted since they are no longer necessary to ensure adequate Shutdown Margin. Therefore, these Required Actions and associated Bases discussions are not included in proposed ITS 3.1.6. An expanded discussion has been incorporated in the l Applicable Safety Analyses portion of the Bases to clarify the requirements for SDM as l it applies to control rod position. These change are consistent with NUREG-1432 as modified by TSTF-67.
Nor a C.en%L rod (sup h k ConDdmd aloon 5 mDsdio *k 2Il rodt in k i rouf inJak bt abo)C & Inowfm im h Palisades Nuclear Plant Page 4 of 5 01/20/98 16-c
i CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO THE DECEMBER 04, 1998 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.1, REACTIVITY CONTROL SYSTEM NRC REQUEST 4
3.1-11 ITS 3.1.7 [STS 3.1.9] Special Test Exception ITS 3.1.7 LC0 Requirements JFD 17 The ITS changes the STS SOM requirement to "21% shutdown reactivity...."
Consent: What is the value of "1% shutdown reactivity" based upon?
Consumers Enerav Resoonse:
The value of "1% shutdown reactivity" is based on engineering judgement and is intended to provide adequate negative reactivity to shut down and maintain the reactor subcritical during Physics Testing and includes margin for calculational uncertainties.
effected Submittal Paaes:
None a
17
CONVERSION TO INPROVED TECHNICAL SPECIFICATIONS RESPONSE TO THE DECEMBER 04, 1998 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.1 REACTIVITY CONTROL SYSTEM NRC REQUEST 3.1-12 ITS 3.1.7 [STS 3.1.9] Special Test Exception ITS 3.1.7 Required Actions B and D JFD 14 and JFD 15 The ITS revises the STS Required Actions making them more logical.
Comment: This information adds clarity. Request that a TSTF be provided to incorporate this information into the STS.
1 Consumers Enerav Resoonst:
Palisades will propose a generic change to NUREG-1432 at the next meeting of the CE Owners Group Licensing Subcommittee to revise the Required Actions associated with ISTS 3.1.9 to make them more logical.
j
&ffected Submittal Paaes:
l None I
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CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO DECEM8ER 09, I?98 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.2, POWER DISTRIBUTION LIMITS NRC REQUEST:
3.2-01 ITS 3.2.1 Linear Heat Rate (LHR)
ITS 3.2.1 LC0 JFD 8 The ITS 3.2.1 LC0 adds, "as determined by an OPERABLE Incore Alarm System or by an OPERABLE Excore Monitoring System," which is neither in the CTS nor the STS.
Consent: The wording of the LCO precludes the Condition A option of "DE LHR, as determined by manual incore readings, not within limits...." Suggest that the LCO be reworded to add the straight forward requirement that the Incore Alarm System and the Excore Monitoring System shall both be operable.
Consumers Enerav Resoonse:
The CTS allows LHR to be monitored by either the Incore Alarm System, or the Excore Monitoring System. If the Incore Alarm System is inoperable and the Excore Monitoring System is not being used to monitor LHR, operations are allowed to continue provided power is reduced to 85% Rated Power and incore readings are manually recorded. Specifying an Operable Incore Alarm System or Excore Monitoring System in the LC0 is necessary to support the structure of the ITS while maintaining the flexibility provided in the CTS. That is, the Incore Alarm System and the Excore Monitoring System must be inoperable (proposed ITS Condition B) before reliance is placed on the manual method of verifying LHR. The third entry in Condition A (LHR, as determined by manual incore detector reading, not within limits specified in the COLR) is necessary to ensure the LHR limits are not violated.
It was correctly identified by the NRC reviewer that the LC0 wording was not straight forward. As such, the LC0 wording has been revised to clearly require the Incore Alarm System or Excore Monitoring System to be Operable for monitoring LHR. In addition, the term "Incore Monitoring System" has been replaced with the term "Incore Alarm System" throughout Specification 3.2.1 and its associated Bases to eliminate the ambiguity of the LC0 requirement.
Conforming changes have also been made to the supporting documents.
19
CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO DECEM8ER 09, 1998 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.2, POWER DISTRIBUTION LIMITS Affected Submittal Paaes Att 1 ITS 3.2.1, pg 3.2.1-1 Att 1 ITS 3.2.1, pg 3.2.1-2 Att 1 ITS 3.2.1, pg 3.2.1-3 Att 2 ITS 3.2.1, pg 8 3.2.1-2 Att 2 ITS 3.2.1, pg B 3.2.1-3 Att 2 ITS 3.2.1, pg B 3.2.1-5 Att 2 ITS 3.2.1, pg B 3.2.1-6 Att 2 ITS 3.2.1, pg B 3.2.1-7 Att 2 ITS 3.2.1, pg B 3.2.1-8 Att 2 ITS 3.2.1, pg B 3.2.1-9 Att 3 DOC 3.2.1, pg 2 of 7 Att 3 DOC 3.2.1, pg 4 of 7 Att 5 NUREG 3.2.1, pg 3.2-1 Att 5 NUREG 3.2.1, pg 3.2-3 ;
Att 5 NUREG 3.2.1, pg B 3.2-4 insert Att 5 NUREG 3.2.1, pg B 3.2-5 insert Att 6 JFD 3.2.1, pg 1 of 5 itt 6 JFD 3.2.1, pg 3 of 5 Att 6 JFD 3.2.1, pg 4 of 5 20
SAl 12.-Of LHR 3.2.1 3.2 POWER DISTRIBUTION LIMITS l M bli k forem h thd3 h 8 b 'o Yk M O.r4 4A Inwt AlarA Sfab OR i i 3.2.1 Linear Heat Rate (LHR) { dg gg g gg$gtlf, l h/ io rn on i+of., LHR. I LC0 3.2.1 LHR, as d ermined by an PtKAULL Incore niternig System '
. or by an PERABLE Excor Monitoring Syste , shall be within Jhe _1_i_m ts specified i the COLR.
Kl '
APPLICABILITY: MODE 1 with THERMAL POWER > 25% RTP.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME 4
A. LHR, as determined by A.1 Restore LHR to within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> 1 limits.
"--iter"( M )(
Moma+ic, !
then System,Incore not within l limits specified in the COLR, as indicated l by four or more coincident incore channels. j E \
l LHR, as determined by l the Excore Monitoring i i
System, not within limits specified in the COLR. I M
LHR, as determined by l
. manual incore detector readings, not within limits specified in the COLR.
l 1
l Palisades Nuclear Plant 3.2.1-1 Amendment No. 01/20/98 20-a
_. - . _. . - - . . . . . . - . . - .- _ .- ~ . . - . . . --- .
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l LHR l
3.2.1- !
i ACTIONS j i
i CONDITION COMPLETION TIME REQUIRED ACTION l l
l l i
1 l B. Incore Alarm and B.1 Reduce THERMAL POWER 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> j Excore Monitoring to s 85% RTP. ,
l Systems inoperable for ;
'! monitoring LHR. E n J
l l Wr.G LHRis tudhin JINS j B.2 C-t:..-ir.; Uta using 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />
- manual incore
! readings. E Once per 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> '
i thereafter i
k '
\
C. Required Action and C.1 Reduce THERMAL POWER 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> l associated Completion to s 25% RTP.-
4 Time not met.
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- SURVEILLANCE REQUIREMENTS a
SURVEILLANCE FREQUENCY g SR 3.2.1.1 -.----.-.-----.-..-NOTE . f. l M M ----.--
j Only required when Incorchiterir; System is being used to monitor LHR.
(
Verify LHR is within the limits specified 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />
- in the COLR.
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, Palisades Nuclear Plant 3.2.1-2 Amendment No. 01/20/98 20-b J
LHR 13.2.1 SURVEILLANCE RE0'lIREMENTS SURVEILLANCE FREQUENCY AAl3,7/0!
'~
SR 3.2.1.2 -------------------NOTE---J$bar/Y'--------
Only required when Incore "eait;rin; System g
- is being used to monitor LHR.
Adjust incore alarm setpoints based on a Prior to measured power distribution. operation > 50%
RTP after each fuel loading MQ 31 EFPD thereafter SR 3.2.1.3 -_-----------------NOTE--------------------
Only required when Excore Monitoring System is being used to monitor LHR.
1 I
Verify measured ASI has been within 0.05 of Prior to each target ASI for last 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. initial use of i Excore I Monitoring System to monitor LHR SR 3.2.1.4 -------------------NOTE--------------------
Only required when Excore Monitoring System
~
is being used to monitor LHR.
Verify THERMAL POWER is less than the APL. I hour i
i Palisades Nuclear Plant 3.2.1-3 Amendment No. 01/20/98 20-c
LHR B 3.2.1 BASES BACKGROUND Power distribution is a product of multiple parameters, (continued) various combinations of which may produce acceptable power distributions.
The limits on LHR, Assembly Radial Peaking. Factor (F/), i Total Radial Peaking Factor (F/), QUADRANT POWER TILT (T,), i and AXIAL SHAPE INDEX (ASI), which are obtained directly .
from the core reload analysis, ensure compliance with the j safety limits on LHR and Departure from Nucleate Boiling ;
J Ratio (DNBR). l RAL Either of the two core power distribution monitoring systems, the Incore mom 4eHf% System or the Excore-Monitoring System, provides adequate monitoring of the corel. h m power distribution and is capable of verifyino that the LHR5 y is within its limits. The Incore Mem4echg System performs ^ ';
this function by continuously monitoring the local power at many points t)roughout the core and comparing the measurements to predetermined setpoints above which the limit on LHR could be exceeded. The Excore Monitoring l
System perfoms this function by providing comparf son of the measured core .'.SI with predetermined ASI limits based on incore r.easurements. An Excore Monitoring System Allowable Power Level (APL), which may be less than RATED THERMAL POWER, and an additional restriction on T,, are applied when i using the Excore Monitoring System to ensure that the ASI limits adequately restrict the LHR to less than the limiting values.
In conjunction with the use of the Excore Monitoring System for monitoring LHR and in establishing ASI limits, the following assumptions are made:
- a. The control rod insertion limits of LCO 3.1.5,
" Shutdown and Part-Length Rod Group Insertion Limits,"
and LCO 3.1.6, "F.egulating Rod Group Position Limits,"
are satisfied;
- b. The additional T, restriction of SR 3.2.1.6 is satisfied; and
- c. Radial Peaking Factors, F/ and F/, do not exceed the limits of LC6 3.2.2.
Palisades Nuclear Plant- B 3.2.1-2 01/20/98 20-d
_ _ _ _ _._ __ _ - _ _ _ _ _ . _._ . _ _ . ._._ ~ . __ ._ __
k LHR
- B 3.2.1 i BASES t
i BACKGROUND The limitations on the Radial Peaking Factors' provided in .
I (continued) the COLR ensure that the assumptions used in.the analysis !
for establishing the LHR limits and Limiting Safety System i l' t Settings allowable(LSSS) controlremain validinsertion during operation limits. at the various k.tto rod group ?
j The Incore MISGReg-System continuously provides a direct - )(
measure of the LHR and the Radial . Peaking factors. . It also' ;
l '
i provides alarms that have been establisheti for the i individual incore detector segments, ensuring that the peak-l LHRs are maintained within the limits specified in the COLR.-
l The setpoints for these alarms include tolerances, set in conservative directions, for:
- f 3
l a. A measurement calculational uncertainty factor
! (as identified in the COLR);
s
- b. An engineering uncertainty factor of 1.03; and ,
i I c. A THERMAL POWER measurement uncertainty factor of 1.02.
' The measurement uncertainties associated with LHR,_ F/ and .
F/ are based on a statistical analysis performed on power i
distribution benchmarking results. .The COLR Includes the applicable measurement uncertainties for fresh and depleted
} incore detector usage. The engineering and THERMAL POWER p
l uncertainties are incorporated in the power distribution calculation performed by the fuel vendor.
l The excore power distribution monitoring system consists of
- Power Range Channels 5 through 8. The power range channels- ;
j monitor neutron flux from 0 to 125 percent full power. They 1 j are arranged symmetrically around the reactor core to I provide information on the radial and axial flux distributions.
j i I
(
The power range detector assembly consists of two
~
J uncompensated ion chambers for each channel. One' detector extends axially along the lower half of the core while the ;
- other, which is located directly above it, monitors flux j 4
from the upper half of the core. The DC current signal from j each of the ion chambers is fed directly to the control room !
drawer' assembly without pre-amplification. Each excore' l detector supplies data to a Thermal Margin Monitor (TMM). l t Each TMM uses these excore signals to calculate Axial Shape l j Index (ASI) on a continuous basis. l j
i Palisades Nuclear Plant B 3.2.1-3 01/20/98 ,
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to-e.,
.I
LHR B 3.2.1- ,
-BASES i
APPLICABLE c. During an ejected rod accident, the fission energy - 1 SAFETY ANALYSES inputtothefuelmustnotexceed280 cal /gm;and 6,0 (continued) d.
A&
The3contro
% nl rods must be capable of shutting down the-U -;
reactor with a minimum required SDM with the highest i worth control rod stuck fully withdrawn (Ref. 3). j The power density at any point in the core must be limited ,
to maintain the fuel design criteria (Ref. 4). This is ,
accomplished by maintaining the power distribution and !
primary coolant conditions so that the peak LHR and DNB parameters are within operating limits supported by accident analyses (Ref. 1), with due regard for the correlations between measured quantities, the power distribution, and 1 uncertainties in detennining the power distribution.
Fuel cladding failure during a LOCA is limited by ,
restricting the maximum linear heat generation rate so that +
the peak cladding temperature does not exceed 2200*F (Ref.4). High peak cladding temperatures are assumed to cause severe cladding failure by oxidation due to a Zircaloy ' -
water reaction.
The LCOs governing LHR, ASI, and the Primary Coolant System q
Operation ensure that these criteria are met as long as the core is operated within the LHR, ASI, F/, F/, and T, limits. !
i The latter are process variables that characterize the three ;
i dimensional power distribution of the reactor core.
Operation within the limits for these variables ensures that l their actual values are within the ranges used in the l accident analyses. i Fuel cladding damage does not necessarily occur while the !
plant is operating at conditions outside the limits of these l LCOs during normal operation. Fuel cladding damage could 5 result, however, 'if an accident occurs from initial l conditions outside the limits of these LCOs. The potential I l for fuel cladding damage exists because changes in the power po i distribution can cause increased power peaking and can correspondingly increase local LHR.
g i i
The IncoreJ4en44ef4*g System provides for monitoring of LHR, d
[ radial peaking factors, and QUADRANT POWER TILT to ensure ;
that fuel design conditions and safety analysis assumptions NN are maintained. The Incore,Non44e*4eg System is also
- utilized to determine the target AXIAL OFFSET (AO) and to y ;
determine the Allowable Power Level (APL) when using the ;
excore detectors. ;
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I Palisades Nuclear Plant B 3.2.1-5 01/20/98 20-f.
LHR B 3.2.1 BASES APPLICABLE The Excore Monitoring System provides for monitoring of ASI SAFETY ANALYSES and QUADRANT POWER TILT to ensure that fuel design (continued) conditions and safety analysis assumptions are me ntained.
The LHR satisfies Criterion 2 of 10 CFR 50.36(c)(2).
LC0 The power distribution LCO limits are based on correlations l between power peaking and certain measured variables used as inputs to the LHR and DNBR operating limits. The power ,
distribution LC0 limits, except T , are provided in the I COLR. ThelimitationontheLHRlnthepeakpowerfuelrod j at the peak power elevation Z ensures that, in the event of Pg a LOCA, the peak temperature of the fuel cladding does not S exceed 2200*F.
nwntuna wdh n "ad h g g g /The LCO requires that LHR be6-^aite^^f"" ed b"^either4= 0" ",A"LE gmw Incore Hen 4Me4gdystem or - E Excore Monitorino g Systeg. When using the Incore MenHee4+Hf System, the LHR is Lx.oMAmff not considered to be out of limits until there are four or more incore detectors simultaneously in alarm. When using to mona x LM )
the Excore Monitoring System, LHR is considered within limits when the conditions are acceptable for use of the Excore Monitoring System and the associated ASI and T, limits specified in the SRs are met.
To be considered OPERABLE, the [ncore M 1 .; /ystem must A' g have at least 160 of the 215 possible incore detectors
- OVtKAULL anO 2 InCore) Ser axial level per Core Quadrant (
OP LE. For Jhe LHR monitoping (automat 1 al4rming) T unction of thf incore monitfring system t be 4onsideredI T_h addihon, h \
PERABLE./tTe required alann setpointsimu beientered into knt /r0% the/1antgomputer.
1 i
CDenfo{ef mJO To be considered OPERABLE, the Excore Monitoring System must
' ( ggE Ag have been calibrated with OPERABLE incore detectors, the ASI l must not have been out of limits for the last 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, and !
THERMAL POWER must be less than the APL.
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Palisades Nuclear Plant B 3.2.1-6 01/20/98 l 1
20 g
1 LHR B 3.2.1 -
)
BASES :
i APPLICABILITY -In MODE 1 with THERMAL POWER > 25% RTP,' power distribution must be maintained within the limits assumed in the accident analysis to ensure that fuel damage does not result following an A00. In MODE-I with THERMAL POWER s 25% RTP,.
and in other MODES, this LCO does not apply because there is not sufficient THERMAL POWER to require a limit.on the core power distribution, and because ample thermal margin exists to ensure that the fuel integrity is not jeopardized and ,
safety analysis assumptions remain valid.
ACTIONS A.d .
S' There are three acceptable methods 'for verifying that LHR .is [
within limits. The LCO requires monitoring by either an >
fQg,ys OPERABLE Incore&"::it;ria; Syste:n or an OPERABLE Excore y-Monitoring System. When both of the required systems are inoperable, Condition B allows for monitoring by taking i manual readings of the incore detectors. Any of these three methods may indicate that the LHR is not within limits. -
With the LHR exceeding its limit,' excessive fuel damage j could occur following an accident. In this Condition, prompt action must be taken to restore the LHR to within the :
specified limits. One hour to restore the LHR to within its specified-limits is reasonable and ensures that the core does not continue to operate'in this Condition. The 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> Completion Time also allows the operator sufficient time for evaluating core conditions and for initiating proper ;
corrective actions.
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Palisades' Nuclear Plant B 3.2.1-7 01/20/98 l 20-h 4
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LHR )
B 3.2.1 BASES 1
ACTIONS B.1 and B.2 1.- 0 (continued) Al.a.nx.
WiththeIncore."eniteringSysteminoperableformonitoringK LHR and the Excore Monitoring System inoperable for monitoring LHR, THERMAL POWER must be reduced to s 85% RTP within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. Operation at s 85% RTP ensures that ample thermal margin is maintained. A 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> Completion Time is adequate to achieve the required plant condition without challenging plant systems. Additionally, with the Incore Alarm and Excore Monitoring Systems inoperable, LHR must be verified to be within limits within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, and every 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> thereafter by manually collecting incore detector readings at the terminal blocks in the control room utilizing a suitable signal detector. The manual readings shall be taken on a minimum of 10 individual detectors per quadrant (to include a total of 160 detectors in a 10 hour1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> i period) . The time interval of 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and the minimum of 10 detectors per quadrant are sufficient to maintain adequate surveillance of the power distribution to detect significant changes until the monitoring systems are returned to service.
l C.d If the Required Action and associated C mpletion Time are not met, THERMAL POWER must be reduce' co s 25% RTP. This reduced power level ensures that tb core is operating within its thermal limits and plates the core in a conservative condition. The allowed Completion Time of 4
4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> is reasonable, based on operating experience, to reach s 25% RPT from full power MODE 1 conditions in an orderly manner and without challenging pla,'t systems.
i Palisades Nuclear Plant B 3.2.1-8 01/20/98 20-1
-LHR B 3.2.1 BASES S h "
SURVEILLANCE SR 3.2.1.1 REQUIREMENTS I The Incore_."; nit; ring System provides continuous monitoring F of LHR through the plant computer. The plant crmputer is used to generate alarm setpoints that are based en measured margin to allowed LHR. As the incore detectors are read by the plant computer, they are continuously compared to the alarm setpoints. If the Incore " " ti,, ,,9 System LHR y monitoring function is inoperable, excore detectors or brm.-y manual recordings of the incore detector readings may he used to monitor LHR. Periodically monitoring LHR ensures that the assumptions made in the Safety Analysis are maintained. This SR is modified by a Note that states that the SR is only applicable when the Incore ":rit; ring System X is being used to monitor LHR. The 12 hout Frequency is.
consistent with an SR which is to be performed each shift.
?
Continuous monitoring of the LHR is provided by the Incore z q" rit;rin; System which provides adequate monitoring of the core power distribution and is capable of verifying that the LHR does not exceed its specified limits.
m Performance of this SR verifies the Incore#;rit:r'n; System (
can accurately monitor LHR by ensuring the alarm setpoints are based on a measured power distribution. Therefore, they *
, are only applicable when the Incore. "c-ite^g System is
. being used to determine the LHR.
The alarm setpoints must be initially adjusted following each fuel loading prior to operation above 50% RTP, and periodically adjusted every 31 Effective Full Power Days (EFPD) thereafter. A 31 EFPD Frequency is consistent with the historical testing frequeicy of the reactor monitoring ,
system. The SR is modified by a Note which allows the SR to j
- be oerformed only when the Incore#0-iterM; System is being x '
4 used to determine LHR.
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4 Palisades Nuclear Plant B 3.2.1-9 01/20/98 20 ,1 i
1 ATTACHMENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.2.1, LINEAR HEAT RATE ,
a A.4 CTS 3.23.1 provides actions when the LHR is being monitored by the excore yAl monitoring system but the system is no longer appropriate for monitoring LHR as '3 indicated by an Axial Offset (AO) of more than 0.05 (ACTION 2). The actions include both " discontinue using the excore monitoring system for monitoring LHR" and " follow the procedure in ACTION 3 below." - rent in entry into CTS 3.23.1 -
ACTION 2 is that the normally used Incore T Sg is inoperable.
Therefore, this situation is one with both the Incore ?'n _ , System and the excore i :
monitoring system inoperable for the purpose of monitoring LHR. This is includ:d as ITS 3.2.1 Condition B. The specific direction to enter this Condition is not included in ITS since this is the normal use and application of the improved STS format.
Therefore, this omission is considered an administrative change.
A.5 CTS 3.23.1 provides actions when the LHR is indicated as not within the limits specified in the COLR by four or more coincident incore alarms (ACTION 1), and when the manually recorded incore readings indicate a local power level greater than the alarm setpoints (ACTION 3). However, no specific action is provided in the CTS for when the LHR is not w-iiin limits as monitored by the excore monitoring system.
The ITS includes a second entry condition for ITS 3.2.1 Condition A specifically for !
when the LHR is determined to be not within limits using the excore monitoring system, Since the appropriate action is the same regardless of the method used to l utermme that LHR is not within limits, the addition of a specific Required Action, entry condition for "LHR, as determined by the Excore Monitoring System, not within .l limits specified in the COLR" is considered an administrative change. i A.6 CTS 3.23.1 ACTION 3 indicates that when the LHR is indicated as not within the limits specified in the COLR by the manually recorded incore readings "the action specified in ACTION 1 above shall be taken." The ITS includes a third entry
- condition for ITS 3.2.1 Condition A specifically for when the LHR is determined to be i not within limits using the manual incore readings, Since these are only different j j formats to require the same action, the addition of a specific Required Action, entry ;
condition for "LHR, as determined by manual incore readings, not within limits J specified in the COLR" is considered an administrative change.
t 4
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Palisades Nuclear Plant Page 2 of 7 01/20/98 q 20-k
ATTACHMENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.2.1, LINEAR HEAT RATE M.2 CTS does not include specific surveillance requirements to verify that LHR remains %,[.oE within limits. Such an SR is included as ITS SR 3.2.1.1. This SR is necessary to /
provide direct verification that the LCO requirements are met when using the Incore A[g,.gMeniming System for monitoring LHR. Consistent with the NUREG, verification ( ,
~ that an OPERABLE Incore"t i;; System does not indicate LHR out of limits is f sufficient to fulfill this SR. This is an additional restriction on plar,t operation.
LESS RESTRICTIVE CHANGES - REMOVAL OF DETAILS TO LICENSEE CONTROLLED DOCUMENTS (LA)
LA.1 CTS 3.23.1 contains specific details regarding the requirements for monitoring of the LHR, i.e., "in the peak power fuel rod at the peak power elevation Z." This information is not required to be provided in NUREG LCO 3.2.1. These details 1 describe elements of the LHR which are addressed by the methodology for determining ,
LHR and are not directly a part of the actual requirement, i.e., Limiting Condition for Operation. Since these details are not necessary to adequately describe the actual t
regulatory requirement, they can be moved to a licensee controlled document without a significant impact on safety. Placing these details in the LCO Bases of ITS 3.2.1 provides adequate assurance that they will be maintained. The Bases are controlled by j the Bases Control Program in Chapter 5 of the proposed Technical Specifications. This
- change is consistent with NUREG-1432.
! LA.2 CTS 3.23.1 ACTION 3 contains specific details regarding the requirements for monitoring of LHR by manual readings of the incore detection system when the incore !
LHR alarm system is inoperable, i.e., " readings shall be taken on a minimum of 10 individual detectors per quadrant (to include a total ntrmber of 160 detectors in a i
10-hour period)." This information is not provided in NUREG LCO 3.2.1. These
- details describe elements of the incore detection system requirements which are addressed by the methodology for proper use of the system and are not directly a part of the actual requirement, i.e., Limiting Condition for Operation. Since these details '
are not necessary to adequately describe the actual regulatory requirement, they can be moved to a licensee controlled document without a significant impact on safety.
Placing these details in the Bases of ITS 3.2.1 provides adequate assurance that they will be maintained. The Bases are controlled by the Bases Control Program in Chapter 5 of the proposed Technical Specifications. Th's change is consistent with NUREG-1432.
Palisades Nuckar Plant Page 4 of 7 01/20/98 1 20-1
n_. . . . . -. - . . -
[bI G.ncl 4h<, I,ncoN $6 .I#h.m of
@ d.Ce<. Na,no hy%,ro chall k Cf4R%f.,
s n a .a u t.
LHR Rpri1W}
s ., . .
3.2 POWER DISTRIBUTION LIMITS h 3.2.1 Linear Heat Rate (LHR) M h
3,11,1 uo LCO 3.2.1 (be taikk LHR shall mer ar**Fd)the limits specified in the COLR. g
& ,ssan/<ra-< W % ss ove e LC 1 </re A tse,. S*, s tl )
,, wAoyspus (e., + i+ ,i /, sp, .r ,
/j
- 3. Z L I Aptt, APPLICABILITY: MODE 1 ,-
ACTIONS j l
CONDITION REQUIRED ACTION COMPLETION TIME h
3.23 i Quforru4te A. LHR, as determined by A.1 Restore LHR to within I hour i Act I the ncore d3RTEDM limits. X l ra System I h{ng,,4,wperusIDellimitjf1ses4 KI-IcElte fecNielh I h-COLR, as Indicated by ]
four or more .
coincident incore channels.
O d nw LHR,as/determinedby , t.f4E , as de +crained ley b _- i n !" ""*\ I*** '**b $
- hh w;&@ex ori iimi
'a not cai46 limiff ,
Y
,ngg g ,,7 Steeik.'eb k A t 40 des dent ente --
1_ireits hafsoacif' ed in 3,n. l tAs rm.7/I-2A) the l
( . M .3 Q1.E. l 1
frA)SCRTH Required Action and .1 g iyM00 ours 121,1 associated Completion Time not r.at. R*d"'8 M"Al Act 3 PcWER _ o_ 5 25% ETP.
CEOG STS 3.2-1 Rev 1, 04/07/95
, 20-m
SECTION 3.2 ,
INSERT f
B. Incore Alarm and B.1 Reduce THERMAL POWER 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 3.Z3.I Excore Monitoring 'to s 85% RTP.
MrZ Systems inoperable for 4 fc) monitoring LHR. AND 13 witM hadt B.2 NS M+^.-.tx LHRAusing 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> 3'13'I manual incore ACT 3 readings. AN_Q Once per 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> thereafter
't i
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3.2-1 20-n
\
LHR[jAfi
- f f'od 3.Z.1 SURVEILLANCE REQUIREMENTS (continued)
SURVEILLANCE FREQUENCY 0 \
Oz SR 3.2.1.10 ----------------
Atarts s a.[10TEFf------------------
r lOnly applicable)when the Incore pr;n to s /
y,gg,g,g ostsetAr McM tormm System is being
- oPrdim '
1 (usedtodetermineLHR. 7 f0 '7o ETP
..... .... .. . ... ____ . i g m ,-e wen, -c._ 6,n ;
~ 1: n; w x w n y " r 3 s~u"
@cxua% <> a - -wa ?. ,, siinssc-2 4.14,l . 2. b 4.14.1.7.c
SECTION 3.2 INSERT A The Incore Alarm System provides for monitoring of LHR, radial peaking factors, and QUADRANT POWER TILT to ensure that fuel design conditions and safety analysis assumptions are maintained. The Incore Alarm System is also utilized to determine the target AXIAL OFFSET (AO) and to determine the Allowable Power Level (APL) when using the excore detectors.
The Excore Monitoring System provides for monitorinF of ASI and QUADRANT POWER TILT to ensure that fuel design conditions and safety analysis assumptions are maintained. 8A(
S INSERT B rei. hine) toi4h.n A.
8m,4J /Ar.iC d an N [.Ol.A W 46, g The LCO requires that LHR]be -9x:d b/bither.anCBERABbE Incore Alarm System o an OPEPf3LE Excore Monitoring Systent When using the Incore Alarm System, the LHR X l
h is'not considered to be out of limits until there are four or more incore detectors simultaneously in alarm. When using the Excore Monitoring System LHR is considered egs within limits when the conditions are acceptable for use of the Excore Monitoring System and the associated ASI and T, limits specified in the SRs are met.
To be considered OPERABLE, theAlcore must have at least 160 of the 215 Y possible incore detectors OPERABLE and 2 incor 1 level per core quadrant OPERABLE.fFor theAHR monitorty_(automatic alarmmpI function of the incore jdonitoring) K rsystem 16 be consido(ed OPERABLE Ahe required alarm setpomts must be entered into the pant computer.
To be considered OPERABLE, the Excore Monitoring System must have been calibrated with OPERABLE incore detectors, the ASI must not have been out of limits for the last 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, and THERMAL POWER must be less than the APL.
T t\ 0.da dton, ths fkink fcgg.,5s C. M d < m an + k O('fgA3\(
a.n ci i
i 20-p B 3.2-4
M
! SECTION 3.2 I
INSERT A
! B.1 and B.2.
J l With the Incore Alarm System inoperable for monitoring LHR and the Excore Monitoring i System inoperable for monitoring LHR, THERMAL POWER must be reduced to s 85% RTP l within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. Operation at s 85% RTP ensures that ample thermal margin is maintained. A 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> Completion Time is adequate to achieve the required unit condition without challenginF
- plant systems. Additionally, with the Incore Alarm and Excore Monitoring Systems inoperable, LHR must be verified to be within limits within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, and every 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> j thereafter by manually collecting incore detector readings at the terminal blocks in the control
- room utilizing a suitable signal detector. The manual readings shall be taken on a minimum of
- 10 individual detectors per quadrant (to include a total of 160 detectors in a 10 hour1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> period).
[ The time interval of 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and the minimum of 10 detectors per quadrant are sufficient to maintain adequate surveillance of the power distribution to detect significant changes until the l
1 monitoring systems are returned to service. '
o l INSERT B 8(
Alarm
- The Incore M+ig System provides continuous monitoring of LHR through the plant x
. computer. The plant computer is used to generate alarm setpoints that are based on measured l margin to allowed LHR. As the incore detectors are read by the plant computer, they are
- continuously compared to the alarm setpoints. If the Incore Alarm System LHR monitoring l function is inoperable, excore detectors or manual recordings of the incore detector readings i may be used to monitor LHR. Periodically monitoring LHR ensures that the assumptions i made in the Safety Analysis are maintained. This SR is modified by a Note that states that the :
SR is only applicable when the Incore Alarm System is being used to monitor LHR. J l
i l
4 t
20-q B 3.2-5 1
! 1
ATTACIBIENT 6 JUSTIFICATION FOR DEVIATIONS SPECIFICATION 3.2.1, LINEAR HEAT RATE (LHR)
Chance Dkcuuinn Note: This attachment provides a brief discussion of the deviations from NUREG-1432 that were made to support the development of the Palisades Nuclear Plant ITS. The Change Numbers correspond to the respective deviation shown on the "NUREG MARKUPS." The first five justifications were used generically throughout the markup of the NUREG. Not all generic justifications are used in each specification.
- 1. The brackets have been removed and the proper plant specific information or value has been provided.
- 2. Deviations have been made for clarity, grammatical preference, or to establish consistency within the Improved Technical Specifications. These deviations are .
editorial in nature and do not involve technical changes or changes of intent. j
- 3. The requirement / statement has been deleted since it is not applicable to this facility. The following requirements have been renumbered, where applicable, to reflect this deletion.
l
- 4. Changes have been made (additions, deletions, and/or changes to the NUREG) to reflect the facility specific nomenclature, number, reference, system description, or analysis description.
- 5. This change reflects the current licensing basis / technical specification. These O
include an ITS 3.2.1 Applicability less restrictive than the NUREG and the f j
addition of an ACTgr determination of LHR using manual readings when '
both the Incore Mc. -. .. System and the excore monitoring system are )( I inoperable for determining LHR. With power reduced to below 85% RTP (per ITS 3.2.1, Required Action B.1), the manual readings of the incore l monitors provide an adequate indication that LHR is within limits. This is l consistent with CTS as approved in Amendment 68. Additionally, the proposed l 4
Applicability for ITS 3.2.1 is actually more restrictive than CTS 3.23.1 which
> 25% RTP" is consistent with the Applicability for the other Power Distribution Limit specifications, and provides for incore adjustments based on power distribution maps prior to exceeding 25% which is consistent with i Quadrant Power Tilt needs for incore adjustments.
Palisades Nuclear Plant Page 1 of 5 01/20/98 i 1
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l ATTACHMENT 6 l JUSTIFICATION FOR DEVIATIONS SPECIFICATION 3.2.I. LINEAR HEAT RATE (LHR)
Chance Discueelan .
- 8. An addition to the LCO in incorporated which equires that the LHR be X f j determined by an OPERABLE Incore Me9 .g System or by an OPERABLE l
- excore monitoring system. Such an LCO requirement is consistent with the j
. NUREG SR Note which requires that the LHR be determined by either the incore detector monitoring system or the excore detector monitoring system.
{
However, incorporating the requirement into the LCO provides a more direct j indication that the LCO is not met when both the incore LHR alarm function and the excore LHR monitoring function are inoperable (which results in entry into ITS Condition B, as discussed in JFD 5).
) 9. The Surveillance Requirements (SRs) for LHR are revised consistent with the I current licensing basis. The NUREG SR Note is inappropriate for Palisades
', Nuclear Plant because manual reading of the incore monitors is also allowed for j determining LHR to be within limits. This is corrected by incorporating the SR j Note requirements directly into the LCO (see JFD 8) and adding an ACTION j for use of the manual incore readings (see JFDs 5 and 7). The NUREG SRs are also inappropriate for all plants since failure of the alarms or serpoints to be properly set does not mean that the LHR is not within limits. However, SR 3.0.1 would require that the LCO be considered not met when any of these
- SRs are not met . This is not consistent with the format and content iment of i the improved STS NUREGs, is considered overly conservative, and is not adopted.
l
! ITS SR 3.2.1.1 specifically requires the verification that LHR is within the i limits specified in the COLR. This SR is a direct verification that the LCO is l being met (which is missing from the NUREG). Mcwever, since the LHR is i normally automatically monitored and alarmed by the incore power distribution
- monitoring system, the SR is only required to be performed when the Incore l M Mc9-i;; System is being used to determme LHR, and is met by )(
administrative verification that thejicore raeMMGgfystem is OPERABLE for N monitoring LHR, and that the Gicore
""~
fystem does not indicate LHR q is not within limits.
i Palisades Nuclear Plant Page 3 of 5 01/20/98 20-s
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i A'ITACIDENT 6 -
i JUSTIFICATION FOR DEVIATIONS
- SPECIFICATION 3.2.1, LINEAR HEAT RATE (LHR) i i
Chnnoe Die = clan
- 9. (continued)
! NUREG SR 3.2.1.2 and SR 3.2.1.3 requirements for incore alarms are
- combined and revised to reflect CTS 4.19.1. ITS SR 3.2.1.2 requires that the
! incore alarm setpoints be adjusted (i.e., the alarms be set) based on a measured l power dist ibution. This Surveillance provides adequate assurance that the Incore# % System is providing accurate monitoring of the LHR. This Y change is consistent with CTS 4.19.1 requirements for adjustments of incore alarm settings.
ITS SR 3.2.1.3, SR 3.2.1.4, SR 3.2.1.5, and SR 3.2.1.6 require the verification
- of parameters that similarly indicate the LHR is within the limits specified in the COLR when using the excore monitoring system. These SRs also provide verification that the parameters are appropriate for use of the excore monitoring system to monitor LHR and that the LCO is being met (which is missing from the NUREG). However, since the LHR is normally automatically monitored and alarmed by the[ core dd5 p) 4 stem, these SRs are only required to be M met when the excore monitoring system is being used to determine LHR. These i SRs are generally consistent with the requirements of CTS 4.19.1.2a, b, c, aM d.
2
- 10. The periodic Frequency of NUREG SR 3.2.1.3 is revised to 31 EFPD.
. CTS 4.19.1.1 provides requirements to adjust the incore alarm settings based on a measured power distribution on a periodic Frequency of "7 days of power operation." Although the CTS Frequency is based on days of power operation, this is inconsistent with the Frequency of ITS Section 3.1 SRs which are based on EFPD, inconsistent with' NUREGs for other vendors (e.g., NUREG-1430 i and NUREG-1431) for Power Distribution Limit SRs which are based on l EFPD, and inconsistent with preferred methods for tracking this Frequency since EFPD is already required to be tracked to for numerous calculations 4 related to burnup and other fuel status para neters. When the plant is operating steadily at full power there is no difference in the NUREG SR 3.2.1.3 periodic Frequency of "31 days" and the proposed "3r ECPD." However, when the.
31 days includes operation at less than full power ine "31 EFPD" is longer than-2 the NUREG would allow. Still, the revision to the SR Frequency is acceptable since the Frequency continues to be sufficient to assure the incore alarm settings are appropriately since any change is a slow process.
Palisades Nuclear Plant Page 4 of 5 01/20/98
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CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO DECEMBER 09, 1998 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.2. POWER DISTRIBUTION LIMITS NRC REQUEST:
3.2-02 ITS 3.2.1 Linear Heat Rate (LHR)
ITS 3.2.1 Surveillance Requirements JFD 9 The STS SRs have been changed in the ITS to be consistent with the CTS.
Coment #1: The ITS SR 3.2.1.1 Note incorrectly references LC0 3.2.5 and LCO 3.2.6. What is the purpose of this note? Recommend deleting note.
Coment #2: Provide ITS SR 3.2.1.1 an appropriate specific frequency.
Consumers Enerav Resconse:
The markup of ISTS SR 3.2.1.1 (Attachment 5 NUREG page 3.2-2 Insert) contains a Note which inappropriately references LC0 3.2.5 and LC0 3.2.6. The markup also inappropriately specified a Frequency of "as required by applicable specification.". The Note was intended to state "Only Required when the Incore Monitoring System is being used to monitor LHR" and to specify a Frequency of "12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />." The intended version of this Note was correctly presented in the clean typed copy of Specification 3.2.1 and its associated Bases found in Attachments 1 and 2, respectively, and appropriately justified in JFD 9 (Attachment 6).
- 8ffected Submittal Paaes Att 5 NUREG, pg 3.2-2 insert 21
SECTION 3.2 a
INSERT SURVEILLANCE REQUIREMENTS
, SURVEILLANCE FREQUENCY 4
i SR 3.2.1.1 --NOTE -
I
'Only r ired to be performed hen specified y 00!
- LCO 3 .5, "Incore Monitor' System," or Lcct -3',2.-O L '
3.2.6 Excore Monitoring stem."
-t i
Verify LHR is within the limits specified in the COLR. I& bovG i
lj Q dIftd Mbth bM 5Idb @ p ggggg S iri = l a Wng ura.A b monder LHR, l
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CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO DECEMBER 09, 1998 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.2, POWER DISTRIBUTION LIMITS NRC RE00EST:
3.2-02 ITS 3.2.1 Linear Heat Rate (LHR)
ITS 3.2.1 Surveillance Requirements JFD 9 The STS SRs have been changed in the ITS to be consistent with the CTS.
Coment #3: ITS SR 3.2.1.3 and ITS SR 3.2.1.5 should appear with the ASI specification; recomend moving to ITS 3.2.4. (4) ITS SR 3.2.1.6 should appear with the T, specification; recommend moving to ITS 3.2.3.
Coment #4: ITS SR 3.2.1.6 should appear with the T, specification; recommend moving to ITS 3.2.3.
Consumers Enerav Resoonse:
ITS SR 3.2.1.3, SR 3.2.1.5, and SR 3.2.1.6 ensure the conditions related to l
. core power distribution are acceptable before using the Excore Monitoring System to monitor LHR, and to ensure LHR remains within limit. The Excore Monitoring System does not determine LHR directly. However, if more ;
restrictive limits are placed on both ASI and T,, Excore readings may be used l
- to assure LHR is within limits. These more restrictive limits are only !
- necessary when the Incore Alarm System is unavailable. The limits imposed by 1 these SRs are more restrictive than the limits imposed in their respective specifications (i .e., ITS 3.2.4, " Axial Shape Index", and ITS 3.2.3, " Quadrant PowerTilt"). Since failure to meet an SR would be failure to meet the LC0 (SR 3.0.1), placing the more restrictive SRs in their respective specifications would invoke inappropriate Required Actions in the event an SR has failed. Therefore, to ensure the appropriate Required Actions are taken when LHR is not within limits as determined by the Excore Core Monitoring, SR 3.2.1.3, SR 3.2.1.5, and SR 3.2.1.6 must be retained in ITS 3.2.1.
I i
Affected Submittal Poaes j None 1
22
CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO DECEM8ER 09, 1998 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.2, POWER DISTRIBUTION LIMITS N8C REQUESE:
3.2-03 ITS 3.2.1 Linear Heat Rate (LHR)
ITS SR 3.2.1.2 Frequency JFD 10 The frequency for ITS SR 3.2.1.2 has been changed, from 7 days in the CTS and 31 days in the STS, to 31 EFPD; a beyond scope change.
Comeent: Recommend retaining the STS frequency of 31 days for ITS SR 3.2.1.2.
- Consumers Enerav Resoonse
The Frequency of proposed ITS SR 3.2.1.2 was changed from units of " days" to "EFPD" (Effective Full Power Days) to be consistent with proposed SR 3.2.2.1 (Specification 3.2.1, DOC L.4). Aligning the Frequency of these two SRs is logical since the input to SR 3.2.1.2 is based on the results of SR 3.2.2.1.
As noted in NRC Request 3.2-05, the Frequency of SR 3.2.2.1 was changed from units of " days of accumulated operation in Mode 1" to "EFPD" (Specification 3.2.2, DOC L.2). These changes were made to establish
- consistency with the methods generally accepted to track core parameters that are sensitive to fuel burnup. These methods are deemed acceptable since power distribution changes are relatively slow over a 31 day period. In addition, nearly all " power operation" is at the full power condition, and when the i
plant is operating at full power there is no difference in a Frequency of 4
31 days and 31 EFPD.
- Although these Frequency changes represent a deviation from NUREG-1432, they are consistent with similar type power distribution Frequencies in NUREG-1430
- (B&W plants) and NUREG-1431 (Westinghouse plants) previously found acceptable by the NRC. As such, Palisades would like to retain the Frequency units of EFPD in SR 3.2.1.2 and SR 3.2.2.1 on the basis it is consistent with the Improved Standard Technical Specifications for power distribution related surveillances.
Affected Subetttal Paats None 23 P
CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO DECEMBER 09, 1998 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.2, POWER DISTRIBUTION LIMITS NRC REQUEST:
3.2-04 ITS 3.2.2 Radial Peaking Factors ITS 3.2.2 Required Action B DOC M.1, JFD 5 and JFD 8 The CTS requires going to Hot Shutdown (similar to Mode 3) in 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> if peaking factors are not within limits, with Power < 50% RTP. The ITS requires going to 5 25% RTP in 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> if peaking factors are not returned to within limits in 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
Coment H: In the STS, when a radial peaking factor is not within limit the first action is to reduce power. The ITS allows 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> delay' prior to reducing power. The ITS should more closely reflect the STS actions.
Consumers Enerav RtSDonst:
When radial peaking factors are not within limit, the Required Actions of both the ISTS and ITS allow 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> to establish compliance with the LCO. The Required Actions of the ISTS are more prescriptive than the ITS since they i include the method for restoring compliance with the LCO. Neither the CTS, nor the ITS provide this same level of detail but simply require the peaking factors be restored to within limits without specifying the method used to accomplish the restoration. Although restoration would typically include a reduction in thermal power, such a reduction may not always be necessary.
Alternatively, correcting the source of the peaking may be the optimum method for restoration. The method to restore peaking factors prescribed in the ISTS is by reducing thermal power while withdrawing the CEAs to or beyond their long term steady state insertion limit. Since only one set of insertion limits is used at Palisades, actions-to be taken if rods are inserted beyond the insertion limits are specified in LCO 3.1.5 or 3.1.6. These actions have a 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> Completion Time. Therefore, a prescriptive Required Action to reduce thermal power may not always be appropriate. As such, Palisades would like to maintain the operational flexibility that exists in the CTS for restoring radial peaking factors to within limits.
Affected Submittal Paaes None 24
1 CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO DECEMBER 09, 1998 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.2, POWER DISTRIBUTION LIMITS NRC RE00EST:
3.2-04 ITS 3.2.2 Radial Peaking Factors ITS 3.2.2 Required Action B DOC M.1, JFD 5 and JFD 8 The CTS requires going to Hot Shutdown (similar to Mode 3) in 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> if peaking factors are not within limits, with Power < 50% RTP. The ITS requires going to 5 25% RTP in 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> if peaking factors are not returned to within limits in 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
Cauwent #2: This change to the CTS is less restrictive and needs to be appropriately justified.
Consumers Enerav Response:
DOC L.1 has been revised to enhance the justification which provides four additional hours to exit the mode of applicability when radial peaking factors I can not be restored within limits. .
1 Affected Submittal Poaes Att 3 DOC 3.2.2, pg 3 of 4 Att 4 NSHC 3.2.2, pg 1 of 5
'f 9
25
j i ATTACHMENT 3 J DISCUSSION OF CHANGES j SPECIFICATION 3.2.2, RADIAL PEAKING FACTORS l LA.2 CTS 4.19.2.1 provides Surveillance Requirements (SRs) for the Radial Peaking )
Factors. However, it contains specific details for monitoring of the peaking factors, l i
! i.e., that the SR is performed by verifying the " measured" radial peaking factors i "obtained by using the incore detection system." This information is not provided in NUREG SR 3.2.2.1. These details describe elements of the radial peaking factor verification which are addressed by the methodology and are not directly a part of the j actual requirement, i.e., Surveillance Requirement. Since these details are not ;
i necessary to adequately describe the actual regulatory requirement, they can be moved l to a licensee controlled doc 6 ment without a significant impact on safety. Placing these (
l j details in the Bases of ITS SR 3.2.2.1 provides adequate assurance that they will be maintained. The Bases are controlled by the Bases Control Program in Chapter 5 of the proposed Technical Specifications. This change is consistent with NUREG-1432.
. LESS RESTRICTIVE CHANGES (L)
(
l L.1 ' 6 3.23.2 pwv-s actions for peaking ractors exceeding their limits based on power) ,
lev The first of these actions is for P (power) < 50%, and requires the plant to be I tin at leas t shutdown, i.e., subcritical, within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. ITS 3.2.2 Required Action
}
1 A.1 provides rs to attempt restoration of the peaking factors to within limits, and if the Required Acti?> its associated Completion Time is not met, then Required Action B.1 requires that AL POWER be reduced to s 25% RTP. This change j is less restrictive in two ways, t, six hours is provided to attempt restoration of the
- peaking factors to within limits that is t provided in the CTS. Second, the default action requires only that the plant to be r to s 25% RTP, rather than subcritical.
The ITS Required Actions are appropriate for the co ' ions and assure the plant will l not operate for an extended period with the peaking facto t within limits. The Completion Time provides a reasonable time for determining roper method, power !
! - level, and associated limits for restoration, and for the restoration the plant to within itions in an
] g} dA limits, and a reasonable time to remove the plant from the applicable c orderly manner and without challenging plant systems. This change is co ' stent with
{UREG-1432 as modified for plant specific parameters and analysis, j l
2 /
$1. 3.P0Y ,
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. Palisades Nuclear Plant Page 3 of 4 01/20/98 25-a
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. INSERT CTS 3.23.2 provides actions for peaking factors exceeding their limits based on power level.
The first of these actions is for P (power) < 50%, and requires the plant to be in at least hot
{ shutdown (i.e., suberitical) within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. ITS 3.2.2 Required Action A.1 provides 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />
, to attempt restoration of the peaking factors to within limits, and if the Required Action'and its associated Completion Time is not met, then Required Action B.1 requires that THERMAL POWER be reduced to s 25% RTP. This change is less restrictive in two . ways. First,
, 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> is provided to attempt restoration of the peaking factors to within limits that is not i provided in the CTS. Second, the default action requires only that the plant to be reduced to s 25% RTP, rather than subcritical, in the subsequent 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
]
4
- The ITS Required Action to restore the radial peaking factors to the within limits specified in i the COLR assure the plant will not operate for an extended period with the peaking factors not I within limits. The Completion Time of 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> provides a reasonable time for evaluating core
! ; conditions, calculating a reduced power level at which the peaking factors would be within ;
limits, determining the proper method for the power reduction (e.g., rod positioning and/or i boration) and, completing the reduction in power. In the event the peaking factors are not i
restored to within limits, an additional 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> is provided to remove the plant from the mode of j applicability. Although CTS 3.23.2 requires the plant to be placed in hot shutdown, terminating -
i the power reduction anywhere below 25% is permissible since CTS LCO 3.0.1 only requires
- co:npliance with an LCO during the plant condition specified in that LCO. Thus, the default r.ction of proposed ITS Required Action B.1 is consistent with the shutdown action for CTS 3.23.2. A Completion Time of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> is reasonable to reduce thermal power below 25% in i
an orderly manner and without challenging plant systems.
1 I
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ATTACHMENT 4 NO SIGNIFICANT HAZARDS CONSIDERATION SPECIFICATION 3.2.2, RADIAL PEAKING LESS RESTRICTIVE CHANGE L.1 -
~ .-
^
CTS 3.23.2 provides actions for peaking factors exceeding their limits based on power level.1 The first of these actions is for P (power) < 50%, and requires the plant to be in at lea ot I shutdown, i.e., subcritical, within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. ITS 3.2.2 Required Action A.1 prov' 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> to attempt restoration of the peaking factors to within limits, and if the Re Action and its associated Completion Time is not met, then Required Action B. tres that THERMAL !
POWER be reduced to s 25% RTP. This change is less active in two ways. First, j six hours is provided to attempt restoration of g factors to within limits that is not i provided in the CTS. Second, the default on requires only that the plant to be reduced to k s 25% RTP, rather than subcritica 5
\
The ITS Required Act' are appropriate for the conditions and assure the plant will not operate for an e ed period with the peaking factors not within limits. The Completion Times prov' s a reasonable time for determining the proper method, power level, and associat limits for restoration, and for the restoration of the plant to within limits, and a reas le time to remove the plant from the applicable conditions in an orderly manner and wi tout challenging plant systems. This change is consistent with NUREG-1432 as modified for plant specific parameters and analysis.
p iM gA1S.P 3
Palisades Nuclear Plant Page 1 of 5 01/20/98 l
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l
INSERT CTS 3.23.2 provides actions for peaking factors exceeding their limits based on power level.
The first of these actions is for P (power) < 50%, and requires the plant to be'in at least hot shutdown (i.e., subcritical) cithin 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. ITS 3.2.2 Required Action A.1 provides 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> to attempt restoration of the peaking factors to within limits, and if the Required Action and its associated Completion Time is not met, then Required Action B.1 requires that THERMAL POWER be reduced to s 25% RTP. .This change is less restrictive in two ways. First, 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> is provided to attempt restoration of the peaking factors to within li: nits that is not i provided in the CTS. Second, the default action requires only that the plant to be reduced to s 25% RTP, rather than subcritical, in the subsequent 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
The ITS Required Acticn to restore the radial peaking factors to the within limits specified in '
the COLR assure the plant will not operate for an extended period with the peaking factors not within limits. The Completion Time of 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> provides a reasonable time for evaluating core conditions, calculating a reduced power level at which the peaking factors would be within limits, determining the proper method for the power reduction (e.g., rod positioning and/or .
boration) and, completing the reduction in power. In the event the peaking factors are not restored to within limits, an additional 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> is provided to remove the plant from the mode of applicability. Although CTS 3.23.2 requires the plant to be placed in hot shutdown, terminating the power reduction anywhere below 25% is permissible since CTS LCO 3.0.1 only requires compliance with an LCO during the plant condition specified in that LCO. Thus, the default
, action of proposed ITS Required Action B.1 is consistent with the shutdown action for CTS 3.23.2. A Completion Time of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> is reasonable to reduce thermal power below 25% in an orderly manner and without challenging plant systems,
{
i J
4 25-d
CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO DECEMBER 09, 1998 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.2, POWER DISTRIBUTION LIMITS NRC RE0UEST:
3.2-05 ITS 3.2.2 Radial Peaking Factors ITS SR 3.2.2.1 Frequency DOC L.2 and JF0 9 The frequency for ITS SR 3.2.2.1 has been changed, from 7 days in the CTS and 31 days in the STS, to 31 EFPD; a beyond scope change.
Convent: Recommend retaining the STS frequency of 31 days for ITS SR 3.2.2.1.
Consumers Enerav ResD0nse:
1 Please see the response to NRC Request 3.2-03.
Affected Submittal Poaes None 26
7-CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RES?0NSE TO DECEMBER 09, 1998 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.2, POWER DISTRIBUTION LIMITS NRC REQUEST:
3.2-06 ITS 3.E.3 [$TS 3.2.4] Power Tilt (T,)
ITS 3.2.3 [STS 3.2.4] LC0 and Required Actions JFD 1 and JFD 5 STS 3.2.4 has been rewritten to reflect CTS limits in ITS 3.2.3. r Comment #1: The ITS has not retained STS Required Action C.3 to restore T, to
< [0.03] prior to increasing thermal power 'if T, is no longer >[0.10]);
submit TSTF for chance to STS. NRC to review.
Comment #2: The ITS has not retained the STS Notes to the Required Action C and the related Completion Times, though similar requirements are retained in administrative controls; submit TSTF for change to STS. NRC to review.
Consumers Enerav Resoonse:
Palisades will propose a generic change to NUREG-1432 at the next meeting of ,
the CE Owners Group Licensing Subcommittee to delete ISTS 3.2.4, Required Action C.3 and its associated Completion Time.
&L,fected Sybetttal Paaes ;
None 27 I
I 1
I
-. . - _ _ _ - - - - _ _ _ _ _ _ _ _ _ _ _ _ _ _l
CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE % 'ECEMBER 09, 1998 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.2, POWER DISTRIBUTION LIMITS NRC RE00EST:
3.2-07 ITS 3.2.3 [STS 3.2.4] Power Tilt (T,)
ITS 3.2.3 [STS 3.2.4] Bases to Required Actions (STS pages B 3.2-23 & B 3.2-24)
- JF0 5 The STS Bases contains two paragraphs addressing STS Required Actions C.1, C.2 und C.3, that have been deleted in the ITS Bases.
Coment: Recommend retaining this information tailored for Palisades.
1 CDBsumers Enerav Resconse:
- The deleted paragraphs were reviewed for information that is appropriate for inclusion in the ITS Bases. While much of the deleted information deals with q
the omitted Actions C.1, C.2, and C.3 the appropriate information has been added to the Bases for ITS Action B.1.
l Affected Submittal Poaes :
Att 2, ITS 3.2.3, page B 3.2.3-2 ,
Att 5, NUREG 3.2.4, page B 3.2-23 I
l i
28
,. T, B 3.2.3 BASES-ACTIONS M If the measured T is > 0.05, T must-be restored within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or F,' and, F,' must be determined to be within the' ,
limits of LCO 3.2.2, and determined to be within these J limits e hours thereafter, as long as T is out of
'[wo Cmits. hours is sufficient time to all,wo the operator CU -
to repor 1on control rods, and significant radial xtnon redistribution cannot occur within this time. The 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> Completion Time ensures changes in F,' and F,' can be identified before the limits of LCO 3.2.2 are exceeded. 1 U
.With the measured T > 0.10, power - : be reduced to h -
l
< 50% RTP within 4 bours, and F," and F,' must be within g' their specified limits to ensure that acceptable flux ,
peaking factors are maintained as required by Condition A
~
(which continues to be applicable). 4as d on opera W 7 EKp;riO.9c0, 4-hvurs is suffisicut time for evaiustivn vi
-tttess. fs;;;rs .
If F,' and F,' are within limits, operation y j ggf y :
may proceed while attempts are made to restore T, to within its limit.
u If T is > 0.15, or if Required- Actions and associated CompletionTimesarenotmet,THERMALPOWERmustbereduced to s 25% RTP. This requirement ensures that the core is operating within its thennal limits and places the core in a
- conservative condition. Four hours is a reasonable time to ,
reach 25% RTP in an orderly manner and without challenging plant systems. ;
If the tilt is generated due to a control rod misalignment. continued operation at < 50% RTP allows for realignment; if the cause is other than control rod misalignment. continued operation may be necessary to i
discover the cause of the tilt. Reducing THERMAL POWER to < 50% RTP. and i the more frequent measurement of peaking factors required by Action A.1
- provide conservative protection from potential increased peaking due to xenon redistribution.
6 Palisades Nuclear. Plant B 3.2.3-2 01/20/98 28-a
N.IIIiU'li'iEl';U'Isi I,!' lUIU!.E'?O'NIUlli II"U'efim T, @
n;;;S'#,'u:"'0'0; itu'.'U.llia'lH2 EU'?" 'h!'m. ,,.
Dg:
BASES (continued) .=.. ,,,.=:i : :=,rmuinr=,::.un:a
- i. ,i.. ...
)
ACTIONS A.1 [ame E2) _
1 6 A If the measured T. is > (LF;0F and <A.14.mdr01stion; l
7
<owserzner N a) T, must be restored within '
2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> o and F; must be determined to be within the limits of L 3.2.2 (EiFCFTES. and deterained to be
@h {60 **g **Aa 5' within t_hese limits every 8 hos.s thereafter, as lon as T, is out of limits. Cd hours is sufficient time to a low the 1 operator to repostt on u gy, and significant radial xenor> i redistribution cannot occur within t s ting. The 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 1 Completion Time ensures changes in nd F, can be identified before the limits of L .2.2 tanruYMJh.
(FRRprtiltT]Pl) are exceeded.
Q is
- Y O.1[i *or i_
_ ~9 If Reautred Actions.and associated Completion Times @ l r-ntMn m are not met, THERMAL POWER must be reduced to 1 5 RTP, This requirement ensures that the core is l operating within its thermal limits and places the core in a conservative condition. Four hours is a reasonable time to reach @ RTP in an orderly _ manner and without challenging j
"" *#k'***l po[er mst be redued to < SD% f.TP b55 60 wih 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, and F[
=
(c x r.a . A C @ - - _ . - - _
g Q 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />) q f j b8
- _
WithYT, > 0.10 and FJ must be within their specified limits to snsure hat. acceptable . flux peaki factors are maintained' Based qn operating experience, is V M *.G(1 su tent too for ttronerate tenvuuate t ese factors.
and F are within limits, operation may proceed 6 pAr if
- s c - urs m ar m a wu- u m urmr1 whil e
/ attempts are made to restore T, to within its limit.( "
l' eftl $ 0 AEnnot he a eved, power.augbereducedto s 505 within 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> . If the tilt gene' rated Bue to CEA salignment, op ating at s 50% allows for the I i j , re very of the~C , Except as a re it of CEA i ; salignment. T. 0.10 is not exp ed; if it occurs
- continued oper ion of the react may be necessary o l
- discover the ause of the tilt. If this procedure s
- followed, eration is restri ed to only those nditions required identify the ca e of the tilt. It s necessary I
- to acc t explicitly for, ower asymmetries ause the J j (continued)
CEOG STS B 3.2-23 Rev 1, 04/07/95 g gn.w s.s - c1 C m., i,. v.q $ ) @.ti c cl c oc'O M l
\ To St. h%CW8
\
l 28-b
.e.
CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO DECEM8ER 09,1998 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.2, POWER DISTRIBUTION LIMITS NRC REQUEST:
3.2-08 ITS 3.2.3 [STS 3.2.4] Power Tilt (T,)
ITS 3.2.3 [STS 3.2.4] Required Actions DOC L.1 and DOC M.3 .
The CTS required action if T, is > 0.15 is to go to Hot Shutdown in 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> while the ITS required action is to decrease power to s 25% RTP in 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
Comment: Confirm that the ITS Applicability is appropriate and that the CTS required action is overly restrictive. When T, becomes too large shutting down may be the appropriate action.
Consumers Enerav Resoonse:
The Applicability for the ITS T, LC0 is unchanged from CTS. The Applicability for T, in both the CTS and ITS is > 25% Rated Power. Although CTS 3.23.3 requires the plant be placed in Hot Standby whenever T, is > 0.15, terminating the power reduction anywhere below 25% is permissible since CTS LC0 3.0.1 only requires compliance with an LC0 during the plant conditions specified in that LCO. At power levels < 25% there is insufficient Thermal Power to require a limit on core power distribution. In addition, ample thermal margin exists to ensure fuel integrity is not jeopardized and safety analysis assumptions remain valid. As such, requiring the plant to be placed in hot standby when T, is not within limit is overly restrictive.
Affected Submittal Paaes None 29 I 1
j
CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO DECEMBER 09, 1998 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.2 POWER DISTRIBUTION LIMITS NRC REQUEST:
3.2-09 ITS 3.2.3 [STS 3.2.4] Power Tilt (T,)
ITS SR 3.2.1.6 CTS SR 4.19.1.1.b Proposed ITS SR 3.2.1.6 (CTS SR 4.19.1.1.b) imposes a limit on T, of 0.03.
Comment: This limit does not appear anywhere in any ITS LC0 limit; why not?
Should the STS limits be adopted?
Consumers Enerav Resconse:
The 3% limit for T, does not appear in any ITS LC0 since it is specified as a surveillance requirement associated with the LCO for LHR. This surveillance requirement ensures the conditions related to core power distribution are acceptable before using the Excore Monitoring System to monitor LHR, and to ensure LHR remains within limit.
Adopting the ISTS limits (i.e., specifying the 3% limit on T, in the Quadrant Power Tilt specification) would not be appropriate since the 3% restriction on T, is only applicable when the Excore Monitoring System is being used to monitor LHR. If T, were to exceed the 3% limit when the Excore Monitoring System is being used to monitor LHR, the appropriate Required Actions would be those actions associated with the LHR LCO, not the T, LCO.
Also see response to RAI 3.2-02. i Affected Submittal Paaes None a
30
____--___-___L
CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO DECEMBER 09, 1998 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.2. POWER DISTRIBUTION LIMITS NRC RE00EST:
3.2-10 ITS 3.2.3 [STS 3.2.4] Power Tilt (T,)
ITS 3.2.3 Required Action C i DOC M.2 ITS Condition C (default actions if required action not met) is an addition to CTS actions.
Comment: Wouldn't the CTS actions implicitly required a similar action?
How is this more restrictive; is this an administrative change?
, Cpnsumers Enerav Resoonse:
If the Actions of CTS 3.23.3 could not be met, then LC0 3.0.3 would require the plant to be placed in hot standby within 7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br />. The "more restrictive" aspect of adding ITS Condition C is the shorter time for completing the shutdown (i.e., 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> in ITS versus 7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br /> in CTS).
< i Affected Submittal Paaes None i
l j
i 31
l l
I l ENCLOSURE 2 i
- CONSUMERS ENERGY COMPANY
. PALISADES PLANT i DOCKET 50-255 4
i l
CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS
! RESPONSE TO DECEMBER 4, 1998 REQUEST FOR ADDITIONAL INFORMATION l
i i EDITORIAL CHANGES A
Control Rod Alignment :
3.1.4 3.1 REACTIVITY CONTROL SYSTEMS 3.1.4 Control Rod Alignment pg LCO 3.1.4 All control rods, including their position indication Y channels, shall be OPERABLE and aligned to within 8 inches of all other rods in their respective group, and the control rod position deviation alarm shall be OPERABLE.
APPLICABILITY: MODES 1 and 2. .
ACTIONS CONDITION REQUIRED ACTION C0HPLETION TIME A. One channel of rod A.1 Perform SR 3.1.4.1 Once within position indication (rod position 15 minutes inoperable for one or veri fication) . following any more control. rods. rod motion in that group B. Rod position deviation B.1 Perform SR 3.1.4.1 Once within alarm inoperable. (rodposition 15 minutes of veri fication) . movement of any control rod C. One control rod C.1 Perform SR 3.2.2.1 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> misaligned by (peaking factor
>8 inches.l fro any verification).
'otherfodin ts I 9froupf i QB g C.2 Reduce THERMAL POWER 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> C*cchY to s 75S; RTP. "," .Cd A" E.
Palisades Nuclear Plant 3.1.4-1 Amendment No. 01/20/98
k Control Rod Alignment I 3.1.4-
)
SURVEILLANCE REQUIREftENTS SURVEILLANCE FREQUENCY l
i l SR 3.1.4.1 Verify the position of each control rod to 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> be within 8 inches of all other control rods in its group.
CM E'd 1 SR 3.1.4.2 Perform a CHANNEL CHECK of the, rod position 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.: '
indication channels. -
i i
SR 3.1.4.3 verify the rod position deviation alarm is 92 days 5
OPERABLE.
i SR 3.1.4.4 Verify control rod freedom of movement by 92 days f
. moving each individual full-length4 rod that 1(
l isnotfullyinsertedintothereactor{ core CAMf0L
- 2 6 inches in either direction. ed '
SR 3.1.4.5 PerformaCHANNELCALIBRATIONoftheSod 18 months position indication channels.
I SR 3.1.4.6 Verify each full-length control rod drop Prior to time is s 2.5 seconds, reactor criticality,
' after each '
reinstallation of the reactor head ,
Palisades Nuclear Plant 3.1.4-3 Amendment No. 01/20/98
SDM B'3.1.1
- i
-BASES l APPLICABLE The minimum required SDM is assumed as an initial condition SAFETY ANALYSES in safety analysis. 'The. safety analysis (Ref. 2) l establishes an -SDM that ensures'specified acceptable fuel- !
' design limits are not exceeded for normal operation and A00s, with the assumption that the control rod of highest reactivity worth is fully withdrawn following a reactor trip. For MODE 5, the primary safety analysis that relies 5 on the SDM limits is the boron dilution analysis.
The acceptance criteria' for the SDM requirements' are that ,
i specified acceptable fuel design limits are maintained. This is done by ensuring that:
i ;
4 j a. The reactor can be made suberitical from all operating :
l conditions, transients, and Design Basis Events; s
I b. The reactivity transients associated with postulated a accident conditions are controllable within acceptable limits (Departure from Nucleate Boiling Ratio (DNBR), -
- fuel centerline temperature limit A00s, and -
i s 280 cal /gm energy deposition for the control rod d
ejection accident); and-i '
l c. The reactor will be maintained sufficiently
- subcritical to preclude inadvertent criticality in the 1
- shutdown condition.
The most limiting accident for the SDM requirements are
- based on a Main Steam Line Break (MSLB), as described in the
) accident analysis (Ref. 2). The increased steam flow l
- resulting from a pipe break in the main steam system causes '
an increased energy removal from the affected Steam I Generator (SG),~and consequently the PCS. This results in a Q.
l reduction of the primary coolant temperature. The resultant i j coolant shrinkage causes a reduction in pressure. In the 1 presence of a negative moderator temperature coefficient, j
- this cooldown causes an increase in core reactivity. A; TG l l t
- ; rete e de --- - . +ka r="--ity O' in "SL9 d::r:::::- !
[ =til the "^M 5 :1 : 1: r:::h:d. The most limiting MSLB, j with respect to potential fuel damage befere a reseter t-4p r ess w se.is a guillotine break of a main steam line d :id: x conte 4 amen 4 initiated at the end of core life. The positive r l reactivity addition from the moderator temperature decrease
- will terminate when the affected SG boils dry, thus ,
terminating PCS heat removal and cooldown. Following the i
! MSLB, a post trip return to power may occur; however,
! THERMAL POWER does not" violate the Safety Limit (SL)
4 1-..2- u- --..m . <-. u + v... no . K b
- ~
- Palisades' Nuclear Plant B 3.1.1-2 01/20/98
__ . . _ . . _ _ ~ ._ _ __ _ . . . _ _ ~ . - ,.
Control Rod Alignment B 3.1.4 I
BASES BACKGROUND The control rods are arranged into groups that are radially
- (continued) symmetric. Therefore, movement of the control rod groups ,
does not introduce radial asymmetries in the core power i
distribution. The shutdown and regulating rods provide the required reactivity worth for immediate reactor shutdown upon a reactor trip. The regulating rods also provide. ,
' reactivity (power level) control during normal operation I
and transients.
The axial position of shutdown and regulating rods is- !
f indicated by two separate and independent systems, which i
are 1) synchro based position indication system, and 2) the reed switch based position indication system.
' The synchro based position indication system measures the phase angle of a synchro geared to the CRDM rack. Full Lj l
' control rod travel corresponds to less than 1 turn of the synchro. Each control rod has its own synchro. The goJg, x.
Primary Information Processor (PIP)* scans and converts
}
i synchro outputs into inches of control rod withdrawal. The g resolution of this system is approximately 0.5 incpes.
i
' Each synchro also hasfgcam operated limit switch %hich can- X V j provide positive indication of control rod position.
' The reed switch based position indication system is referred to as the Secondary Position Indication (SPI) i system. This system provides a highly accurate indication
' of actual control rod position, but at a lower precision than the synchros. The reed switches are wired so that-the voltage read across the reed switch stack is proportional to rod position. The reed switches are spaced along a tube i
with a center to center spacing _ distance of 1.5 inches.
The resolution of the SPI reed 46 tacks is 1.5 inches. lne Q( ,,,,,
reed switches also provide input to the matrix indication 1
lights which provide control rod status indication for various key positions. To increase the reliability of the kr(
d system, there are redundant :;:t;;n; reed switcht51+veh which prevent false r:d 9:p indication in the event 4 reed (r j
- switch fails,te c h e. gn gg i
l' A control rod position deviation alarm is provided to
- alert the operator when any two control rods in the same l
group are more than 8 inches apart. This helps to ensure {J
^
any control rod misalignments are minimized. 4 j (A.o. k cu b<. y us.d Q e. M +m. $ff d*% W flf nb0t i
rm u, tm S PL o pM , m Csn yc.f& w et h + h<. bd Cn ?M)
M rtdondad +5 int. fif mde. to tN. foam ofr csn+r4L Nd ~
2 r*torbrtatni . Un frd. rod man i kina - a.nd fumit kof au2h n i .
.J' ()
- Palisades Nuclear. Plant .B 3.1.4-2 ~01/20/98 J ,
l H
Control Rod Alignment .]
B 3.1.4 ]
j BASES APPLICABLE Control rod' mis' alignment accidents are analyzed in the i SAFETY-ANALYSES safety analysis (Refs. 3 and 4). The accident analysis defines control rod misoperation as any event, with the . j exception of sequential group withdrawals, which could ,
- result from a single malfunction in the reactivity. control I systems. For example, control rod misalignment may be !
caused by a malfunction of the Rod Control System, or by operator error. A stuck rod may be caused by mechanical
- jamming. Inadvertent withdrawal of a single control rod
- may be caused by an electrical or mechanical failure in the
, Rod Control System. ' A dropped control rod could be caused 2
by an electrical or e e r m ' failure in the CRDM.- V/
^
%eMal
, The acceptance criteria for addressing control rod inoperability/misalignmentarethat:
- a. There shall be no violations of:
- 1. Specified Acceptable Fuel Design Limits (SAFDL),
j or l 2. Primary Coolant System (PCS) pressure boundary
- integrity; and l b. The core must remain subcritical after accident j transients.
Three types of misoperations are discussed in the safety analysis (Ref. 4). During movement of a group, one control i rod may stop moving while the other control rods in the group continue. This condition may cause excessive power j peaking. The second type of misoperations occurs if one control rod fails to insert upon a reactor trip and remains i stuck fully withdrawn. This condition requires an evaluation to determine that sufficient reactivity worth is held in the remaining control rods to meet the SDM requirement with'the maximum worth rod stuck fully withdrawn. If a control rod is stuck in the fully
. withdrawn position, its worth is added to the SDM
. requirement, since the safety analysis does not take two stuck rods into account. The third type of misoperations i occurs when one rod drops partially or fully into the reactor core. This event causes an initial power reduction r followed by a return towards the original power, due to
> positive reactivity feedback from the negative moderator temperature coefficient. Increased peaking during the power increase may result in excessive local Linear Heat Rates (LHRs).
l
. Palisades Nuclear Plant 'B 3.1.4-3 01/20/98
l l
Control Rod Alignment l B 3.1.4 BASES APPLICABLE The most limiting static misalignment occurs when Bank 4 is '
SAFETY ANALYSES fully inserted with one rod fully withdrawn (Bank 4 is (continued) 99 inches out of alignment with the rated Power Dependent Insertion Limit (PDIL). This event was bounded by the dropped full-length control rod event (Ref. 4).
Since the control rod drop incidents result in the most 4'Ch rapid approach to SAFDLs caused by a control rod misoperation, the accident analysis analyzed a single p,.ha$
full-length control rod drop. 'he -~ + "epid =p?"ca9 +^ g p f the GNRD 9AFnt may hp causpd by a d nnla full larm+h-c 0 F. t "^1 "^d d"^p.
The above control rod misoperations may or may not result in an automatic reactor trip. In the case of the full-length rod drop, a prompt decrease in core average power and a distortion in radial power are initially produced, which, when conservatively coupled, result in a local power and heat flux increase, and a decrease in DNBR parameters.
The results of the control rod misoperation analysis show that during the most limiting misoperation events, no i violations of the SAFDLs, fuel centerline temperature, or '
PCS pressure occur.
Control rod alignment satisfies Criteria 2 and 3 of 10CFR50.36(c)(2).
LC0 The limits on shutdown, regulating, and part-length rod alignments ensure that the assumptions in the safety analysis will remain valid. The requirements on OPERABILITY ensure that upon reactor trip, the control rods will be available and will be inserted to provide enough negative reactivity to shut down the reactor. The OPERABILITY requirements also ensure that the control rod banks maintain the correct power distribution and control rod alignment and that each control rod is capable of being moved by its CRDM. The OPERABILITY requirement for the part-length rods is that they are fully withdrawn and are capable of being moved by their CRDMs.
Palisades Nuclear Plant B 3.1.4-4 01/20/98
Control Rod Alignment B 3.1.4-enu feJ.hj e.t k & M nok or BASP C E OPIC*'-
LCO The requirement is to maintain'the ontrol rod alignment to ig (continued) within 8 inches between any control rod and all other rods in its group.To4+s helpe ensure thi r quirement is met, f the control rod position deviation ala must be OPERABLE and provide an alarm when any control rod becomes misaligned > 8 inches from any other rod in its group. The .
safety analysis assumes a total misalignment from fully withdrawn to fully inserted. This case bounds.the: safety analysis for a single rod in any intermediate position.
The primary rod position indication system is considered OPERABLE, for purposes of this specification, if the digital position readout, the PPC display, or the cam operated position indication lights give positive indication of rod position. The secondary rod position indication system is considered OPERABLE if the ,
magnetically operated reed switches are providing positive indication of rod position either via the plant process 1 1
computer or taking direct readings of the output from the j magnetic reed switches.
t Failure to meet the requirements of this LC0 may produce 1
- unacceptable power peaking factors and LHRs, or l unacceptable SDM, any of which may constitute initial l conditions inconsistent with the safety analysis, i
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- APPLICABILITY The requirements on control rod OPERABILITY and alignment are applicable in MODES 1 and 2 because these are the only MODES in which (e.g.,
the OPERABILITY neutron (or fission))
trippability power isofgenerated, and and alignment control rods have the potential to affect the safety of the i plant. In MODES 3, 4, 5, and 6, the alignment limits do l
not apply because the reactor is shut down and not
- producing fission power. In the shutdown MODES, the ,
OPERABILITY of the shutdown and regulating rods has the potential to affect the required SDM, but this effect can i
be compensated for by an increase in the boron
- concentration of the PCS. See LCO 3.1.1, " SHUTDOWN MARGIN
- (SDM)," for SDM in MODES 3, 4, and 5, and LC0 3.9.1, " Boron Concentration," for boron concentration requirements during
! refueling.
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Palisades Nuclear Plant B 3.1.4-5 01/20/98
Control Rod Alignment.
83.1.4' BASES ,
ACTIONS L,1 (continued) ,
When a Required Action cannot be completed within the required Completion Time, a controlled shutdown should be commenced. The allowed Completion Time of 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> is reasonable, . based on operating experience for reaching MODE 3 from full power conditions in an orderly manner and without challenging plant systems.
i SURVEILLANCE SR 3.1.4.1 -
l REQUIREMENTS Verification that individual control rod positions are i
within 8 inches of all other. control rods in the group at a f i 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Frequency allows the operator to detect a control l
rod that is beginning to deviate from its expected position. The specified Frequency takes into account other control rod position information that is continuously l
available to the operator in the control room, so that j during control rod movement, deviations can be detected.
Also protection can be provided by the control ~ rod
- deviation alarm. 1 SR 3.1.4.2 OPERABILITY of two control rod position indicator channels is required to determine control rod positions, and thereby i ensure compliance with the control rod alignment and- Ed insertion limits. Perfa n nce of a CHANNEL CHECK on the I i primary and secondary 7roTposition indication channels i provides confidence in the accuracy of the rod position
- indication systems. The control rod " full in" and " full l 4 out" lights, whi.ch correspond to the lower electrical limit !
and the upper electrical limit respectively, provide an l
. additional means for determining the control rod positions I when the control rods are at either their fully inserted or l fully withdrawn positions, j 1
i The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Frequency takes into consideration other l
.information continuously available to the operator in the
' control room, so that during control' rod movement, deviations can be detected, and protection can be provided by the control rod deviation alarm.
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Palisades Nuclear Plant B 3.1.4-9 01/20/98 l t- ,
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Shutdorn and Part-Length Rod Group Insertion Limits B 3.1.5
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BASES BACKGROUND They affect core power, burnup distribution, and add (continued) negative reactivity to shut down the reactor upon receipt of a reactor trip signal.
l APPLICABLE Accident analysis assumes that the shutdown rod groups are SAFETY ANALYSES fully withdrawn any time the reactor is critical. This ensures that:
- a. The minimum SDM is maintained; and
. b. The potential effects of a control rod ejection accident are limited to acceptable limits.
ed
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i Control rods are considered fully withdrawn at 128 inches,
- since this position places them in at@@ insignificant )(
reactivity worth region of the integral worth curve for each bank.
On a reactor trip, all full-length control rods (shutdown and regulating), except the most reactive rod, are assumed i to insert into the core. The shutdown and regulating rod groups shall be at or above their insertion limits and available to insert the required amount of negative reactivity on a reactor trip signal. The regulating rods may be partially inserted in the core as allowed by LC0 3.1.6, " Regulating Rod Group Position Limits." The
, shutdown rod group insertion limit is established to ensure i
that a sufficient amount of negative reactivity is
- available to shut down the reactor and maintain the
- required SDM (see LCO 3.1.1, " SHUTDOWN MARGIN (SDM))
following a reactor trip from full power. The combination of regulating rod and shutdown rods (less the most reactive 4 rod, which is assumed to remain fully withdrawn) is '
sufficient to take the reactor from full power conditions at rated temperature to zero power, and to maintain the required SDM at rated no load temperature (Ref. 2). The shutdown rod group insertion limit also limits the reactivity worth of an ejected shutdown rod.
Palisades Nuclear Plant B 3.1.5-2 01/20/98
Shutdown and Part-Length Rod Group Insertion Limits B 3.1.5 1
BASES I (d
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- SURVEILLANCE SR 3.1.5.1 (continued) l REQUIREMENTS C4wed.
l Since the/thutdown and oart-runatNVrod grcups are )h
- positioned manually by the control room operator,
- verification of shutdown and part-length rod group position '
- at a Frequency of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> is adequate to ensure that the-
{ shutdown and part-length rod groups are within their j insertion limits. Also, the 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Frequency takes into
- account other information available to the operator in the control room for the purpose of monitoring the status of
, the shutdown and part-length rod groups.
- REFERENCES 1. FSAR, Section 5.1
- 2. FSAR, Section 14.2
- 3. FSAR, Section 14.6 i
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Palisades Nuclear Plant B 3.1.5-6 01/20/98 ,
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R'egulating Rod Group Position Limits l B 3.1.6
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BASES l '
i LC0 The Power Dependent Insertion Limit (PDIL). alarm circuit is '
(continued) required to be OPERABLE for notification that the
- regulating rod groups are outside the required insertion limits. The Control Rod Out Of Sequence (CR005) alarm
! circuit is required to be OPERABLE for notification that the rods are not within the required sequence and. overlap j limits. When the PDIL or th2 CR005 alarm circuit is '
N
- inoperable, the verification of rod group positions is
' increased to ensure improper rod alignment is identified before unacceptable flux distr;bution occurs. 3 g r
! APPLICABILITY The requiating rod group sequence, overlap, and physical
- insertion limits shall be naintained with the reactor in MODES 1 and 2. These limits must be maintained, since they I l preserve the assumed power distribution, ejected rod worth, SDM, and reactivity rate insertion assumptions. 1
! Applicability in MODES 3, 4 and 5.is not required, since
! neither the power distribution nor ejected rod worth assumptions would be exceeded in these MODES. SDM is preserved in MODES 3, 4, and 5 by adjustments to the
- soluble boron concentration.
i
) The Applicability has been modified by a Note indicating the LC0 requirement is suspended SR 3.1.4.4 (rod exercise test). Control rod exercising verifies the freedom of the rods to move, and requires the individual regulating rods g
to move below the LCO limits which = 'd = u H y violate X
- the LC0 for their group. coasd l
j i I l ACTIONS A.1 and A.2 j Operation beyond the insertion limit may result in a loss l 1 of SDM and excessive peaking factors. The insertion limit i snould not be violated during normal' operation; this
! violation, however, may occur during transients when the operator is manually controlling the regulating rods in response to changing plant conditions.
3 % PhLcad Chos alams can L peld/ 6 8th wncke, kaAd
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7 Palisades Nuclear Plant B 3.F.6-6 01/20/98 i l 1
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i ENCLOSURE 3
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CONSUMERS ENERGY COMPANY l PALISADES PLANT 1 DOCKET 50-255 t
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,! CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS l RESPONSE TO DECEMBER 4, 1998 !
REQUEST FOR ADDITIONAL INFORMATION i
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! REVISED PAGES FOR CHAPTER 2.0 l
1 I !
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CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO DECEMBER 4, 1998 REQUEST FOR ADDITIONAL INFORMATION REVISED PAGES FOR CHAPTER 2.0 Paoe Chance Instructions Revise the Palisades submittal for conversion to Improved Technical Specifications by removing the pages identified below and inserting the attached pages. The revised pages are identified by date and contain vertical lines in the margin indicating the areas of change.
REMOVE PAGES IMIRT PAGES REV DATE NRC COMENT#
ATTACHMENT 1 TO ITS CONVERSION SUBMITTAL No page changes ATTACHMENT 2 TO ITS CONVERSION SUBMITTAL B 2.1.2-3 B 2.1.2-3 02/05/99 RAI 2.0-01 ATTACHMENT 3 TO ITS CONVERSION SUBMITTAL $
No page changes ATTACHMENT 4 10 ITS CONVERSION SUBMITTAL No page changes ATTACHMENT 5 IQ_ITS CONVERSION SUBMITTAL l B 2.0-8 8 2.0-S 02/05/99 RAI 2.0-01 l l
l ATTACHMENT 6 TO ITS CONVERSION SUBMITTAL No page changes 1 l
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PCS Pressure SLs B 2.1.2 BASES SAFETY LIMITS The maximum transient pressure allowable in the PCS pressure vessel under the ASME Code, Section III, is 110% of design pressure. The maximum transient pressure allowable in the PCS piping, valves, and fittings under 120% of design pressure (Ref. 6). The most limiting of these two allowances is the 110% of design pressure; therefore, the SL on maximum allowable PCS pressure is established at 2750 psia.
APPLICABILITY SL 2.1.2 applies in MODES 1, 2, 3, 4, 5, and 6 because this SL could be approached or exceeded in these MODES due to overpressurization events. In MODE 6 with the reactor vessel head installed and the reactor vessel head closure bolts less than fully tensioned the potential for an over pressurization event still exists. Although overpressurization of the PCS is impossible once the reactor vessel head is removed, the requirements of this SL apply as long as fuel is in the reactor. Once all the fuel has been removed from the reactor, the requirements of SL 2.1.2 no longer apply.
SAFETY LIMIT The following SL violation responses are applicable to the VIOLATIONS PCS pressure SLs.
2.2.2.1 If the PCS pressure SL is violated when the reactor is in MODE 1 or 2, the requirement is to restore compliance and be in MODE 3 within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
With PCS pressure greater than the value specified in SL 2.1.2 in MODE l'or 2, the pressure must be reduced to below this value. A pressure greater than the value specified in SL 2.1.2 exceeds 110% of the PCS design pressure and may challenge system integrity.
The allowed Completion Time of I hour provides the operator time to complete the necessary actions to reduce PCS pressure by terminating the cause of the pressure increase, removing mass or energy from the PCS, or a combination of tt.ase actions, and to establish MODE 3 conditions.
Palisades Nuclear Plant B 2.1.2-3 02/05/99
RCS pressure SLl W aitkr)l B 2.1.2 i
BASES (continued)
APPLICABILITY SL 2.1.2 applies in MODES 1, 2, 3, 4,, because this SL '
could be approached or exceeded in these MODES due to overoressuritation events. JThe SL 1s not pplicacie in MODE 6 ecause the reacKr vessel head c sure bolts are 01, t fully ightened, makinf it unlikely tha the RCS can be 1
/ pres urized.l SAFETY LIMIT The following SL violation responses are applicable to the j V10LAT10N3 S pressure SLs. h
- maer 2.2.2.1 hfAd inficlkd and i ftador
{tMal C./0$vfc hold us han If the 5 pressure SL is violated when the reactor is in MODE 1 'or 2, the requirement is to restore cumpliance and be lh tull'/ fcnsioned; 4N, %nia l 6r an carre,33pq c hen ey in MODE 3 within I hour.
Shu- CMU, Alfhe h With S pressure greater than the value specified in SL 2.1.2 in MODE 1 or 2, the pressure must be reduced to lh OWrfftnuriphen o f he fti below this value. A pressure greater E the value lh 13 le%3rdit once tne (cQ 4hr specified in SL 2.1.2 exceeds 110% of the SCS design g VfA4'l hoed 13 (tmowd , tM, pressure and may challenge system integrityV, h-k%*M3 d blJ 3L The allowed Completion Time of I hour provides the perator C.pp(j cc s kL time to complete the necessary actions to reduce E S pressure by terminating the cause of the pressure increase, lh
- 3 M Ynt fledBC Onu.
G Lt. fAs ( W ) h u [, e removing mass or energy from the 5, or a combination of these actions, and to establish ODE 3 conditions, lh i
(c.rh6Vtd bm M kb6 f Yh It if WAund og 2.2.2.2 Sb b.2 no inhet If the S pressure SL is exceeded in MODE 3, 4, pressure must be restored to within the SL value within RCS I
l Eff d . 5 minutes. g l Exceeding the S pressure SL in MODE 3, 4,97@ is potentially more severe than exceeding this SL in MODE 1 h '
or 2, since the reactor vessel temperature may be lower and the vessel material, consequently, less ductile. As such, i pressure must be reduced to less than the SL within 5 minutes. This action does not require reducing MODES, since this would require reducing temperature, which would j l
(continued)
CEOG STS B C5% Rev 1, 04/07/95 I
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ENCLOSURE 4 j-i l i CONSUMERS ENERGY COMPANY )
PALISADES PLANT i 4 DOCKET 50-255 l i
CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS l RESPONSE TO DECEMBER 4, 1998 l REQUEST FOR ADDITIONAL INFORMATION i
l l REVISED PAGES FOR SECTION 3.1 j; ,
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CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO DECEMBER 4, 1998 REQUEST FOR ADDITIONAL INFORMATION REVISED PAGES FOR SECTION 3.1 fage Chanae Instructions Revise the Palisades submittal for conversion to Improved Technical Specifications by removing the pages identified below and inserting the attached pages. The revised pages are identified by date and contain vertical lines in the margin indicating the areas of change.
REMOVE PAGES INSERT PAGES REV DATE NRC COMENT#
ATTACHMENT 1_TO ITS CONVERSION SUBMITTAL ITS 3.1.1-1 ITS 3.1.1-1 02/05/99 RAI 3.1-01 ITS 3.1.1-2 ------ RAI 3.1-01 ITS 3.1.4-1 ITS 3.1.4-1 02/05/99 editorial ITS 3.1.4-2 ITS 3.1.4-2 02/05/99 RAI 3.1-04 ITS 3.1.4-3 ITS 3.1.4-3 02/05/99 editorial ITS 3.1.5-1 ITS 3.1.5-1 02/05/99 RAI 3.1-07
ITS 3.1.5-2 02/05/99 RAI 3.1-07 ITS 3.1.6-1 ITS 3.1.6-1 02/05/99 Tech change ATTACHMENT 2 TO ITS CONVERSION SUBMITTAL ITS B 3.1.1-2 ITS B 3.1.1-2 02/05/99 editorial ITS B 3.1.1-3 ITS B 3.1.1-3 02/05/99 RAI 3.1-01 RAI 3.1-04 ITS B 3.1.1-5 ITS B 3.1.1-5 02/05/99 RAI 3.1-01 ITS B 3.1.1-6 ITS B 3.1.1-6 02/05/99 RAI 3.1-01 ITS B 3.1.4-1 ITS B 3.1.4-1 02/05/99 RAI 3.1-04 ITS B 3.1.4-2 ITS B 3.1.4-2 02/05/99 editorial ITS B 3.1.4-3 ITS B 3.1.4-3 02/05/99 editorial ITS B 3.1.4-4 ITS B 3.1.4-4 02/05/99 RAI 3.1-04 ITS B 3.1.4-5 ITS B 3.1.4-5 02/05/99 editorial ITS B 3.1.4-8 ITS B 3.1.4-8 02/05/99 RAI 3.1-04 ITS B 3.1.4-9 ITS B 3.1.4-9 02/05/99 editorial ITS B 3.1.4-10 ITS B 3.1.4-10 02/05/99 RAI 3.1-04 ITS B 3.1.4-11 ITS B 3.1.4-11 02/05/99 RAI 3.1-04 ITS B 3.1.5-1 ITS B 3.1.5-1 02/05/99 RAI 3.1-04 ITS B 3.1.5-2 ITS B 3.1.5-2 02/05/99 editorial ITS B 3.1.5-3 ITS B 3.1.5-3 02/05/99 RAI 3.1-07 ITS B 3.1.5-4 ITS B 3.1.5-4 02/05/99 RAI 3.1-04 ITS B 3.1.5-5 ITS B 3.1.5-5 02/05/99 RAI 3.1-07 ITS B 3.1.5-6 ITS B 3.1.5-6 02/05/99 editorial ITS B 3.1.6-1 ITS B 3.1.6-1 02/05/99 RAI 3.1-04 ITS B 3.1.6-4 ITS B 3.1.6-4 02/05/99 RAI 3.1-04 ITS B 3.1.6-5 ITS B 3.1.6-5 02/05/99 RAI 3.1-10 ITS B 3.1.6-6 ITS B 3.1.6-6 02/05/99 editorial ITS B 3.1.6-7 ITS B 3.1.6-7 02/05/99 editorial ITS B 3.1.6-8 ITS B 3.1.6-8 02/05/99 editorial
CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO DECEMBER 4, 1998 REQUEST FOR ADDITIONAL INFORMATION REVISED PAGES FOR SECTION 3.1 ATTACHMENT 3 TO ITS CONVERSION SUBMITTAL CTS 3.1.1, pg 3-50 CTS 3.1.1, pg 3-50 02/05/99 RAI 3.1-01 CTS 3.1.4, pg 3-52 CTS 3.1.4, pg 3-52 02/05/99 Tech change CTS 3.1.4, pg 4-11 CTS 3.1.4, pg 4-11 02/05/99 Tech change CTS 3.1.5, pg 3-53 CTS 3.1.5, pg 3-53 02/05/99 RAI 3.1-07 DOC 3.1.1, pg 1 of 6 DOC 3.1.1, pg 1 of 5 02/05/99 RAI 3.1-01 through through DOC 3.1.1, pg 6 of 6 DOC 3.1.1, pg 5 of 5 DOC 3.1.4, pg 4 of 7 DOC 3.1.4, pg 4 of 7 02/05/99 Tech cnange DOC 3.1.5, pg 1 of 5 DOC 3.1.5, pg 1 of 4 02/05/99 RAI 3.1-07 through through DOC 3.1.5, pg 5 of 5 DOC 3.1.5, pg 4 of 4 4
ATTACHMENT 4 TO ITS CONVERSION SUBMITTAL NSHC 3.1.1, pg 1 of 3 NSHC 3.1.1, pg 1 of 1 02/05/99 RAI 3.1-01 through NSHC 3.1.1, pg 3 of 3 AIIAClitimL5 TO ITS CONVERSJDN SUBMITTAL
, NUREG 3.1-1 NUREG 3.1.1 02/05/99 RAI 3.1-01 NUREG 3.1-1 insert -----
02/05/99 RAI 3.1-01 NUREG 3.1-8 NUREG 3.1-8 02/05/99 editorial l NUREG 3.1-11 insert NUREG 3.1.11 insert 02/05/99 editorial NUREG 3.1-12 NUREG 3.1-12 02/05/99 editorial NUREG 3.1-13 NUREG 3.1-13 02/05/99 RAI 3.1-07 l NUREG 3.1-15 NUREG 3.1-15 02/05/99 Tech change NUREG B 3.1-2 NUREG B 3.1-2 02/05/99 editorial NUREG B 3.1-3 NUREG B 3.1-3 02/05/99 editorial NUREG B 3.1-4 NUREG B 3.1-4 02/05/99 RAI 3.1-01 NUREG B 3.1-5 NUREG B 3.1-5 02/05/99 RAI 3.1-01 NUREG B 3.1-6 insert NUREG B 3.1-6 insert 02/05/99 RAI 3.1-01 NUREG B 3.1-24 insert NUREG B 3.1-24 insert 02/05/99 editorial NUREG B 3.1-26 NUREG B 3.1-26 02/05/99 editorial NUREG B 3.1-26 insert NUREG B 3.1-26 insert 02/05/99 editorial NUREG B 3.1-31 insert NUREG B 3.1-31 insert 02/05/99 editorial i
NUREG B 3.1-32 NUREG B 3.1-32 02/05/99 Tech change NUREG B 3.1-34 NUREG B 3.1-34 02/05/99 RAI 3.1-04 NUREG B 3.1-34 insert NUREG B 3.1-34 insert 02/05/99 Tech change NUREG B 3.1-35 NUREG B 3.1-35 02/05/99 editorial NUREG B 3.1-36 insert NUREG B 3.1-36 insert 02/05/99 RAI 3.1-04
(#1 - 6, 2 pgs) (#1-6,1pg) RAI 3.1-07 1
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l CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS
- RESPONSE TO DECEMBER 4, 1998 REQUEST FOR ADDITIONAL INFORMATION REVISED PAGES FOR SECTION 3.1 ATTACHMENT 1 TO ITS CONVERSION SUBMITTAL (continued)
NUREG B 3.1-37 NUREG B 3.1-37 02/05/99 RAI 3.1-07 NUREG B 3.1-37 insert NUREG B 3.1-37 insert 02/05/99 RAI 3.1-07 NUREG B 3.1-39 NUREG B 3.1-39 02/05/99 editorial NUREG B 3.1-42 insert NUREG B 3.1-42 insert 02/05/99 RAI 3.1-10 NUREG B 3.1-43 NUREG B 3.1-43 02/05/99 editorial ATTACHMENT 6 TO ITS CONVERSION SUBMITTAL JFD 3.1.1, pg 2 of 3 JFD 3.1.1, pg 2 of 3 02/05/99 RAI 3.1-01 JFD 3.1.5, pg 2 of 7 JFD 3.1.5, pg 2 of 7 02/05/99 Tech change JFD 3.1.6, pg 2 of 4 JFD 3.1.6, pg 2 of 4 02/05/99 Tech change JFD 3.1.6, pg 4 of 4 JFD 3.1.6, pg 4 of 4 02/05/99 RAI 3.1-07 JFD 3.1.7, pg 4 of 5 JFD 3.1.7, pg 4 of 5 02/05/99 RAI 3.1-10
- SDML
. 3.1.1-3.1 -REACTIVITY CONTROL SYSTEMS 3.1.1 SHUTDOWN' MARGIN (SDM)
LCO 3.1.1 lSDM shall be within the' limits specified in the COLR. -l APPLICABILITY: MODE 3, 4, and 5.
ACTIONS ,_
' CONDITION REQUIRED ACTION COMPLETION: TIME A. SDM not within limit. A.1 Initiate boration to 15 minutes restore SDM to within limit.
SURVEILLANCE REQUIREMENTS SURVEILLANCE I
- FREQUENCY
- SR 3.1.1.1 Verify SDM to be within limits. 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> ll I
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i Palisades Nuclear Plant :3.1.1-1 Amendment No. 02/05/99 l
Control Rod Alignment 3.1.4 3.1 REACTIVITY CONTROL SYSTEMS 3.1.4 Control Rod Alignment LC0 3.1.4 All control rods, including their position indication l channels, shall be OPERABLE and aligned to within 8 inches of all other rods in their respective group, and the control rod position deviation alarm shall be OPERABLE.
APPLICABILITY: MODES 1 and 2.
ACTIONS CONDITION REQUI. LED ACTION COMPLETION TIME i
A. One channel of rod A.1 Perform SR 3.1.4.1 Once within position indication (rod position 15 minutes inoperable for one or verification) . following any more control rods. rod motion in that group l B. Rod position deviation 8.1 Perform SR 3.1.4.1 Once within alarm inoperable. (rod position 15 minutes of veri fication) . movement of any control rod C. One control rod C.1 Perform SR 3.2.2.1 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> misaligned by (peaking factor
> 8 inches. verification). l l
DB C.2 Reduce THERMAL POWER 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> to s 75% RTP.
Palisades Nuclear Plant 3.1.4-1 Amendment No. 02/05/99
)
Control Rod Alignment. -
3.1.4 _
ACTIONS ;
CONDITION REQUIRED ACTION COMPLETION TIME .)
l D. One full-length D.1 Restore control rod Prior to l j control rod innovable, to OPERABLE status. entering M00E'2 1 but trippable. from MODE 3-i t
E. Required Action and E.1 Be in MODE 3. 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />s-associated Completion i Time not net.
- 08 !
One or more control l rods inoperable for .
reasons other than !
Condition D. 1
! I 08 )
- i i Two or more control l j rods misaligned'by i > 8 inches.
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Both rod position
< indication channels '
inoperable for one or more control rods.
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- Palisades Nuclear Plant- 3.1.4-2 -Amendment No. 02/05/99 ;
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Control Rod Alignment '
3.1.4 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.4.1 Verify the position of each control rod to 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> be within 8 inches of all other control rods in its group.
SR 3.1.4.2 Perform a CHANNEL CHECK of the control rod 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> I position indication channels.
SR 3.1.4.3 Verify control rod freedom of movement by 92 days l moving each individual full-length control l rod that is not fully inserted into the reactor core 2 6 inches in either di rection.
SR 3.1.4.4 Verify the rod position deviation alarm is 18 months l OPERABLE.
SR 3.1.4.5 Perform a CHANNEL CALIBRATION of the 18 months control rod position indication channels. l SR 3.1.4.6 Verify each full-length control rod drop Prior to time is s 2.5 seconds. reactor criticality, after each reinstallation of the reactor head Palisades Nuclear Plant 3.1.4-3 Amendment No. 02/05/99
Shutdown and Part-Length Rod Group Insertion Limits .
3.1.5 :
i 3.1 REACTIVITY CONTROL SYSTEMS i 3.1.5 Shutdown ans.Part-Length Control Rod Group Insertion Limits l ,
LCO 3.1.5 All shutdown and part-length rod groups shall be withdrawn to 2 128 inches.
APPLICABILITY: MODE 1, MODE 2 with any regulating rod withdrawn above 5' inches.- j
NOTE---------------------------- l This LC0 is not applicable while performing.SR 3.1~.4.3 .l 3 (rodexercisetest). I ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more shutdown A.1 Declare affected Immediately l l or part-length rods control rod (s) l l not within limit. inoperable and enter l 3 the applicable l .
Conditions and l ;
Required Actions of l
LCO 3.1.4. l u
1 B. Required Action and 8.1 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> j j associated Completion
! Time not met.
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! Palisades Nuclear' Plant '3.1.5-1. Amendment _No.- 02/05/99
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Shutdown and Part-Length Rod Group Insertion Limits <-
3.1.5' SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.5.1 Verify each shutdown and part-length rod 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> group is withdrawn 2 128 inches. f 1
l Palisades Nuclear Plant' 3.1.5-2 Amendment No. 02/05/99
Regulating Rod Group Position Limits 3.1.6 3.1 REACTIVITY CONTROL SYSTEMS 3.1.6 Regulating Rod Group Position Limits LC0 3.1.6 The Power Dependent Insertion Limit (PDIL) alarm circuit and the Control Rod Out Of Sequence (CR005) alarm circuit shall be OPERABLE, and the regulating rod groups shall be limited to the withdrawal sequence, overlap, and insertion limits specified in the COLR.
APPLICABILITY: MODES 1 and 2.
NOTE----------------------------
This LC0 is not applicable while performing SR 3.1.4.3 l (rodexercisetest).
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME I
A. Regulating rod grou'ps A.1 Restore regulating 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> inserted beyond the rod groups to within insertion limit. limits.
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A.2 Reduce THERMAL POWER 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 2
to less than or equal
. to the fraction of RTP allowed by the regulating rod group position and ,
insertion limits l I
specified in the COLR.
Palisades Nuclear Plant 3.1.6 1 Amendment No. 02/05/99 !
SDM B 3.1.1 BASES APPLICABLE The minimum required SDM is assumed as an initial condition SAFETY ANALYSES in safety analysis. The safety analysis (Ref. 2) establishes an SDM that ensures specified acceptable fuel design limits are not exceeded for normal operation and 1 A00s, with the assumption that the control rod of highest reactivity worth is fully withdrawn following a reactor trip. For MODE 5, the primary safety analysis that relies on the SDM limits is the baron dilution analysis.
The acceptance criteria for the SDM requirements are that specified acceptable fuel design limits are maintained. This is done by ensuring that:
- a. The reactor can be made subcritical from all operating conditions, transients, and Design Basis Events;
- b. The reactivity transients associated with postulated :
accident conditions are controllable within acceptable I limits (Departure from Nucleate Boiling Ratio (DNBR),
fuel centerline temperature limit A00s, and i s 280 cal /gm energy deposition for the control rod I ejection accident); and
- c. The reactor will be maintained sufficiently subcritical to preclude inadvertent criticality in the shutdown condition.
The most limiting accident for the SDM requirements are based on a Main Steam Line Break (MSLB), as described in the accident analysis (Ref. 2). The increased steam flow resulting from a pipe break in the main steam system causes an increased energy removal from the affected Steam Generator (SG), and consequently the PCS. This results in a reduction of the primary coolant temperature. The resultant coolant shrinkage causes a reduction in pressure. In the presence of a negative moderator temperature coefficient, this cooldown causes an increase in core reactivity. The l most limiting MSLB with respect to potential fuel damage is a guillotine break of a main steam line initiated at the end of core life. The positive reactivity addition from the moderator temperature decrease will terminate when the affected SG boils dry, thus terminating PCS heat removal and cooldown. Following the MSLB, a post trip return to power may occur; however, THERMAL POWER does not violate the Safety Limit (SL) requirement of SL 2.1.1. l Palisades Nuclear Plant B 3.1.1-2 02/05/99
SDM B 3.1.1 BASES APPLICABLE In addition to the limiting MSLB transient, the SOM SAFETY ANALYSES requirement for MODES 3 and 4 must also protect against an (continued) inadvertent boron dilution; (Ref. 3) and an uncontrolled control rod bank withdrawal from subcritical conditions (Ref. 5).
Each of these events is discussed below.
In the boron dilution analysis, the required SDM defines the reactivity difference between an initial subcritical baron concentration and the corresponding critical boron concentration. These values, in conjunction with the configuration of the PCS and the assumed dilution flow rate, directly affect the results of the analysis. This event is most limiting at the beginning of core life when critical beron concentrations are highest.
The witharewal of a control rod bank from subcritical conditions adds reactivity to the reactor core, causing both the core power level and heat flux to increase with corresponding increases in reactor coolant temperatures and pressure. The withdrawal of control rod banks also produce i a time dependent redistribution of core power. !
Depending on the system initial conditions and reactivity i insertion rate, the uncontrolled control rod banks i withdrawal transient is terminated by either a high power ;
trip or a high pressurizer pressure trip. In all cases, ;
power level, PCS pressure, linear heat rate, and the DNBR do !
not exceed allowable limits.
l SDM satisfies Criterion 2 of 10 CFR 50.36(c)(2). j l
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LCO The MSLB (Ref. 2) and the boron dilution (Ref. 3) accidents I are the most limiting analyses that establish the value for l !
SDM. For MSLB accidents, if the LC0 is violated, there is a l potential to exceed the DNBR limit and to exceed 10 CFR 100, l
" Reactor Site Criteria," limits (Ref. 4). For the boron dilution accident, if the LC0 is violated, then the minimum required time assumed for operator action to terminate dilution may no longer be applicable.
SDM is a core physics design condition that can be ensured through full-length control rod positioning (regulating and l shutdown rods) and through the soluble boron concentration. j Palisades Nuclear Plant B 3.1.1-3 02/05/99
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SDM B 3.1.1 BASES SURVEILLANCE SR 3.1.1.1 l REQUIREMENTS f SDM is verified by a reactivity balance calculation,
- considering the listed reactivity effects
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- b. Control rod positions;
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- c. PCS average temperature;
- d. Fuel burnup based on gross thermal energy generation;
- e. Xenon concentration; and i
- f. Isothermal Temperature Coefficient (ITC).
. Using the ITC accounts for Doppler reactivity in this
, calculation because the reactor is subcritical and .the fuel temperature will be changing at the same rate as the PCS.
Samarium is not considered in the reactivity analysis since the analysis assumes that the negative reactivity due to Samarium is offset by the positive reactivity of Plutonium i built in.
SR 3.1.1.1 requires SDM to be within the limits specified in the COLR. This SDM value ensures the consequences of an !
MSLB, will be acceptable as a result of a cooldown of the
- PCS which adds positive reactivity in the presence of a negative moderator temperature coefficient as well as the
, other events described in the Applicable Safety Analysis.
As such, the requirements of this SR must be met whenever the plant is in MODES 3, 4, and 5.
The Frequency of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> for the verification of SDM is _
l based on the generally slow change in required boron j concentration, and also allows sufficient time for the operator to collect the required data, which may include performing a boron concentration analysis, and completing the calculation.
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Palisades Nuclear Plant B 3.1.1-5 02/05/99
, . SDM B 3.1.1 BASES
- REFERENCES' '1. FSAR,-Section 5.1 1
- 2. FSAR, Section 14.14
- 3. FSAR, Section 14.3 1 4. 10 CFR 100
- 5. FSAR, Section 14.2 j
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Palisades Nuclear Plant B 3.1.1-6 02/05/99 f
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lCcntrol Rod Alignment.-
B 3.1.4 l B 3.1 REACTIVITY CONTROL SYSTEMS. !
B 3.1.4 Control Rod Alignment' ;
. t BASES BACKGROUND The OPERABILITY (e.g., trippability) of the shutdown anJ.
regulating rods is an initial assumption in all safety analyses that assume full-length control' rod insertion upon l-i reactor trip. Maximum control rod misalignment is'an initial assumption in the safety analysis that directly , ,
affects core power distributions and assumptions of:
3 available SDN.
The Palisades Nuclear. Plant design criteria contain the.
applicable criteria for these reactivity and power distribution design r3quirements'(Ref. 1). < s Mechanical or electrical failures may cause a control rod '
to become inoperable or to become misaligned from its group. Control rod misalignment may cause increased power l '
peaking, due to the asymmetric reactivity' distribution,. and a reduction in the total available control ~ rod worth for .
reactor shutdown. Therefore, control. rod alignment and OPERABILITY are related to core operation in design power peaking limits and the core design requirement of a minimum SDN. ,
Limits on control rod alignment and OPERABILITY.have been established, and all control rod positions are monitored
, and controlled during power operation to ensure that the power distribution and reactivity limits defined by the
- design power peaking and SDM_ limits are preserved.
! Control rods are moved by their Control Rod Drive
} rate of approximately 46 inches per minute. Although the- ,
j ability to move a full-length control rod by its drive l mechanism is not an initial assumption used in the safety ,
analyses, it is required to support OPERABILITY. As such, i the inability to move a full-length control rod results in ,
that full-length control rod being inoperable.
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p . Palisades Nuclear Plant B 3.1.4-1 02/05/99 .
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Control Rod Alignment B 3.1.4 BASES BACKGROUND The control rods are arranged into groups that are radially (continued) symmetric. Therefore, movement of the control rod groups does not introduce radial asymmetries in the core power i distribution. The shutdown and regulating rods provide the '
required reactivity worth for imediate reactor shutdown upon a reactor trip. The regulating rods also provide ;
reactivity (power level) control during normal operation l and transients.
The axial position of shutdown and regulating rods is indicated by two separate and independent systems, which are 1) synchro based position indication system, and 2) the reed switch based position indication system.
The synchro based position indication system measures the phase angle of a synchro geared to the CRDM rack. Full control rod travel corresponds to less than 1 turn of the synchro. Each control rod has its own synchro. The Primary Information Processor (PIP) node scans and converts l synchro outputs into inches of control rod withdrawal. The resolution of this system is approximately 0.5 inches.
Each synchro also has cam operated limit switches which can l provide positive indication of control rod position.
The reed switch based position indication system is referred to as the Secondary Position Indication (SPI) system. This system provides a highly accurate indication of actual control rod position, but at a lower precision than the synchros. The reed switches are wired so that the voltage read across the reed switch stack is proportional to rod position. The reed switches are spaced along a tube with a center to center spacing distance of 1.5 inches.
The resolution of the SPI reed switch stacks is 1.5 inches. l The reed switches also provide input to the matrix indication lights which provide control rod status indication for various key positions. To increase the reliability of the system, there are redundant reed switches which prevent false indication in the event an individual reed switch fails.
A control rod position deviation alarm is provided to alert the operator when any two control rods in the same group are more than 8 inches apart. This helps to ensure any control rod misalignments are minimized. The alarm can be generated by either the SPI system or PIP node since the SPI system, in conjunction with the host computer, is redundant to the PIP node in the task of control rod measurements, control rod monitoring, and limit processing.
Palisades Nuclear. Plant B 3.1.4-2 02/05/99
Control Rod Alignment B 3.1.4 BASES APPLICABLE Control rod misalignment accidents are analyzed in the SAFETY ANALYSES safety analysis (Refs. 3 and 4). The accident analysis defines control rod misoperation as any event, with the exception of sequential group withdrawals, which could result from a single malfunction in the reactivity control systems. For example, control rod misalignment may be caused by a malfunction of the Rod Control System, or by operator error. A stuck rod may be caused by mechanical jamming. Inadvertent withdrawal of a single control rod may be caused by an electrical or mechanical failure in the Rod Control System. A dropped control rod could be caused by an electrical or mechanical failure in the CRDM. l The acceptance criteria for addressing control rod inoperability/misalignmentarethat:
- a. There shall be no violations of:
- 1. Specified Acceptable Fuel Design Limits (SAFDL),
or
- 2. Primary Coolant System (PCS) pressure boundary integrity; and
- b. The core must remain subcritical after accident transients.
Three types of misoperations are discussed in the safety analysis (Ref. 4). During movement of a group, one control rod may stop moving while the other control rods in the group continue. This condition may cause excessive power peaking. The second type of misoperations occurs if one control rod fails to insert upon a reactor trip and remains stuck fully withdrawn. This condition requires an evaluation to determine that sufficient reactivity worth is ,
held in the remaining control rods to meet the SDM l requirement with the maximum worth rod stuck fully l withdrawn. If a control rod is stuck in the fully j withdrawn position, its worth is added to the SDM requirement, since the safety analysis does not take two stuck rods into account. The third type of misoperations ,
occurs when one rod drops partially or fully into the !
reactor core. This event causes an initial power reduction followed by a return towards the original power, due to positive reactivity feedback from the negative moderator temperature coefficient. Increased peaking during the l power increase may result in excessive local Linear Heat i Rates (LHRs).
l Palisades Nuclear Plant B 3.1.4-3 02/05/99
Control Rod Alignment B 3.1.4 BASES APPLICABLE The most limiting static misalignment occurs when Bank 4 is SAFETY ANALYSES fully inserted with one rod fully withdrawn (Bank 4 is (continued) 99 inches out of alignment with the rated Power Dependent Insertion Limit (PDIL). This event was bounded by the dropped full-length control rod event (Ref. 4).
Since the control rod drop incidents result in the most rapid approach to SAFDLs caused by a control rod misoperation, the accident analysis analyzed a single full-length control rod drop. l The above control rod misoperations may or may not result in an automatic reactor trip. In the case of the full-length rod drop, a prompt decrease in core average power and a distortion in radial power are initially produced, which, when conservatively coupled, result in a local power and heat flux increase, and a decrease in DNBR parameters.
The results of the control rod misoperation analysis show that during the most limiting misoperation events, no violations of the SAFDLs, fuel centerline temperature, or PCS pressure occur.
Control rod alignment satisfies Criteria 2 and 3 of 10 CFR 50.36(c)(2).
LC0 The limits on shutdown, regulating, and part-length rod 3
alignments ensure that the assumptions in the safety
, analysis will remain valid. The requirements on OPERABILITY ensure that upon reactor trip, the full-length l control rods will be available and will be inserted to provide enough negative reactivity to shut down the reactor. The OPERABILITY requirements also ensure that the control rod banks maintain the correct alignment and that each full-length control rod is capable of being moved by its CRDM. The OPERABILITY requirement for the part-length-rods is that they are fully withdrawn. (
Palisades Nuclear Plant 0 3.1.4-4 02/05/99-
Control Rod Alignment B 3.1.4 BASES LC0 The requirement is to maintain the control rod alignment to (continued) within 8 inches between any control rod and a'11 other rods in its group. To help ensure this requirement is met, the control rod position deviation alarm generated by either the PIP node or the SPI system, must be OPERABLE and provide an alarm when any control rod becomes misaligned
> 8 inches from any other rod in its group. The safety analysis assumes a total misalignment from fully withdrawn to fully inserted. This case bounds the safety analysis for a single rod in any intermediate position.
The primary rod position indication system is considered OPERABLE, for purposes of this specification, if the digital position readout, the PPC display, or the cam operated position indication lights give positive indication of rod position. The secondary rod position indication system is considered OPERABLE if the i magnetically operated reed switches are providing positive !
indication of rod position either via the plant process computer or taking din.ct readings of the output from the ,
magnetic reed switches.
Failure to meet the requirements of this LC0 may produce unacceptable power peaking factors and LHRs, or unacceptable SDM, any of which may constitute initial conditions inconsistent with the safety analysis, i
APPLICABILITY The requirements on control rod OPERABILITY and alignment I are applicable in MODES 1 and 2 because these are the only l MODES in which neutron (or fission) power is generated, and i the OPERABILITY (e.g., trippability) and alignment of control rods have the potential to affect the safety of the ;
plant. In MODES 3, 4, 5, and 6, the alignment limits do
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not apply because the reactor is shut down and not producing fission power. In the shutdown MODES, the OPERABILITY of the shutdown and regulating rods has the !
potential to affect the required SDM, but this effect can i I
be compensated for by an increase in the boron concentration of the PCS. See LCO 3.1.1, " SHUTDOWN MARGIN (SDM)," for SDM in MODES 3, 4, and 5, and LC0 3.9.1, " Boron l Concentration," for boron concentration requirements during i refueling.
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Palisades Nuclear Plant f 1.4-5 02/05/99 l
t Control Rod Alignment B 3.1.4 BASES ACTIONS Q.d (continued) . .
Condition D is entered whenever it is discovered that a -
single full-length control rod can not be moved by its l-i operator yet the control rod is still capable of being tripped. Although the ability to move a full-length l control rod is not an. initial assumption used in the safety
- analyses, it does relate to full-length control rod j OPERABILITY. The inability to move a full-length control -
- rod by its operator may be indicative of a systemic failure (other than trippability) which could potentially affect 3
other rods. Thus, declaring a full-length control rod l inoperable in this instance is conservative since it~1imits the number of full-length control rods which can not be j moved by their operators to only one. The Completion Time to restore an inoperable control rod to OPERABLE status is stated as prior to entering MODE 2 from HODE 3. Thi.
- Completion Time allows unrestricted operation in MODES 1 and 2 while conservatively preventing a reactor startup with an imovable full-length control rod. l
. E.d If the Required Action or associated Completion Time of Condition A, Condition B, Condition C, or Condition 0 is not met; one or more control rods are inoperable for reasons other than Condition 0; or two or more control rods are misaligned by > 8 inches, or two channels of control rod position indication are inoperable for one or more control rods, the plant is required to be brought to MODE 3. By being brought to MODE 3, the plant is brought outside its MODE of applicability. Continued operation is not allowed in the case of more than one control rod i
misaligned from any other rod in its group by > 8 inches, or two or more rods inoperable. This is because these cases may be indicative of a loss of SDM and power re-distribution, and a loss of safety function, respectively.
i Also, if no rod position indication exists for one or more control rods, continued operation is not allowed because the safety analysis assumptions of rod position cannot be ,
ensured.
4 Palisades Nuclear Plant B 3.1.4-8 02/05/99 4
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Control Rod Alignment l
B 3.1.4 J BASES ACTIONS E.d (continued) l l When a Required Action cannot be completed within the
- required Completion Time, a controlled shutdown should be J comenced. The allowed Completion Time of 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> is l i reasonable, based on operating experience, for reaching j MODE 3 from full power conditions in an orderly manner and '
j without challenging plant systems.
SURVEILLANCE SR 3.1.4.1 REQUIREMENTS Verification that individual control rod positions are
! within 8 inches of all other control rods in the group at a 4 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Frequency allows the operator to detect a control rod that is beginning to deviate from _its expected position. The specified Frequency takes into account other control rod position information that is continuously I availabie to the operator in the control room, so that during control rod movement, deviations can be detected. j Also protection can be provided by the control rod i 4
deviation alarm. l i
i SR 3.1.4.2 OPERABILITY of two control rod position indicator channels is required to determine control rod positions, and thereby ensure compliance with the control rod alignment and insertion limits. Performance of a CHANNEL CHECK on the primary and secondary control rod position indication l .
channels provides confidence in the accuracy of the rod i position indication systems. The control rod " full in" and
, " full out" lights, which correspond to the lower electrical j limit and the upper electrical limit respectively, provide an additional means for determining the control rod i positions when the control rods are at either their fully 1 inserted or fully withdrawn pcsitions. i The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Frequency takes into consideration other l information continuously available to the operator in the '
control room, so that during control rod movement, deviations can be detected, and protection can be provided by the control rod deviation alarm.
Palisades Nuclear Plant B 3.1.4-9 02/05/99 s
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Control Rod A11gnnen't B 3.1.4 BASES SURVEILLANCE SR 3.1.4.3~ l REQUIREMENTS
- (continued) Verifying each full-length control rod is trippable would require that each. full-length control rod be tripped. In H0 DES 1 and 2, tripping each full-length control rod would result in radial or axial power tilts, or oscillations.
Therefore, individual full-length control rods are exercised every 92 days to provide increased confidence that all full-length control rods continue to be trippable. l:
even if they are not regularly tripped. A movement of 6 inches is adequate to demonstrate motion without exceeding the alignment limit uhen only one control rod is being moved. The 92 day frequency takes into consideration other information available to the operator in.the control room and other surveillances being performed more frequently, which add to the determination of OPERABILITY i of the control rods. At any time, if a control rod (s) is
-inoperable, a determination of the trippability of the I control rod (s) must be made, and appropriate action taken.
SR 3.1.4.4 l Demonstrating the rod position deviation alarm is OPERABLE verifies the alarm is functional. The 92 day Frequency takes into account other information continuously available to the operator in the control room, so that during control rod movement, deviations can be detected.
Palisades Nuclear Plant 8 3.1.4-10 , - 02/05/99 1
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Centrol Rad' Alignment B 3.l.4
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I BASES SURVEILLANCE, SR~ 3.1.4.5 i i
' REQUIREMENTS (continued) Performance of a CHANNEL CALIBRATION of each control: rod l position indication channel. ensures the channel is 0PERABLE i' and capable of indicating control rod position over the.
entire length of the . control rod's travel. with the exception of the secondary rod' position indicating channel) dead band near the. bottom of travel. This dead band exists. .
because the control rod drive mechanism housing seismic- !
support prevents operation of the reed switches. Since '
this Surveillance must be performed when the reactor is shut down, an 18 month Frequency to be coincident with. ,
refueling outage was selected. Operating experience has .
shown that these components usually pass this Surveillance when performed at a Frequency of once every 18 months.
Furthermore, the Frequency takes into account other.
surveillances being performed at shorter Frequencies, which I determine the OPERABILITY of the control rod position 3 indicating systems.
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SR 3.1.4.6 Verification of full-length control rod drop times determines that the maximum control rod drop time is consistent with the assumed drop time used in that safety analysis (Ref. 2). The 2.5 second acceptance criteria is measured from the time the CRDM clutch is deenergized by the reactor protection system or test switch to 90%
insertion. This time is bounded by that assumed in the safetyanalysis(Ref.2). Measuring drop times prior to reactor criticality, after reactor vessel head reinsta11ation, ensures that reactor internals and CRDMs will not interfere with full-length control rod motion or l )
drop time and that no degradation in these systems has ,
occurred that would adversely affect full-length control j rod motion or drop time. Individual full-length control rods whose drop times are greater than safety analysis assumptions are not OPERABLE. This SR is performed prior to criticality, based on the need to perform this Surveillance under the conditions that apply during a plant !
outage and because of the potential for an unplanned plant 1 transient if the Surveillance were performed with the reactor'at power. ;
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I Palisadss.PaclearPlant 'B 3.1.4-11 02/05/99 :
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Shutdown and Part-Length Rod Group Insertion Limits B 3.1.5 4
B 3.1 REACTIVITY CONTROL SYSTEMS B 3.1.5 Shutdown and Part-Length Rod Group Insertion Limits BASES BACKGROUND The insertion limits of the shutdown rods are initial
. assumptions in all safety analyses that assume full-length l control rod insertion upon reactor trip. The insertion limits directly affect core power distributions and i assumptions of available SDM, ejected rod worth, and initial reactivity insertion rate.
The Palisades Nuclear Plant design criteria (Ref.1) and j 10 CFR 50.46, " Acceptance Criteria for Emergency Core i Cooling Systems for Light Water Nuclear Power Reactors,"
J contain the applicable criteria for these reactivity and power distribution design requirements. Limits on shutdown rod insertion have been established, and all rod positions i are monitored and controlled during power operation to
- ensure that the reactivity limits, ejected rod worth, and SDM limits are preserved.
4 The shutdown rods are arranged into groups that are '
radially symmetric. Therefore, movement of the shutdown rod groups does not introduce radial asymmetries in the i core power distribution. The shutdown and regulating rod groups provide the required reactivity worth for immediate reactor shutdown upon a reactor trip.
The Palisades Nuclear Plant has four part-length control rods installed. The part-length rods are required to remain completely withdrawn during power operation except
< during rod exercising performed in conjunction with SR 3.1.4.3. The part-length rods do not insert on a l reactor trip.
The design calculations are performed with the assumption that the shutdown rod groups are withdrawn prior to the regulating rod groups. The shutdown rods can be fully withdrawn without the core going critical. This provides available negative reactivity for SDM in the event of boration errors. All control rod groups are controlled l manually by the control room operator. During normal plant operation, the shutdown rod groups are fully withdrawn.
The shutdown rod groups must be completely withdrawn from i the core prior to withdrawing any regulating rods during an approach to criticality. The shutdown rod groups are then left in this position until the reactor is shut down.
< Palisades Nuclear Plant B 3.1.5-1 02/05/99 J
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Shutdown and Part-Length Rod Grcup Instrtion Limits B 3.1.5 BASES BACKGROUND They affect core power, burnup distribution, and add (continued) negative reactivity to shut down the reactor upon receipt of a reactor trip signal.
APPLICABLE Accident analysis assumes that the shutdown rod groups are S.AFETY ANALYSES fully withdrawn any time the reactor is critical. This ensures that:
- a. The minimum SDM is maintained; and
- b. The potential effects of a control rod ejection accident are limited to acceptable limits.
Control rods are considered fully withdrawn at 128 inches, since this position places them in an insignificant l reactivity worth region of the integral worth curve for each bank.
On a reactor trip, all full-length control rods (shutdown and regulating), except the most reactive rod, are assumed to insert into the core. The shutdown and regulating rod groups shall be at or above their insertion limits and available to insert the required amount of negative reactivity on a reactor trip signal. The regulating rods may be partially inserted in the core as allowed by LC0 3.1.6, " Regulating Rod Group Position Limits." The shutdown rod group insertion limit is established to ensure that a sufficient amount of negative reactivity is available to shut down the reactor and maintain the required SDM (see LC0 3.1.1, " SHUTDOWN MARGIN (SDM))
following a reactor trip from full power. The combination of regulating rod and shutdown rods (less the most reactive rod, which is assumed to remain fully withdrawn) is sufficient to take the reactor from full power conditions at rated temperature to zero power, and to maintain the required SDM at rated no load temperature (Ref. 2). The shutdown rod group insertion limit also limits the reactivity worth of an ejected shutdown rod.
Palisades Nuclear Plant B 3.1.5-2 02/05/99
i Shutdown and Part-Length Rod Group Insertion Limits B 3.1.5 BASES APPLICABLE The acceptance criteria for addressing shutdown rods as SAFETY ANALYSES well as regulating rodfinsertion limits and inoperability (continued) or misalignment are that:
- a. There be no violation of:
- 1. Specified acceptable fuel design limits, or i
- 2. Primary Coolant System pressure boundary damage; and
- b. The core remains subcritical after accident transients.
As such, the shutdown and part-length rod group insertion limits affect safety analyses involving core reactivity, ejected rod worth, and SDM (Ref. 2). The part-length control rods have the potential to cause power distribution l envelopes to be exceeded if inserted while the reactor is cri tical . Therefore, they must remain withdrawn in accordance with the limits of the LC0 (Ref. 3).
The shutdown and part-length rod group insertion limits satisfy Criterion 2 of 10 CFR 50.36(c)(2).
I LC0 The shutdown and part-length rod groups must be within their insertion limits any time the reactor is critical or approaching criticality. For a control rod group to be ,
considered above its insertion limit, all rods in that group must be above the insertion limit. Maintaining the l shutdown rod groups within their insertion limits ensures that a sufficient amount of negative reactivity is available to shut down the reactor and maintain the required SDM following a reactor trip. Maintaining the part-length rod group within its insertion limit ensures that the power distribution envelope is maintained.
Palisades Nuclear Plant B 3.1.5-3 02/05/99 4
Shutdown and Part-Lcngth Rod Group Insertion Limits B 3.1.5 BASES APPLICABILITY The shutdown and part-length rod groups must be within their insertion limits, with the reactor in MODES 1 and 2.
In MODE 2 the Applicability begins anytime any-regulating rod is withdrawn above 5 inches. This ensures that a sufficient amount of negative reactivity is available to shut down the reactor and maintain the required SDM following a reactor trip. In MODE 4, 5, or 6, the shutdown rod groups are inserted in the core to at least the lower electrical limit and contribute to the SDM. In MODE 3 the shutdown rod groups may be withdrawn in preparation of a reactor startup. Refer to LCO 3.1.1, "SHUTOOWN MARGIN (SDM)," for SDM requirements in MODES 3, 4, and 5.
LC0 3.9.1, " Boron Concentration," ensures adequate SDM in MODE 6.
The Applicability has been modified by a Note indicating )
the LC0 requirement is suspended during SR 3.1.4.3 (rod l l exercise test). Control rod exercising verifies the i freedom of the rods to move, and requires the individual !
shutdown rods to move below the LCO limits for their group. l l Only the full-length rods are required to be tested by l SR 3.1.4.3. The part-length rods may also be moved '
-however, if a part-length rod is moved below the limit of ;
the associated LCO, the Required Actions of Condition A )
must be taken.
Positioning of an individual control rod within its group is addressed by LC0 3.1.4, " Control Rod Alignment."
Palisades Nuclear Plant B 3.1.5-4 02/05/99
E Shutdown"and Part-Length Rod Group-Insertion Limits 1 s
B 3.1.5 BASES ACTIONS L1 :
l Prior to entering this condition, the. shutdown and' .
l part-length rod groups were fully withdrawn. If.a shutdown- 1 rod group is.then inserted into the core, its potential . i negative reactivity is added to the core as it is inserted. '
If one or more shutdown or limits, the affected rod (s)part-length rods inoperable must be declared are not within and the applicable Conditions and Required Actions of.LCO 3.1.4 ,
entered immediately. This Required Action is: based on the; recognition that the shutdown and part-length rods are.
I normally withdrawn beyond their insertion limits and-are
- capable of being moved by their control rod drive i mechanism. Although the requirements of this LCO are not
- applicable during performance of the control rod exercise j test, the inability to' restore a control rod to within the ,
! limits of the LCO following rod exercising would be~
indicative of a problem affecting the'0PERABILITY of the F control rod. Therefore, entering the applicable Conditions and Required Actions of LCO 3.1.4 is appropriate since 'they !
provide.the applicable compensatory measures commensurate H
- with the inoperability of the control rod.' j i
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! When Required Action A.1 cannot be met or completed within
- the required Completion Time, a controlled shutdown should i be coninenced. The allowed Completion Time of 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. is
! reasonable, based on operating experience, for reaching
. MODE 3 from full power conditions in an orderly manner and
- without challenging plant systems.
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. Palisades l Nuclear Plant' B 3.1.5-5' 02/05/99 w4= =r e,wc ' hew se , , trap '
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Shutdown and Part-Length Rod Gr;up Insertion Limits B 3.1.5 BASES SURVEILLANCE SR 3.1.5.1 REQUIREMENTS Verification that the shutdown and part-length rod groups are within their insertion limits prior to an approach to criticality ensures that when the reactor is critical, or being taken critical, the shutdown. rods will be available to shut down the reactor, and the required SDM will be maintained following a reactor trip.: Verification that the part-length rod groups are within their insertion limits ensures that they do not adversely affect power distribution requirements. This SR and Frequency ensure that the shutdown and part-length rod groups are withdrawn before the regulating rods are withdrawn during a plant startup.
Since control rod groups are positioned manually by the l control room operator, verification of shutdown and part-length rod group position at a Frequency of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> j is adequate to ensure that the shutdown and part-length rod groups are within their insertion limits. Also, the 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Frequency takes into account other information available to the operator in the control room for the purpose of monitoring the status of the shutdown and part-length rod groups. !
1 REFERENCES 1. FSAR, Section 5.1
- 2. FSAR, Section 14.2 l l
- 3. FSAR, Section 14.6 l
Palisades Nuclear Plant B 3.1.5-6 02/05/99
Regulating Rod Group Position Limits B 3.1.6 B 3.1 REACTIVITY CONTROL SYSTEMS B 3.1.6 Regulating Rod Group Position Limits BASES BACKGROUND The insertion limits of the regulating rod groups are l initial assumptions in all safety analyses that assume full-length rod insertion upon reactor trip. The insertion l limits directly affect core power distributions, assumptions of available SDM, and initial reactivity insertion rate. The applicable criteria for these reactivity and power distribution design requirements are contained in the Palisades Nuclear Plant design criteria (Ref.1), and 10 CFR 50.46, " Acceptance Criteria for Emergency Core Cooling Systems for Light Water Nuclear
- Power Reactors" (Ref. 2).
Limits on regulating rod group insertion have been established, and all regulating rod group positions are monitored and controlled during power operation to ensure '
that the power distribution and reactivity limits defined by the design power peaking, ejected rod worth, reactivity insertion rate, and SDM limits are preserved.
The regulating rod groups operate with a predetermined amount of position overlap, in order to approximate a ,
linear relation between rod worth and rod position ,
(integral rod worth). The regulating rod groups are withdrawn and operate in a predetermined sequence. The i group sequence and overlap limits are specified in the COLR.
l The regulating rods are used for precise reactivity control i of the reactor. The positions of the regulating rods are i manually controlled. They are capable of addin ]
very quickly (compared to borating or diluting)g reactivity The power density at any point in the core must be limited I to maintain specified acceptable fuel design limits, l including limits that preserve the criteria specified in l 10 CFR 50.46 (Ref. 2). Together, LC0 3.1.6; LC0 3.2.3, j
" QUADRANT POWER TILT (T,)"; and LCO 3.2.4, " AXIAL SHAPE 1 INDEX (ASI)," provide limits on control component operation and on monitored process variables to ensure the core operates within the linear heat rate (LC0 3.2.1, " Linear Heat Rate (LHR)") and radial peaking factor F/ and F/ 1 (LC0 3.2.2, " Radial Peaking Factors) limits in the COLR. l
' l Palisades Nuclear Plant B 3.1.6-1 02/05/99
Regulating Rod Group Position Limits B 3.1.6 BASES APPLICABLE Fuel cladding damage does not occur when the core is SAFETY ANALYSES operated outside these LCOs during normal operation.
(continued) However, fuel cladding damage could result, should an accident occur with simultaneous violation of one or more of these LCOs. Changes in the power distribution can cause increased power peaking and corresponding increased local LHRs.
The SDM requirement is ensured by limiting the regulating and shutdown rod group insertion limits, so that the allowable inserted worth of the rods is such that sufficient reactivity is available to shut down the reactor to hot zero power. SDM assumes the maximum worth rod remains fully withdrawn upon trip (Ref. 4).
The most limiting SDM requirements for Mode 1 and 2 conditions at Beginning of Cycle (BOC) are determined by the requirements of several transients, e.g., Loss of Flow, etc. However, the most limiting SDM requirements for MODES 1 and 2 at End of Cycle (E0C) come from just one transient, Main Steam Line Break (MSLB). The requirements of the MSLB event at E0C for the full power and no load conditions are significantly larger than those of any other event at that time in cycle and, also, considerably larger than the most limiting requirements at BOC.
Although the most limiting SDM requirements at E0C are much larger than those at B0C, the available SDMs obtained via ;
tripping the full-length control rods are substantially l !
larger due to the much lower boron concentration at E0C.
To verify that adequate SDMs are available throughout the i cycle to satisfy the changing require?ents, calculations 1 are performed at both B0C and E0C. it has been determined that calculations at these two times in cycle are sufficient since the difference between available SDMs and the limiting SDM requirements are the smallest at these times in cycle. The measurement of full-length control rod bank worth performed as part of the Startup Testing Program demonstrates that the core has the expected shutdown capability. Consequently, adherence to LC0 3.1.5,
" Shutdown and Part-Length Rod Group Insertion Limits," and LC0 3.1.6 provides assurance that the available SDM at any l time in cycle will exceed the limiting SDM requirements at that time in cycle.
Palisades Nuclear Plant B 3.1.6-4 02/05/99
+ s n - - . -
-=m Regulating Rod Group Position Limits B 3.1.6 BASES-APPLICABLE Operation at the insertion limits or ASI limits may 4 SAFETY ANALYSES approach the maximum allowable linear heat generation rate (continued) or peaking factor, with the allowed T, present. Operation at the insertion limit may also indicate the maximum ejected rod worth could be equal to the limiting value in fuel cycles that have sufficiently high ejected rod worth.
The regulating and shutdown rod insertion limits ensure i that safety analyses assumptions for reactivity _ insertion rate, SDM, ejected rod worth, and power distribution peaking factors are preserved.
l
- The regulating rod group position limits satisfy Criterion 2 of 10 CFR 50.36(c)(2).
LC0 The limits on regulating rod group sequence, overlap, and physical insertion, as defined in the COLR, must be maintained because they serve the function of preserving power distribution, ensuring that the SDM is maintained, ensuring that ejected rod worth is maintained, and ensuring adequate negative reactivity insertion on trip. The overlap between regulating rod groups provides more uniform rates of reactivity insertion and withdrawal and is imposed to maintain acceptable power peaking during regulating rod group motion. For a control rod group to be considered above its insertion limit, all rods in that group must be above the insertion limit.
The Power Dependent Insertion Limit (PDIL) alarm circuit is required to be OPERABLE for notification that the regulating rod groups are outside the required insertion limits. The Control Rod Out Of Sequence (CR00S) alarm circuit is required to be OPERABLE for notification that the rods are not within the required sequence and overlap limits. When the PDIL or the CR005 alarm circuit is inoperable, the verification of rod group positions is increased to ensure improper rod alignment is identified before unacceptable flux distribution occurs. The PDIL and CROOS alarms can be generated by either the synchro based Primary Indication Processor (PIP) node, or the reed switch based Secondary Position Indication (SPI) system since the SPI system, in conjunction with the hest computer, is redundant to the PIP node in the task of control rod measurement, control rod monitoring and limit processing.
Palisades Nuclear Plant B 3.1.6-5 02/05/99
Regulating Rod Group Position Limits B 3.1.6 BASES APPLICABILITY The regulating rod group sequence, overlap, and physical insertion limits shall be maintained with the reactor in MODES 1 and 2. These limits must be maintained, since they preserve the assumed power distribution, ejected rod worth, SDM, and reactivity rate insertion assumptions.
Applicability in MODES 3, 4, and 5 is not required, since neither the power distribution nor ejected rod worth assumptions would be exceeded in these MODES. SDM is preserved in MODES 3, 4, and 5 by adjustments to the soluble boron concentration.
The Applicability has been modified by a Note indicating the LC0 requirement is suspended SR 3.1.4.3 (rod exercise l
, test). Control rod exercising verifies the freedom of the rods to move, and requires the individual regulating rods to move below the LC0 limits which could violate the LC0 l 4 for their group.
ACTIONS a.g.3nd A.2 Optration beyord the insertion limit may result in a loss i of SDM and excessive peaking factors. The insertion limit should not be violated during normal operation; this violation, however, may occur during transients when the operator is manually controlling the regulating rods in j response to changing plant conditions. '
WPen the regulating groups are inserted beyond the I insertion limits, actions must be taken to either withdraw the regulating groups beyond the limits or to reduce THERMAL POWER to less than or equal to that allowed for the actual rod group position limit. Two hours provides a reasonable time to accomplish this, allowing the operator to deal with current plant conditions while limiting !
peaking factors to acceptable levels.
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l Palisades Nuclear Plant B 3.1.6-6 02/05/99 ;
Regulating Rod Group Position Limits B 3.1.6 BASES ACTIONS IL1 (continued)
Operating outside the regulating rod group sequence and overlap limits specified in the COLR may result in excessive peaking factors. If the sequence and overlap limits are exceeded, the regulating rod groups must be restored to within the appropriate sequence and overlap.
Two hours provides adequate time for the operator to restore the regulating rod group to within the appropriate sequence and overlap limits.
L.1 When the PDIL or the CROOS alarm circuit is inoperable, l performing SR 3.1.6.1 once within 15 minutes following any
! rod motion ensures improper rod alignments are identified
! before unacceptable flux distributions occur.
IL.1 When a Required Action cannot be completed within the required Completion Time, a controlled shutdown should be commenced. The allowed Completion Time of 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> is reasonable, based on operating experience, for reaching MODE 3 from full power conditions in an orderly manner and l
without challenging plant systems.
SURVEILLANCE SR 3.1.6.1 REQUIREMENTS With the PDIL alarm circuit OPERABLE, verification of each regulating rod group position every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> is sufficient to detect rod positions that may approach the acceptable limits, and to provide the operator with time to undertake the Required Action (s) should the sequence or insertion limits be found to be exceeded.
The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Frequency also takes into account the indication provided by the PDIL alarm circuit and other information about rod group positions available to the operator in the control room.
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Palisades Nuclear Plant B 3.1.6-7 02/05/99 ]
1
P Regulating Rod Grcup Position Limits B 3.1.6 ,
4
. BASES ,
p SVRVEILLANCE SR 3.1.6.2 l REQUIREMENTS. .
(continued) Demonstrating the PDIL alarm circuit OPERABLE verifies that. :
' the PDIL alarm circuit is functional. The 31 day Frequency takes into. account other Surveillances being performed-at
- shorter Frequencies that identify improper control rod alignments. ;
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{ SR 3.1.6.3 l a
i Demonstrating the CR005 alarm circuit OPERABLE verifies ;
i that the CR005 alarm circuit is' functional. The 31 day ;
l Frequency takes into account other Surveillances being 4
performed at shorter Frequencies that identify improper i control rod alignment. :
a J
l 1 I REFERENCES 1. FSAR, Section 5.1
- 2. 10 CFR 50.46 I 3. FSAR, Section 14.16 i
i 4. FSAR, Section 14.4 ;
- 1 1
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' Palisades Nuclear Plant B 3.1.6-8 02/05/99 4
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-l 3,10 (CONTROL A00 AND POWER O!STiilBUTIGN LIMITS)
Aeolicabiliti .
Applies to operation of CONTROL R005 a hot chahnel factors during I operation 4[
Obiecti To spe ify limits of CONTROL ROO m ement to assure an acceptable power ! l distr button during power operatio , limit worth of individual r es to ;
valv s analyzed for accident cond tions, maintain adequate shut wn {
mar in after a reactor trip and specify acceptable power li ts for ,
l po er tilt conditions.
I
$9 Soecifications s .
3.@.1 Shutdown Marain Recuirements 4 A. I
- a. (Withfourprimarycoolantpumpsinoperationathotshutdownand, k ~ '
] oove',gthe shutdown margin shall be 2%. LA . I A.5
- b. (With less than four_ primary coolant pumps in operatios j-shutdown and_a gyj, horation shall be imediate} Y_ iqltlategjlo [$b.\
increase and aintain the shutdown margin at 23.75%)
_ , , ,l C. fat less than the hot shutdown condition),fith at least one primaryl I(coolant pump in operaT, ion or at seast one shutdown cooling pump in operation, with a flow rate 22810 gom, the boron concentration ;
shall be greater than the cold shutdown boron concentration for O
Liiomal cooldowns and heatups, ie, non emergency conditions. -) '
During non omergenc conditions,atlessthanthehofshutdown condition with no erating primary coolant pumps apd a primary system recirculati g flow rate < 2810 gpm but 2 650' gpm, then within one hour ther:
. 1. (a) Esta ish a shutdown margin of 2 3.5% nd (b) As re two of the three charging pu ps are electrically d abled.
- 2. At east every 15 minutes verify th t no charging pumps are o erating, if one or more chargin pumps are determined to be perating in any 15 minute surve1 lance period, terminate
/ charging pump operation and insu, e that the shutdown margin requirementsaremetandmaintpned.
( ADh [f_O Ilifft:aMUf N
( ADD RAAl n -/ Ab 0 $( req) Amendment No. !!, ti, 57, 50, 70. !!!, 162 October 26, 1994 A
- 3. l. l. j a nc[ 3 50
... ..a fa,w I 0: 2 o t 02/os/99
AlH Q .
- 3. CONTR0! Vl9 AND PfLIFR DISTRIR/ TION ! IM!TS ' kC6c.,trAly sofrol, [k (3.1/' 4) M m li ed nr innnerable r0 Ro! ROD nr Part -t afeth Rnd nir4 h (( j
- a. A CONTROL R00 or a part length rod is Considered misaligned if it LCO 3 l.9 1s out of position from the remainder of the bank by more than 8 inchespJ% rad fm ku,akn alw hlt QMJ
- b. A CONTROL R00 is considered inoperabie if it cannot ce moved by 1 1ts operator or if it Cannot be tripped. A part length rod 15 M Considered inoperable if it is not fully withdrawn from the core and cannot be moved by its operator t li more (non one LUNIHUL
$[ ROD or part length rod becomes misaligned or inoperable. the reactor shall be placed in thedi6t 4nuurwn) Condition within hb (Yl. 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, trioDc. 3
- c. IfaCONTROLR00orabartlengthrodismisaligned hot Channel g g-l factors must promptly .>e shown to be within design ilmits or j :
reactor power shall_be reduced to 75% or lessiot rated power C . 7. w,tn,n two hours . / In adoit,orv snutcown margin ian inciviouas rod wor 11mits must be met. Inolviouai roa wor n calculations will C 51 der the effects of xenon redistribut10i and reduced fuel urnup in the region o the misaligned C0F OL ROD Or part-len rod.
3.10.5 Roan 1at to Genon Insertion iimits
- a. ,e regulating grou equence overlap,and ps shall be 1~ ited to the withdrawallimits specified in the OLR.
insertio
- b. 'With any regulating group ins rted beyond its limit,
- 1. Restore all regulating roups to within insertlo limit within 2__ hours.
4 $ e.e. 3.1/.s) 8,7- ADb AND4l.t/
@- ( Aub RAB)
Q L Ann se 3.th'.Q Q < Anb RA b)
Amendment No. E4. C4, M2,169 July 26, 1995
. 3 52 i Revised 02/05/99 f)4 Ik$ '
- 3. t . 4 4.2 EQUIPMENT SAMPLING AND TESTS Table 4.2.2 Minimus Fr cuencias for Ecuioment T sts FSAR Section est recuency REFERENCE pr,,r% re,4.<c. fir.at.3 4 5+tr (Ar.k re.me el a f St.
50.l.G 1. CONTROL ROOS D'op Times of All lif,efup'in) P'#" kW '7.6.1 full Length Rods SR 3.l.93 2. CONTROL R005 Partial Movement Every 92 Days .6.1.3 of all Rods (Minimus of 6 in) fsd -
- 3. Pressurizer Safety Valves fetPoint / One Each / 4.3.7 3,4.
/ / Refueling /
(p f
~5.7 6
- 4. Main Steam SafetyValves/
/ Set Point
/
/ Five Each Refueling /
/ 4.3.4 r 5. Refueling Sys em functioning Prior to 9.11.4 l % Interlocks Refueling i (3.9 Operations j i
- 6. Service Wa er Functioning Refueling 9.1.2 System Va e (I"Ib)- Actuatio on SISandpS
[5M qsA f
- 7. Primar/ Syr'.es Lenkap Evaluate / Daily 4.7.1 l __
@ 8.Delfed
- 9. Boric Acid Verify proper
[ Daily
[ l h Heat Tracing temperature readings.
- 10. Safety injection Verify / hat level and Each ift Tank Level and pressupe indication 5 Pressure is be een independent 3.5 high igh/ low alams for evel and pressure.
Amendment No. H , M , M 4, Ma, 4 4 , M ;, M e,
'ELEC CHANGES' 4-11
%e N
. Revised 02/05/99
Spec G k. 3,l, 5 3.@ CONTAdLR00ANDpdWERDISTRIB[IONLIMITS. Red,% C, b.)
[3M $hutdown E RodJ imits l* ^$N
" (7]IT5fs u A.As)
Arch L C,0 @. All sh'utdown rocs snall be withdrawn before any regulating .j rods are withdrawn f y g % -,
- b. The shrtdown rods shall not be withdrawn until normal water I level is established in the pressurizer.
Apfpui.,'l:4 [cc o h The shutdown rods shall not be inserted below their exercise 1 limit until all regulating rods are inserted. ,
3 610.7 Low Power /hysics TesA @ N f
l Sections may (3.ls'.1.m: from be deviated M10.1.b]3.10.3.(idQ during b oowe~r 4 ,_
s us nny/ .b,-3./0.5)and MID '- ' 3.10.6 tosti_wa una CPDM lSec3.1%
Ub tuercises/tf necessary to perform a test but on y for the tim _p '
Q :tary to perform the test o -
j g A.8 l nau .
Sufficient CONTROL R shall' be' withdrawn at all time to assure that the reactivity decrea'se from a reactor trip provides adequate shutdown margin. The availa/le worth of withdrawn rods must l'nclude the reactivity defect 'f power and the failure of the withdrawn rod of highest worth to nsert. The requirement for a s tdown margin of 2.0% ,
in reactivity w) h 4 pump operation, and of 3.755 in reactivity with !
less than 4 pupp operation, is consistent with the assumptions used in ,
the analysis df accident conditions (including / steam line break) as I reported in,Aeference 1 and additional analys/s. ;tequiring the boron concentration to be at cold shutdown boron concentration at less than ~h3) '
i 1
hot shutdown assures adequate shutdown margin exists to ensure a return j to power /does not occur if an unanticipated cooldown accident occurs.
This rpuirement applies to normal operat'ing situations and not during emergpcy conditions where it is necess/ry to perfom operations to mitigate the consequences of an accidefit. By imposing a minimum shutdown cooling pump flow rate of 2$10 gpe, sufficient time is provided l foi the operator to terminate a boron dilution under asynsnetric genditions. For operation with n'oftrimary coolant pumps operating and a ;
j/ recirculating flow rate less than/2810 gpa the increased shutdown margin ;
and controls on charging pump operability or alternately the i 4
surveillance of the charging pymps will ensure that the acceptance criteria,'"for an inadvertent heron dilution event will not be l
} l violated. Thechangeininfertionlimitwithreactorpowerinsures l l that the shutdown requiremep'ts for 4 pump operation 1. met at all power j levels. The 2.5 second drpp time specified for the CONTROL R005 is the
- drop time used in the trytsient analysis."' _
! ( Au 4A Al ( L17 M i Amendment No. M H, H, 64. H4, H4. Me,169 July 26, 1995 i 3 53
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- n. ~
02/0s/99-
I ATTACHMENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.1.1, SHUTDOWN MARGIN ADMINISTRATIVE CHANGES (A)
A.1 All reformatting and renumbering are in accordance with NUREG-1432. As a result, the Technical Specifications (TS) should be more readily readable, and therefore l i
understandable by plant operators as well as other users. The reformatting, renumbering, and rewording process involves no technical changes to existing Technical Specifications.
Editorial rewording (either adding or deleting) is made consistent with NUREG-1432.
During Improved Technical Specification (ITS) development certain wording preferences or English language conventions were adopted which resulted in no technical changes (eitner actual or implied) to the TS. Additional information has also been added to more fully describe each subsection. This wording is consistent with NUREG-1432. Since the design is already approved by the NRC, adding more details j does not result in a technical change.
A.2 CTS 3.10.la and 3.10.lb specify requirements for SHUTDOWN MARGIN in terms of f
...at hot shutdown and above." In the proposed ITS, MODE 3 is essentially equivalent to the CTS " HOT SHUTDOWN" as specified in the Discussion of Changes ,
for Section 1.0. The "...and above" portion of the CTS "...at hot shuidown and l above" applies up through the CTS " HOT STANDBY" and " POWER OPERATIONS" j which corresponds to the ITS MODES I and 2. The insertion limit requirements of l CTS 3.10.5, Shutdown Rod Limits, and CTS 3.10.6, Regulating Group Insertion l Limits, ensure that adequate SHUTDOWN MARGIN exists when the plant is at power. l Therefore, in the proposed ITS, the required amount of SHUTDOWN MARGIN in )
MODES 1 and 2 is verified through the Palisades ITS 3.1.5, Shutdown and Part-Length l Rod Group Insertion' Limits, and 3.1.6, Regulating Rod Group Position Limits. Smce l the requirements of the CTS are maintained and only restructured to meet the ITS l format, these changes are considered.to be administrative changes. These changes are consistent with NUREG-1432.
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L i ATTACHMENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.1.1, SHUTDOWN MARGIN i
j A.3 CTS 3.10.lc specifies SHUTDOWN MARGIN requirements at "less than the hot shutdown condition" (below 525 F). In the proposed ITS this corresponds to MODE 3
< 525 F, MODE 4, and MODE 5. The requirements for the refueling condition j (MODE 6) are addressed in proposed ITS 3.9.1. This is an administrative change to
- reflect the NUREG-1432 defined MODES. This change is consistent with the intent of
I i A.4 CTS 3.10.lc includes the statement "...with at least one primary cociant pump in i operation or at least one shutdown cooling pump in operation, wiin a flow rate
] a 2810 gpm, the boron concentration shall be greater than the cold shutdown boron l concentration." In the proposed ITS for operation with Tave < 525 F, SHUTDOWN
! MARGIN (SDM) will be within the limits specified in the COLR regardless of the l
- primary system flow rate and throughout the temperature range as a cooldown occurs. l
, Overall, this is considered to be an administrative change since the " cold shutdown l i boron concentration" requirement is replaced by the requirement to have SDM within l the limits specified in the COLR throughout the temperature range. This change could j be more or les< r strictive depending on a particular primary coolant temperature j evaluated, however, overall the requirement is considered an administrative
. " substitution" of one requirement for another while still preserving the SDM requirements. 1 i
! A.5 CTS 3.10.lb states in part that "...boration shall be immediately initiated to increase and main'ain the shutdown margin at...." In the proposed ITS this statement becomes
]
Action A :nd the term "immediately" is changed to 15 minutes. In the proposed NUREG-1432, the time frame of 15 minutes is used in lieu of "immediately" to specify a specific time in which an action must be started. The terminology conveys the same
- meaning in the CTS in that quick action must be taken. In NUREG-1432, a
{ Completion Time of "immediately" is defined in Section 1.3 as " pursue continuously in a controlled manner without delay." .Therefore, while a Completion Time of i "15 minutes" is used in the proposed ITS as compared to the CTS "Immediately" the
! cffective meaning is the same. Therefore, this is considered to be an Administrative Change. This change is consistent with NUREG 1432.
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- l ATTACHMENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.1.1, SHUTDOWN MARGIN A.6 CTS 3.10.la, CTS 3.1.10.lb and CTS 3.1.10c contain the requirements for l i
SHUTDOWN MARGIN. The amount of required SHUTDOWN MARGIN is dependent on the plant operating conditions (e.g., above or below hot shutdown) and the number of primary coolant pumps in operation. To establish consistency with the i format and style of the ITS, the values of the required SHUTDOWN MARGIN have been moved to the COLR including the plant specific operating conditions and pump l l configurations (See DOC LA.1) A new LCO statement has been added which states l that the SHUTDOWN MARGIN must be within the limits specified in the COLR, and l ,
an Applicability of MODES 3,4, and 5 stipulated. These changes do not alter the actual CTS requirement for SHUTDOWN MARGIN, nor do they impose any additional requirements. These changes merely present the same infonnation in a different format necessary to convert to the ITS. As such, these changes are considered administrative in nature.
I MORE RESTRICTIVE CHANGES (M)
M.1 CTS 3.10.la specifies "With four primary coolant pumps in operation at hot shutdown and above, the shutdown margin shall be 2%." However there is no action specified in the CTS if the shutdown margin is found to be less than 2% and so the plant would have to enter LCO 3.0.3. In the proposed ITS, if the SHUTDOWN MARGIN is found to be below the limit, boration must be initiated within 15 minutes. This is similar to the restoration action specified in CTS 3.10.lb which specifies if shutdown margin is below the required amount that "boration shall be immediately initiated to increase and maintain the shutdown margin." Since in the CTS, LCO 3.0.3 would be have to be i entered if the SHUTDOWN MARGIN was found to be below the 2% limit, the 15 minutes to initiate boration is considered to be a more restrictive change. Initiating boration to restore the required amount of SHUTDOWN MARGIN is the appropriate action to take in this situation to return the plant to a safe condition. Furthermore, CTS 3.10.lc does not specify actions to take if flow is a 2810 and the shutdown margin requirements (boron concentration greater than the cold shutdown boron concentration) have not been met. Therefore, if the SHUTDOWN MARGIN was not met, and the plant was above the CTS Cold Shutdown (210 F) then the plant would have to be shutdown in accordance with LCO 3.0.3. In the proposed ITS, ACTION A requires that if the SHUTDOWN MARGIN (SDM) requirement is not within limit, then baration must be initiated within 15 minutes to restore SDM to within limit. Therefore, since the proposed ITS requires that action be taken with 15 minutes, it is considered to be a more restrictive action. This change is consistent with NUREG-1432.
Palisades Nuclear Plant Page 3 of 5 02/05/99
A'ITACHMENT 3
! DISCUSSION OF CHANGES :
l SPECIFICATION 3.1.1, SHUTDOWN MARGIN M.2 The Palisades Nuclear Plant CTS does not contain an explicit surveillance requirement
! for SHUTDOWN MARGIN even though there was a requirement that the limits be met as specified in 3.10.1. Proposed ITS 3.1.1 adds SR 3.1.1.1 to verify SHUTDOWN l MARGIN "every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />." Since the requirement to verify SHUTDOWN MARGIN was not explicitly required in the CTS, the addition of the proposed Frequency is considered a "more restrictive" change. This change is consistent with NUREG-1432.
M.3 CTS 3.10.7 includes an exception which allows a deviation from the requirement for shutdown margin during performance of CRDM exercises. Proposed ITS 3.1.1 does not contain this same exception since violation of the LCO is not expected during the performance of the control rod drive exercise surveillance (SR 3.1.4.4). During the performance of SR 3.1.4.4, control rods will be exercised between 6 inches and 8 inches. The change in reactivity as a result of this movement is small due to the relative worth of the control rods which is largely determined by their position in the core at the time this SR is performed. This small change in reactivity is not enough to cause a violation of the Shutdown Margin requirements ofITS 3.1.1. Thus, reliance on the exception contained in CTS 3.10.7 is not needed. This change is consistent with NUREG-1432.
RESTRICTIVE CHANGES - REMOVAL OF DETAILS TO LICENSEE CONTROLLED DOCUMENTS (LA)
LA.1 CTS 3.10.1 contains the requirements for Shutdown Margin including specific values l based on plant conditions and configuration. This proposed change relocates the values l for Shutdown Margin to the COLR in order to provide core design and operational l flexibility that can be used for improved fuel management and to solve plant specific l issues. Placing the Shutdown Margin values in the COLR allows the core design to be l finalized after shutdown when the actual end of cycle burnup is known. This would l save redesign efforts if the actual burnup differs from the projected value. Current l reload design efforts and the resolution of plant specific issues are restricted by the l guidelines to not change the Shutdown Margin since it would result in a License l Amendment Request. Although the actual value of Shutdown Margin is not derived l through calculations, it is assumed to be an initial input in the plant safety analyses. As l such, a change in Shutdown Margin must be evaluated for its impact on the safety l analyses to determine if the revised value results in an unreviewed safety question. l Placing the Shutdown Margin limits in the COLR does not result in a significant impact l on plant safety since changes to the safety analyses (including a change in Shutdown l Margin limits) are done in accordance with NRC approved methodologies. l Palisades Nuclear Plant Page 4 of 5 02/05/99
1 ATTACHMENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.1.1, SHUTDOWN MARGIN
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LESS RESTRICTIVE CHANGES (L) i There were no "Less Restrictive" changes associated with this specification. l h
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' ATTACHMENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.1.4, CONTROL ROD ALIGNMENT M.3 If the rod position deviation alarm is inoperable, Condition B of the proposed ITS requires that SR 3.1.4.1 (rod position verification) be performed within 15 minutes of movement of any control rod. This action ensures that the rods are maintained within their alignment limits and is consistent with other action times in the Palisades CTS for verifying rod position indication such as when a channel of rod position indication is lost. The addition of this requirement is considered a more restrictive change since the CTS does not address requirements for the rod position deviation alarm. This change is consistent with NUREG-1432 with the exception of the Completion Time which is consistent for other CTS Completion Times for performing the rod position verification.
M.4 The proposed ITS includes SR 3.1.4.4 which verifies that the rod position deviation l alarm is OPERABLE every 18 months. This surveillance frequency is adequate for l ensuring that'the rod position deviation alarm remains OPERABLE given the other indications available to the operator of rod position to detect if a deviation has occurred. The addition of this requirement is considered a more restrictive change since the CTS does not address requirements for the rod position deviation alarm. l M.5 CTS 3.10.1d (" Shutdown Margin Requirements") states if a control rod cannot be tripped, shutdown margin shall be increased by boration as necessary to compensate for the worth of the withdrawn inoperable control rod. In addition, CTS 3.10.4b
(" Misaligned or Inoperable Control Rod or Part Length Rod") states, in part, that if i more than one control rod becomes inoperable, the reactor shall be placed in the hot !
shutdown condition within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. The intent of these two CTS requirements is to l provide compensatory measures to allow continued plant operations with one inoperable (untrippable) control rod, and to provide the necessary required actions when more than one control rod is inoperable. Although the CTS allows unrestricted operations with an untrippable control rod, this allowance is inconsistent with the assumptions used in the safety analysis. Therefore, ITS 3.1.4 has been proposed to place the plant in Mode 3 whenever one or more control rods are inoperable for reasons other than a single control rod being immovable. This change is consistent with NUREG-1432.
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ATTACHMENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.1.5, SHUTDOWN AND PART-LENGTH ROD GROUP INSERTION LIMITS ADMINISTRATIVE CHANGES (A)
A.1 All reformatting and renumbering are in accordance with NUREG-1432. As a result, the Technical Specifications (TS) should be more readily readable, and therefore understandable by plant operators as well as other users. The reformatting, renumbering, and rewording process involves no technical changes to existing Technical Specifications.
Editorial rewording (either adding or deleting) is made consistent with NUREG-1432.
During Improved Technical Specification (ITS) development certain wording preferences or English language conventions were adopted which resulted in no technical changes (either actual or implied) to the TS. Additional information has also been added to more fully describe each subsection. This wording is consistent with NUREG-1432. Since the design is already approved by the NRC, adding more details does not result in a technical change.
A.2 The Bases of the current Technical Specifications for this section have been completely ,
replaced by the revised Bases that reflect the format and applicable content consistent I with NUREG-1432. The revised Bases are shown in the proposed Technical SpeciGcation Bases..
A.3 CTS 3.10.3, Part-length Control Rods, specifies that "The part-length control rods will be completely withdrawn from the core..." In the proposed ITS, the part-length control rods are required to be 2128 inches as opposed to " completely withdrawn."
Requiring the part-length rods to be withdrawn 2128 inches has the same effect as completely withdrawn in that the rods are removed from the active region of the core.
This is consistent with NUREG-1432 in that the requirement for rods to be withdrawn is specified in terms of inches withdrawn. This is considered to be an administrative change.
A.4 CTS 3.10.3 specifies that the part-length controls will be completely withdrawn from the core "(except for the control rod exercises and physics test)." The exception for control rod exercises is addressed as part of the Applicability Note. The physics tests exceptions are no longer needed because the part-length rods are not required to be moved during PHYSICS TESTS. These changes are considered to be administrative changes since no requirements have changed. These changes maintain consistency with NUREG-1432.
Palisades Nuclear Plant Page I of 4 02/05/99
ATTACHMENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.1.5, SHUTDOWN AND PART-LENGTH ROD GROUP INSERTION LIMITS A.5 CTS Table 4.17.6 Item 2 requires that the Rod Position Indication have a CHANNEL
- CHECK performed every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. This requirement becomes SR 3.1.5.1 in the proposed ITS. Proposed SR 3.1.5.1 requires " Verify each shutdown and part length rod is withdrawn 2128 inches every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />." The surveillance in the proposed ITS functions to perform the same verifications as that intended in the CTS " CHANNEL CHECK" since the CTS definition of " CHANNEL CHECK" includes the statement "A j CHANNEL CHECK shall include verification that the monitored parameter is within the limits imposed by the Technical Specifications." CTS 3.10.6 requires that the shutdown rods shall be withdrawn before any regulating rods are withdrawn.
CTS 3.10.4b in part states that a part-length rod is considered inoperable if it is not l fully withdrawn. CTS 3.10.3 requires that the part-length rods be completely withdrawn. Therefore, the proposed surveillance performs this by proposed ITS SR 3.1.5.1 ensuring that the shutdown and part-length rods are withdrawn 2128 inches. This is considered to be an administrative change since the requirements have not changed but have been reformatted in accordance with NUREG-1432.
A.6 CTS 3.10.6a states "All shutdown rods shall be withdrawn before any regulating rods i are withdrawn." In the proposed ITS, the phrase "above 5 inches" is added to clarify what is intended by " withdrawn." Allowing the regulating rods to be withdrawn up to 5 inches facilities normal operation of the control rod drive motors which are
" bumped" to bring the rods off the bottom before they are withdrawn. This area of the core is very insignificant with respect to the integral worth of the rod. This also corresponds to the Shutdown Rod Insertion interlock which prevents the shutdown rods from being inserted once the regulating rods are withdrawn greater than 5 inches. This change is a clarification to define what " withdrawn" means with respect to the regulating rods.
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ATTACHMEST 3 DISCUSSION OF CHANGES SPECIFICATION 3.1.5, SHUTDOWN AND PART-LENGTH ROD GROUP INSERTION LIMITS A.7 CTS 3.10.6a states " All shutdown rods shall be withdrawn before any regulating rods are withdrawn." CTS 3.10.6c states "The shutdown rods shall not be inserted below their exercise limit until all regulating rods are inserted." The proposed ITS 3.1.5 LCO states "All shutdown and part/ length rod groups shall be withdrawn to a 128 inches." The Applicability for LCO 3.1.5 is MODE 1, MODE 2 with any regulating rod withdrawn above 5 inches. The proposed ITS wording for the LCO and Applicability is equivalent to the CTS wording in 3.10.6b. In the ITS, the shutdown i rods must be withdrawn 2128 inches by the LCO before the regulating rods are
- withdrawn above 5 inches (see DOC A.6 for discussion on 5 inches criteria). In addition, the CTS 3.10.6c requiremem that the shutdown rods cannot be inserted below their exercise limit is also maintained in the ITS. This is because the shutdown rods cannot be inserted, except for rod exercising allowed by Applicability note, until out of j the MODE of Applicability which required the regulating rods to be s 5 inches ,
withdrawn. Therefore, the CTS and the proposed ITS are equivalent. I A.8 CTS 3.10.7 includes an exception which allows a deviation from the requirement for shutdown rod limits during performance of CRDM exercises. The exception contains a l i qualifying statement which reads "if necessary to perform a test but only for the time necessary to perform the test." The Applicability Note for proposed ITS 3.1.5 which also provides an exception from the requirement for shutdown rod limits during performance of CRDM exercise does not contain this same qualifier since these type details are governed by the usage rules for the ITS. Therefore, deletion of this information is considered administrative in nature. This change is consistent with NUREG-1432.
a Palisades Nuclear Plant Page 3 of 4 02/05/99
ATTACHMENT 3 DISCUSSION OF CHANGFE SPECIFICATION 3.1.5, SHUTDOWN AND PART-LENGTH ROD GROUP INSERTION LIMITS A.9 CTS 3.10.3 and CTS 3.10.6 stipulate the requirement for rod position on an individual l l rod basis (i.e., all shutdown and part-length rod must be fully withdrawn). In addition, j l l
CTS 3 A.10.4 ccquires that a control rod must be aligned within 8 inches from the l remainder of the bank. The CTS does not specify rod positions on a group basis, and l does not contain actions when controls rods are misaligned from their groups by less l ;
than 8 inches. Proposed IT.', 3.1.5 establishes insertion limits for the shutdown and l 1 part-length rod groups by requiring them to be withdrawn 2128 inches. Required l Action A.1 of ITS 3.1.5 requires that any shutdown or part-length rod group that is not l 1
within its group insertion limb ha telared inoperable and the Conditions of ITS 3.1.4 l 1 entered immediately. If the Required Action and associated Completion Time are not l 1 met, Required Action B.1 requires the plant to be in Mode 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. To ensure l )
compliance with the requirements of LCO 3.1.5, for a control rod group to be l considered above its insertion limit, all rods in that group must be above the insertion l limit. The addition ofITS Required Actions A.1 and B.1 is characterized as an l administrative change since the action taken when a shutdown or part-length rod exceed l its insertion limit is consistent with the CTS actions for an inoperable control rod. l l
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MORE RESTRICTIVE CHANGES (M) :
J There were no "More Restri:tive" changes associated with this specification. l LESS RESTRICTIVE CHANGES - REMOVAL OF DETAILS TO LICENSEE CONTROLLED DOCUMENTS (LA)
LA.1 CTS 3.10.6b states "The shutdown rods shall not be withdrawn until normal water level is established in the pressurizer " This requirement was included to help assure an inadvertent criticality will not occur with the PCS water solid. This statement is more appropriate for being addressed in plant procedures and is not included in the proposed ITS. Changes to plant procedures are made in accordance with the plant procedure change process. This change maintains consistency with NUREG-1432.
LESS RESTRICTIVE CHANGES (L)
There were no "Less Restrictive" changes associated with this specification.
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ATTACHMENT 4 NO SIGNIFICANT HAZARDS CONSIDERATION SPECIFICATION 3.1.1, SHUTDOWN MARGIN '
i LESS RESTRICTIVE CHANGES L.1 There were no "Less Restrictive" changes associated with this Specification. l a
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lh SDMET] > 3dO'F fAnalog) /
J.4.4 3.1 REACTIVITY CONTROL SYSTEMS
, ' 3.1.1 SHUTDOWN MARGIN (SDH k , p 20f'F (A plogy
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LCO 3.1.1 SDM shall bed (5)/AkM L.oMin tend 5 psdidcd trs M O C DL.f-.
f .
g g ,g APPLICABILITY: MODES 3, @ 4j a,nd $
i ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. SDM not within limit. A.1 Initiate boration to 15 minutes restore SDM to within limit.
- \TET F- 9 '
SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.1.1 Verify SDM(i,/ /[WAl(MO4o bc. 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />s- l oiikm l[mM .
i
@ i;
)
@ l CEOG STS 3.1-1 Rev 1.-04/07/95 !
I Revised 02/05/99
^
l U*@
(onir4kJ hAlignment '* (Anl 3.1 REACTIVITYC0glSYSTEMS 3.1 Control IElednt Assedly (CEA)) Alignment OrglogD b+rs, re ds, #6clJ. br /6k andwoAn dends, LCO 3.1 A11 (c'ffsYshall e OPERABLE and aligned to within inches
' f(indvtated eositieni of their respective L g 3 ,3 qq (CEF motioh inhtbitiand f the Q deviation gei geu ),shall be QIs.em q
OPERABLE]
", of chu febs'tn .
, Dcv Atrn. AkW APPLICABILITY: MODES 1 and 2.
C6 3R6kTIONS A *
@ CONDITION REQUIRED ACTION COMPLETION TIME ptLR] A (conhot r2 C. Z. Z i
c@. OJnor mo@ regu tffs triob ible ing
@ Reduce THERHAL POWER to 5 (G RTP.
(Dhou@
7 misaligned from tt Tc;%
- group [_by >>ThinchesP '@,1 e and 5 15 inche .
W 4.2.1 erify SON is I hour itTH/7 E (4.5]% Ak/k.
On regulatin CEA M tr ppable an m saligned f om its A.2. Initiate bora on to Ih r roup by restore SDM t within
[15 inch s]. limit.
gp .3.1 Restore t e 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> gg misalign CEA(s) to I
within inches]
(indica ed position) fgTf.lyj}
of its group.
MM dl(J \,.L/ g (continued)
C..l fcf4tno SR i?. Z. I (pr. dog ftrder LMctl J. Lour.s h oR .
I CEOG STS 3.1-8 Rev 1, 04/07/95 Revised 02/0s/99
SECTION 3.1 v
, INSERT 3 i
COMPLETION CONDITION. REQUIRED ACTION TIME 1
E.- ......... E .1 . . . . . . . . . . . ..........
DE Both rod position ,
indication channels inoperable for one or more control rods. ,
INSERT 4 1
Perform a CHANNEL CHECK of the control rod position indication channels. l l l
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3.1-11 Revised 02/05/99
- h. bol bd) @ Alignment RValfg3 3.149 h
i SURVEILLANCE REQUIREMENTS (continued)
SURVEILLANCE FREQUENCY
{TsfF12.7 / u SR 3.1 e deviation ccicetrTU is d P'5 6*A
% / @rylondi h
\ y3 E.a .1 .aj
'Ta\ < H.L1 SR 3.1 @ @ Ver_ify ffh freedom of movement \Os (trro aMI1uI by moving each individual l@
.2 days
~2 ggg.)44 gM$ that is liot fully inserted into the l@
reactor core nches)ineither }@
g Q direction, g yg h ca A,NEL C A U SMTtoeJ f Mk 4ID SR 3.1.$.$ Perform alCHAYfEL FUNCTif)NAL TESTlof the 18 months ll TW ~ t t reed s.w)tch) position)trafismitt4rl channey.
T' % holria{} lI d ubo d
.y 3,a y f.ed<.1c.cd 9
W' *1 SR 3.1.$.$@ (~Oleach Verif JewnQ & drop time is Prior to reactor D<~ i s' , 7 seconds, g h criticality, Lfter each rus,,gde/***^oa o of the reactor head i
CEOG STS 3.1-12 Rev 1, 04/07/95 Revised
. 02/0s/99
Shutdotn Insertion Limits 1(M 757F. r3(o / g h edPuM43A LJ C,ry 5 )
3.1 REACTIVITY CONTROL SYSTEMS 3.1[ ShutdownlConJ/rol Elemed Assembly /CEA)l Insertion Limits handh.,tL.er$& %J &h} @ ,
LCO 3.1 @g All shutdown @shall be withdrawn to 2 MVinches.
4 ea,t.le SA rod va-P3
- '^
APPLICABILITY: MODE 1, MODE 2withanyregulatingICEAnatfully16sertedl.2 h 4
............................N0TE------------------
I. .I.$$.$.$!.$. .$.!! $ .!!. . '. .. .......
(ud emcise tes+)
ACTIONS __
CON 0! TION REQUIRED ACTION COMPLETION TIME A. One or more shutdown A.l.1 Ve ify SDH 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> g not within limit. _
2 (4.5]% Ak/k.
h jorput.).y & cd y..e] s g ,
A.l.2 Initiate boration to I hour h7[*bl j o e SOM to w thin
, \
A.2 Restor shutdown 2 ho s CEA(s to within !
limi . j i I l
l B. Required Action and 8.1 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> l associated Comoletioq Time not met. )
)
Dec3o rt. a f-lech d comb o! In.nd M1 rod (d inopero.bl<. ctnd f n44 r -the ypLt to h kf-Condihon s ud 8ep"d A
\ chons oJ LC O 3. l. 4. 3 CEOG STS 3.1-13 Rev 1, 04/07/95 I
l
= Ilevised 02/05/99
T5TF. l% @ F egulating K4Afinsp'rtfon) Limits Wal(qh k
Kocl Cx o dp % % N\
3.1 REACTIVITY CONTROL SYSTEMS cy g 5 3.1 cvs r.u, ) o b rw a 67 Regulating [Contr/l Elemey/t Assembly /tCEA) Insertion} Limits WalpQ thed Group bit %M h
LCO 3.1 The jBwer 4 pendent 25sertion 8Imit (P0ll alarm circuit l h,,
shall be OPERABLF., and the regt.lating groups shall be l 3 2
- bd'*I N limited to the withdr'awal sequerce and insertion limits specified in the COLR.
, OJ Cf Sepw (f Reos ,
jbm c.; , mlt-b0A,@ P APPLICA8ILITY: MODES I and 2. g
............................N0TE...................U*E......
ThisLCOisnotapplicablewhileperformingSR3.1$$'
kjuring/ reactor tower cutback operation]J.
ACTIONS CON 0! TION REQUIRED ACTION COMPLETION TIME l ~
cn 3 m s b.l h A. h Regulating"@ groups A.I.1 Ver y SDM 1 ur inserted beyond the 2 4.5]% ak/k.
g ttrAnti4nn insertion limit. Q3 A.I.2 Initiate boration to I hour
[qW.6') /3 restore SDM to with limit.
A. tore regulating 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> groups to within imits.
s.- QE (continued)
CE0G STS 3.1 15 Rev 1, 04/07/95 Revised o 02/05/99
~
l SDMi-TbE/F (Agalgd)N B 3.1.1 BASES (continued) 1 APPLICABLE The minimum required SDM is assumed as an initial condition l SAFETY ANALYSES in safety analysis. The safety analysis (Ref. 2) establishes an SDM that ensures specified acceptable fuel DM M'H" design limits are not exce.eded for normal operation and A00s, with the assumptionfof the highest worth CEA stuck out) h5 rf' %- >
Lfolpwingafeactortri).j9 7
The acceptance criteria for the SDM requirements are that i
f_f\)[j specified acceptable fuel design limits are maintained. This
! is done by ensuring that: -
- a. The reactor ca'i be made subcritical from all operating conditions, transients, and Design Basis Events; j b. The reactivity transients associated with postulated j accident conditions are controllable within acceptable limits (departure from nucleate boiling ratio (DNBR),
fuel centerline temperature limit A00s and
, s 280 cal /gm energy deposition for thef, f;5 ejection lh accident); and go.a . . \ <.4)
- c. The reactor will be maintained sufficiently J suberitical to preclude inadvertent criticality in the i shutdown condition. <
i The most limiting accident for the SDM requirements are pa nF 4 based on a main steam line break (MSLB), as described in the
] accident analysis (Ref. 2). The increased steam flow l 1 resulting from a pipe break in the main steam system causes j
! an increased energy removal from the affe ed fleam l f) i g' ginerator(SG),andconsequentlythe$5. This results$(fn a redugtign_qf_the 5 HcW b coolant tempera ure. The resultant
{^ ~ -Q coolant shrinkagi causes a reduction in pressure. In the presence of a negative moderator temperature coefficient. ,
th_is cooldown causes an increase in core reactivity f As } S] { J J emperpure decreasy, Ine severify of an Rdul oecreasew >
nti,1.zt_he MODE 5 ve ue is reached f Tht.most imiting M5L !
with respect to potential fuel damage (berfope a reactorin)tf_B.
(BTeu'iO u is a guillotine break of a main steam lineOtdidp )
- g. p t $ 1 d (1girttai#mesOinitiated at the end of core life. The positive l Teactivity addition from the moderator temperature decrease l will terminate when the affected SG boils dry, thus I terminatinggKS heat removal and cooldown. Following the MSLB a post tri n j
[ fuel,damageocc_u_preturntopowermayoccur;however,jnors l L2) af'a; 1
(continued)
CEDG STS B 3.1-2 Rev 1, 04/07/95 l
l
= n.vt .d I 02/05/99
SDK-T,/; > 20p"F (6(al06b 5 3.1.1 1
BASES APPLICABLE fpger, Adj THERMAL POWER does not violate the Safety Limit b SAFETY ANALYSES (SL) requirement of SL 2.1.1.
(continued) gig (f) In addition to the limiting MSLB transient, the SOM requirementp ust also protect against0 o n g q a h g Inadvertent boran dilution:
aa,.o.m e (Of, $h d { h ,
"- @ uncontrolled GWWithdrawal fromgf subcritical @
oowerl condition ( gd s),
I
Each of these events is discussed below.
In the baron dilution analysis, the required SDM defines the reactivity difference between an initial subcritical boron concentration and the corresponding critical baron concentration. Thesg values, in conjunction with the configuration of ther$5 and the assumed dilution flow rate, directly affect the results of the analysis. This event is most limiting at the beginning of core life when critical boron concentrations are biohest.
M The withdraw m uberitical br /fow pose conditions adds reactivity to the reactor core, tausing both the core power level and heat flux to increase with corresponding increases in reactor coolant temperatures and pressure. The withdrawal of also produces a time dependentredistributionofc@ ore} power. -
9 e c t rod 6 <T)
Depending on the system initial onditions ana reactivity insertion rate, the uncontrolle ($ withdrawal transient is terminated by either a high power Trip or 3 high pressurizer pressure trip. In all cases, power level,RRES pressure, linear heat rate, and the DNBR do not exceed allowable limits.
'The artup of an inact ye RCP will not result in " cold l Iwat " criticality, ev n if the maximum differenc in *
\teperatureexistsbeweentheSGandthecore, he maximum
'p sitive reactivity ddition that can occur due .o an nadvertent RCP st t is less than half the mi imum requiredj (continued)
CEOG STS B 3.1-3 Rev 1, 04/07/95
. n.wis.d 02/05/99
U Sg WF (M BASES -
APPLICABLE SDM. An idl/ RCP cannot, therefofe, produce a returr/to '
SAFETY ANALYSES power from the hot standby condition. /
(continued)
SDM satisfies Criterto'n 2 of lth/NRC Pol)(y Statq4ent[. I hO C,FR So.%(c,72)l LCO The MSLB (Ref. 2) and the boron dilution (Ref. 3) idents /
are the most limiting analyses that establish th SDM v lue r of the LCO. For MSLB accidents, if the LCO is violated, there is a potential to exceed the DNBR limit and to exceed 10 CFR 100, " Reactor Site Criteria," limits (Ref. 4). For the boron dilution accident, if the LCO is violated, then the minimum required time assumed for operator action to terminate dilution.may no longer be applicable.
G +<.1 n d)
SDM is a e ' physics design condition that can be ensured through through positioning (regulating and shutdown e soluble boron concentration.
and h
APPLICABILITY In MODES 3 @ @ the SOM requirements are applicable to -
provide sufficient negative reactivity to meet the assumptions of the safety analyses discussed above. In MODES I and 2, SDM is ensured by complying w~ith LCO 3.1.M hutTo@Ccontrdl Element' Assembly KEA) ,iiser,Ti.on_ Lir!Qtid l h bTF-O[h and LCO 3.1.fftJIf the sertion limits of LCO 3.1.6 or ICO 3.1 7 are not bein complied with, 5 is not autom ically violate . The SDM must b calculated by
[757F-G{@ perfo ing a reactiv y balance calcul ion (consider ng the, llis d reactivity e ects in Bases Se ion SR 3.1.1. ). In
'HO 5, SDM is add essed by LC0_3.1.
~~
"3HUTDOWN ~ IN
( )-T m <200*.flnA0TE6,the'shutdownreactivity i requirements are given in LCO 3.9.1, " Boron Concentration."
ACT!0ftS AJ If the SDM requirements are not met, boration must be
- initiated promptly. A Completion Time of 15 minutes is adequate for an operator to correctly align and start the required systems and components. It is assumed that boration will be continued until the SDM requirements are i met.
(continued)
CEOG STS B 3.1-4 Rev 1, 04/07/95 4
- ...,... l 02/05/99 1
SOM{p snder or i BASES -
ACTIONS Ad (continued)
In the determination of the required combination of boration flow rate and boron concentration, there is no unique requirement that must be satisfied. Since it is imperative
'
- b) m to raise the boron concentration poTsibl'eTthTboronicongtntratloft should of be the ($CS as a highly h as h soon concentrated solution, such as that normally found in the
@6c,4 4bboric acid storage tankfor the' boratedhater sto(age tank.h The operator should borate with the best source available for the plant conditions.
In determining the boration flow rate, the time ore life l@
- must be considered. For instance, the most difficult time in core life to increase thetBS boron concentration is at the beginning of cycle, when the boron concentration may aptroach or exceed 2000 ppm. Assuming that a value of
@-~{IX MB must be recovered and a boration flow ratel@ of35} g m, of thepRES by 100_ ppm in approximatel/D6 minutes. If a f@ *-
It E-9 ogj'rrrr!'T borin7artE ofTfp acm/pp!!i is assumed, t .mbination of parameters will' increase the SDM by 1% g,J These boratien I parameters of {g ' gpm and {L.t-ppm represent typical values I and are provided for the purpose of offering a specific example.
I SURVEILLANCE SR 3.1.1.1 REQUIREMENTS SDM is verified by ;:-f;=S; a reactivity balance calculation, considering the listed reacttvity effects:
a.(2)(8tS boron concentration; l@
- b. positions; {h c.blCSaveragetemperature;{'
J
- d. Fuel burnup based on gross thermal energy generation; I
- e. Xenon concentration; l
If. / Samarium foncentratiof; land lh hf. Isothermal temperature ,c#oefficient (!TC). (h (continued)
CE06 STS B 3.1-5 Rev 1, 04/07/95 l
l l
1 l
l
- Revised l I
02/05/99
. . . . . _ _ _- ~ _ _ . _ . . _ . . ___ _ .__
SECTION 3.1 INSERT 1 Samarium is not considered in the reactivity analysis since the analysis assumes that the negative reactivity due to samarium is offset by the positive reactivity of plutonium build in.
INSERT 2 l SR 3.1.1.1 requires SDM to be within the limits provided in the COLR. This SDM value ensures l l the consequences of an MSLB will be acceptable as a result of a cooldown of the PCS which adds positive reactivity in the presence of a negative moderator temperature coefficient, as weli as the other events described in the Applicable Safety Analysis. As such, the requirements of this SR must be met whenever the plant is in MODES 3,4, and 5. l i
l i
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i 1
1 4
l i
i I
i B 3.1-6 i i
llevised 02/05/99
i 1
SECTION 3.1 I I
INSERT The synchro based position indication system measures the phase angle of a synchro geared to l the CRDM rack. Full control rod travel corresponds to less than 1 turn of the synchro. Each l s control rod has its own synchro. The Primary Information Processor (PIP) node scans and l l 1 converts synchro outputs into inches of control rod withdrawal. The resolution of this system .
is approximately 0.5 inches. Each synchro also has cam operated limit switches which can l
) provide positive indication of control rod position.
I !
! The reed switch based position indication system is referred to as the Secondary Position ,
- Indication (SPI) system. This system provides a highly accurate indication of actual control l
, rod position, but at a lower precision than the synchros. The reed switches are wired so that ;
the voltage read ac~oss the reed switch stack is proportional to rod position. The reed switches l j are spaced along a tube with a center to center spacing distance of 1.5 inches. The resolution I of the SPI reed switch stacks is 1.5 inches. The reed switches also provide input to the matrix l -l indication lights w'aich provide control rod status indication for various key positions. To l 1 increase the reliability of the system, there are redundant reed switches which prevent false l
] indication in the event an individual reed switch fails. l l
A control rod pt,sition deviation alarm is provided to alert the operator when any two control rods in the samt group are more than 8 inches apart. This helps to ensure any control rod misalignments ara minimized. The alarm can be generated by either the SPI system or PIP l 1 node since the SPl system, in conjunction with the host computer, is redundant to the PIP node l 4
in the task of contril rod measurement, control rod monitoring, and limit processing. l l
4 4
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- l t
1
- B 3.1-24 -l Revised 02/05/99-i l
dnpt [o/ @ Alignment 6nlTogb
..qq BASES APPLICABLE 'fdetermingthattherequfredSDMismet/withthemaximugi l SAFETY ANALYSES worth CFA also fully wJthdrawn (Ref 5). /
(continued) rop iricidents result in the most rapid l Since the.@Daamina =ce.otanin nimi ansion riattegSAFDLH @
CeMol r c3 J. approach't t causec oy a M miso ation, t cident analysis analyzed 1
. drop. e st rapTo ap on t I
} thrDN8a single F full len t Us}rre may* a ngle full le gth drop o a subgroup / rop,' depe ing upon in ,tal y Jg.nd Jon 1
,oired rna or r,*y nF M Jhe above automatic reactor isoperations result in an
- p. In the case 01 the full length @ Qh b l drop, a prompt decrease in core average power and a 4 distortion in radial power are initially produced, which, when conservatively coupled, result in a local power and heatfluxincrease,{+dadecreaseinDNBRparameters.
- s. rod The results of the(C'5D+misoperation analysis show that during the most limiting misoperation events, no violations h of the SAFDLs, fuel centerline temperature, or(%Spressure occur.
hte.t.ro$ @ alignment satisfies Criteria 2 and 3 of EFe71R77dTicF1 star ==ent] fo cFt, 50 9 Wth h aaJ Port ( p oj LCO ) regulating, allgnments ensure The limits that the on shutdown'n@the assumptions i safety ana@ lysis will remain valid. The requirements on OPERA 81LITY ensure that upon c.amc coa reactor trip, the@ will be available and will be Inserted to provide enough negative reactivity to shut down th ctor. The OPERABIL'ITY recuirements also ensure that banks maintain the correct power distribution and q 9 ignmentg
! 3 Th requirement is maintain the CEA Alignment to wit n l
- g* @ W [ inches) between ny CEA and its grodp. The minin.us m alignment assu d in safety analysis is L15 inches} and l y some cases, a potal misalignment from fully withdrawn to Jully inserted ir assumed. / /
f j Failure to meet the requirements of this LCO may produce
, unacceptable power peaking factors an:! LHRs, or unacceptable I (continued)
CE0G STS B 3.1-26 Rev 1, 04/07/95 4
9 Revised f 02/05/99 l
1 l
SECTION 3.1 i
I INSERT i
.....and that each control rod is capable of being moved by its CRDM. The OPERABILITY requirement for the part-length rods is that they are fully withdrawn. l l The requirement is to maintain the control rod alignment to within 8 inches between any control rod and all other rods in its group. To help ensure this requirement is met, the control l rod position deviation alarm generated by either the PIP node or SPI system, must be .
1 OPERABLE and provide an alarm when any control rod becomes misaligned > 8 inches from any other rod in its group. The safety analysis assumes a total misalignment from folly withdrawn to fully inserted. This case bounds the safety analysis for a single rod in any intermediate position.
The primary rod position indication syem is considered OPERABLE, for purposes of this specification, if the digital position readaut, the PPC display, or the cam operated position indication lights give positive indication of rod position. The secondary rod position indication system is considered OPERABLE if the magnetically operated reed switches are providing I
positive indication of rod position either via the plant process computer or taking direct readings of the output from the magnetic reed switches.
i l
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l l
l B 3.: 26 l 1
Revised )
02/05/99
SECTION 3.1 INSERT 1 Performance of a CHANNEL CHECK on the primary and secondary control rod position l indication channels provides confidence in the accurracy of the rod position indication systems.
INSERT 2
... which correspond to the lower electrical limit and the upper electrical limit respectively, ;
i
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B 3.1-31 Revised 02/05/99
(C,.<he.t K.d] AlignmentI Wnal5 B 3.1 $
N h '
V BASES SURVEILLANCE SR 3.1.5.3 (continued)
REQUIREMENTS 3 can be de cted, and protecti n can be provided the CEA deviatto circuits.
, 97
@ h{ SR 3.1. .fh Q De'nonstratin the deviationtifcuridisOPERABLEverifies @
b the'tiretrit) s func onal. The(@ day Frequency takes in account other information continuously available to the .< S
~
operator in the control room, so that during R8 movement deviations can be detectedl and Arotection gen De provided)--Q3 lby th VCEA motion JifhibitF-h h SR 3.1. .f h g ,a,,g ,g Verifying each Q8 is trippable would required,that each be tripped. In MODES I and 2, tripping each @ would result in radial or axial power tilts, or oscillations.
.htt.ic,3\ L e, ode.1 < .a Jd Inerefore, TMisiduaTM are exercised every 92 days to c.+.1 ~
provide increased confidence that all (C(AYconfin'ue t6'be r J' trippable, even if they are not regularly tripped. A 0' 6) movementoft$inchesPisadequatetodemonstratemotioncr*t<g without exceeding the alignment limit when only one is being moved. The 92 day Frequency takes into consideration other information available to the operator in the control room and other surveillances being performed more frequently, which add to the determination of OPERABILITY of the CEAs. (Between r ' ired performances f SR 3.1.5.5, fi a C9(s/isdiscovered be immovable, bu remains tripoabre AnLRligng_Qb.e C is__cp/)tLdered to oprRAntr fat any bI '*$ _._ time, if a CDe(s) is immovable, a determination s of the @
trippability ((OPyRABIt.171) of the${y) must be made, a appropriate action taken. gj
< Y
@ @( SR 3. l BL8)@ gg Performance of a CHANNEL LFUJET!0NAL/TESTlof each free 6 switchl phod E4ca position 1transmitt'erichannel ensures the channel is OPERABLE
(
Q,,g
, .dp ----4nd.capabl.Lof. indicttingQ length of the'129's position travel,. Since this over the entire Surveillance must be Lp~J I performea wnen the reactor is shut down, an 18 month L lAU.EI Frequency to be coincident with refueling outage was (continued)
CEOG STS B 3.1-32 Rev 1, 04/07/95
- Revised 02/05/99
Shutdown Insertion Lizitts b and L vL @ R.J 6 p}
B 3.1 REACTIVUYCONTROLSYSTEMS B 3.1 Shutdown l Con (rol Elem/nt Assemb)'y (CEA)l Insertion 9 Limits [Ap BASES
~J L+-Leg, R.4 G Q BACKGROUND The insertion limits of the shutdown are initial '-
assumptions in all safety analyses that assume 71nsertion upon reactor trip. The insertion limits directly affect core pow r distributions and assumptions of available SDN, ejected rth, and initial reactivity insertion rate. lh
, I,[1J3 gf{ T / Q 4(heapplicablecriteriaforthesereactivityandpower I distribution design requirements __.Jare 10 R 50, Appepdix A, IMC To Reactor Design! anTGDC 26, " eactivity (mits" (Ref. ), and 10 CFR 50. 6 "Acceptan Criteria r Oct Emerjency,_ Cort fool _ing itemi_for L ht Water N clear _Pjtwtr_
Reuters3 (Ref d2 hbeen efraETTFed,J. and Limits on shutdownJfB all T&& insertionand positions'are monitored haveh controlled during power operation to ensure that the reactivity limits, ejected G S worth, and SDN limits are lh preserved.
Q The shutdown are arranged into groups that are radially l h synenet ric. oes _ _ t o not introduceT radial erefore, movementinofthe asymmetries theco shutdown wer h,R 8_'*-'1 distribution. The shutdown and regulating rovide the required reactivity worth for immediate reacto shutdown J. jus ut T 2. a, up n a react r trip.
ga y..,) _
yyQ The design calculation are performed with the assumpti~on ,
f,a t t the shutdown re withdrawn prior to the regulating
<,.c (. The shutdown can be fully withdrawn without the core going critical. This provides available negative __
reactivit for SDM in the event of boration errors. dWD~
ll ecM y.- Edsn are controlled manually f or/ auto'natitallyl by the control room operator. 8 ,
l During normal e shutdown M are fully withdrawn. Q&D operation, The shutdown @ Qf'j/
must be completely withdrawn from the core prior to '
withdrawing any regulating LC As F during an approach to d, - ;d I criticality. The shutdown CEAsl are then lef t in this @ 9'--ID bT'
~ -
position until the reactor is shut down. They affect core power, burnup distribution, and add negative reactivity to shut down the reactor upon receipt of a reactor trip signal.
(continued)
CEOG STS B 3.1-34 Rev 1, 04/07/95
- Revised j 02/05/99
SECTION 3.1 INSERT 1 -
The Palisades Nuclear Plant design criteria (Ref.1) and 10 CFR 50.46, " Acceptance Criteria for Emergency Core Cooling System for Light Water Nuclear Power Reactors," contains..
INSERT 2 The Palisades Nuclear Plant has four part-length control rods installed. The part length rods are required to remain completely withdrawn during power operations, except during rod exercising performed in conjunction with SR 3.1.4.3. The part-length rods do not insert on a l reactor trip.
B 3.1-34 Revised-02/05/99~
Shutdown nsertion Limits [XWaldgih
@ ws @u u + eaa, f 12 BASES (continued)
G.d y e os)
APPLICABLE Accident analysis assumes that the shutdown M are fully ,h SAFETY ANALYSES withdrawn any time the reactor is critical. This ensures that:
- a. The minimum SDM is maintained; an g
- b. The potential effects of a ejection accident are ,
limited to acceptable limits. I are considered fully withdrawn at I nches, since Q <.1 ro<l h @this position places them[outsife the act/ve region pj
& in s . n s. r u Le s a r ue+. .h s.,s < aa .
J & ; 4c,,.QJ uo,4 <t-m G< e ek L. . k. "
.- ) ,
p \e,gk cohol <cJs On a reactor trio. allHAf1(shutdown and regulating), '
except the most reactife M are assumed to insert into the lg-1
[ v .d .i cope,~ Tfie shutdown and regulating QA2 shall be at,.their'@:f D I insertion limits and available to insert the CLHmd!Lamou tv Ws of negative reactivity on a reactor trip signal. The <* d B ,
. regulating D TQ may be partially inserted in the core as e allowed by'LC0_3.1@,G'RegulatingIContrM Element Assem6ly lh Q)
{G r o g P.s a,,pS (inh rtfori limi to is established t sensure
. " Thethatshutdown U$ insertion a sufficient amount _ limit of BY-1)g negative reactivity is available to shut down the reactor I_ and maintain the ranuired SDM (see LCO 3.1.1, " SHUTDOWN o,
~TSU Ol MARGIN ull Jower.(SDM){T[fcomoinauono>
Tri 290'F$
regulating following CTii and shutdown
(
a reactor Otrip
' @ - (less the most
@ fully withdrawn) isreactTve'
, which is assumed to sufficient to take the reactor from e~.-
f power conditions at rated temperature to zero power, and to 4
maintain the required SDM at rated no load temperature n
@m(_Re f. ))) . The shutdownfCE insertion limit also lim reactivity worth of c1 aniejected v.-b shutdown g Q gGot y~o) l @
The acceptance riteriaforaddressingshutdownffsaswell I t as regulating insertion limits and inoperability or misalignment are t atT(<4y Q
- a. There be no violation of:
- l. specified acceptable fuel design limits, or
- 2. ts4etc6 Coolant System pressure boundary damage; and Q
- b. The core remains subcritical after accider.t transients.
(continued)
CEOG STS B 3.1-35 Rev 1, 04/07/95
- Revised 02/05/99
SECTION 3.1 INSERT 1 The part length rods have the potential to cause power distribution envelopes to be exceeded if inserted while the reactor is critical. Therefore, they must remain withdrawn in accordance with the limits of the LCO (Ref. 3).
INSERT 2 For a control rod group to be considered above its insertion limit, all rods in that group must l be above the insertion limit. l Maintaining the shutdown rod groups within their insertion limits... -
INSERT 3 Maintaining the part length rod group within its insertion limit ensures that the power distribution envelope is maintained.
INSERT 4 l In MODE 2, the Applicability begins anytime any regulating rod is withdrawn above 5 inches.
INSERT 5
...to at least the lower electrical limit, and contribum to the SDM. In MODE 3, the shutdown rod groups may be withdrawn in preparation for a reactor startup.
INSERT 6 The Applicabilty has been modified by a Note indicating the LCO requirement is suspended
- during SR 3.1.4.3 (rod exercise test). Control rod exercising verifies the freedom of the rods I to move, and requires the individual shutdown rods to move below the LCO limits for their I group. Only the full-length rods are required to be tested by SR 3.1.4.3. The part-length rods I may also be moved however, if a part-length rod is moved below the limit of the associated l LCO the Required Action of Condition A must be taken. l Positioning of an individual control rod within its group is addressed by LCO 3.1.4, " Control Rod Alignment."
B 3.1-36 s.vi..d 02/05/99
l '
Shutdown EA Insertion Li:: lits ((An gog) /
s J.s y
.hd aci 2. q % l9wd 6twf _ 5 BASES ACYIONS h A.1 7 A #.2( ad8 A M (continued) core as it is inserted. IIf Dor concentration is t
, Jchanged at t(is time, $0M shou not change. This, however, O c* '
,, is verified ivithin I hour, or oration is initiate to bring i theSOMtojithinlimit,ifteCEA(s)isnotres red to within limiAs crior to this ime.
I If the EA(s) i not restor d to withi limits wit n I ho e 50H is within limi , then an ditional 1 our is and all ed for r storing th CEA(s) to
~
thin 1Tmit . The @
2 our tota Completion ime allows he operato adequ e me to ad st the CE s) in an or erly manner and te f onsiste with the r quired Com etion Time in [. s3.1.5, 7 "
/ @
N Control / Element As ably (CEA) lignment."
Q-N.5E12T1 I L.1 -
When Required Action A.1 W cannot be met or completed.
within the required Completion Time, a controlled shutdown should be commenced. The allowed Cuipletion Time of 6 ')ours is reasonable, based on operating experience, for reaching MODE 3 from full power conditions in an orderly manner and without challenging plant systems.
I l
SURVE!LLANCE SR 3.1.6.1 ad for+ fg e rod $revf/.
REQUIREMENTS Verification that the shutdown are within their insertion limits prior to an approach to criticality ensures that when the tor is critical, or being taken critical, O/% MFtheshutdo
- will be available to shut down the a l
r reactor~ ~Td a e required SON will be maintained following a \.2/
~.LNIMT[ reactor trip.s This SR and Frequency ensure that the :
_snutcown urn are withdrawn before the regulating ($TAs)are l withdrawn during 4 M$*startup.
W nis i Co& and PoA fodgry J hh{/ Sincetheshutdo2 are positioned manuall control room operat@or, verification of shutdow he position at a Fre_q c of 12 Tours is ade1UTG"to ~ ensure t at the
, shutdowfi are within their insertion limits. Also, the 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Frequency takes into account other information available to the operator in the control room for
_ purpose of monitoring the status of the shutdow '
(continued)
CEOG STS B 3.1-37 Rev 1, 04/07/95
' Revised 02/o5/99
I l
SECTION 3.1 INSERT 1 '
If one or more shutdown or part length rods are not within limits, the affected rod (s) must be declared inoperable and the applicable Conditions and Required Actions of LCO 3.1.4 entered immediately. This Required Action is based on the recognition that the shutdown and part length rods are normally withdrawn beyond their insertion limits and are capable of being moved by their control rod drive mechanism. Although the requirements of this LCO are not applicable during performance of the control rod exercise test, the inability to restore a control rod to within the limits of the LCO following rod exercising would be indicative of a problem affecting the OPERABILITY of the control rod. Therefore, entering the applicable Conditions and Required Actions of LCO 3.1.4 is appropriate since they provide the applicable compensatory measures commensurate with the inoperability of the control rod.
INSERT 2 i
4 Verification that the part length rod groups are within their insertion limits ensures that they do not adversely affect power distribution requirements.
i
)
4 B 3.1-37 i
I
+
hvised 02/0$/99
RegulatingRfA Wsert4onLLisits f(panIT i
8 3*l #
@ R./(m yfackdl 1
B 3.1 REACTIVITY CONTROL SYSTEMS l
8 3.lf Regulating IContr41 Elemerit AssembTy (CEA) Irdertionl Limits KAnayogA b
BASES Rod Gr. p Pos%
l k
to 'QrDAOD The insertion limits of the regulating (EbTare initial "
~
1 BACKGROUND assumptions in all safety analyses that assume Wrtion O't upon reactor trip. The insertion limits directly affect
- core power distributions, assumptions of available SON, and ,
j initial reactivity insertion rate. The applicable criteria
- co,AA cJ .. & for these reactivity and power distribution desian ,
i requirementsareJ10CFR50,AppendixA,GDC10."Reactorf
- M
- 5.Jes MW' Pk# _! Design," and GOC 26, " Reactivity Limits *l(Ref. 1), and
~
'10~CFR 50.46, " Acceptance criteria for Emergency Core
- 9 3:,$ , , , % , s Cooling Systems for Light Water Nuclear Power Reactors" l
(Ref. 2). e.d y.3 Limits on regulating insertion have been established, tg
{ and all@ positions are monitored and controlled during f g.3
("_ ,r.JL -$,o4~~ J power operation to ensure that the power distribution and reactiv.ity limits defined by the design power peaking, i Q e.jected'0@ worth, reactivity insertion rate, and SOM limits l 4 are preserved.
The regulating up; operate with a predetermined in order to approximate a line lh l.
_. 4m99.n,1 of position overlaprelatioT6cfsiend worth indG position 4(integral I worth). The regulating @ groups are withdrawn and operate in a predetermined sequence. The group sequence and overlap ,
limits are specified in the COLR. l
<d . i The regulati are used or precise reactivity contro 4 q l of the reactor, e positions of the regulating Cess are I manually controlled. They are capable of adding reactivity very quickly (compared to borating or diluting).
The power density at any point in the core must be limited to maintain specified acceptable fuel design limits, including limits that preserve the criteria specified in 10_ CFR 50.46 (Ref. 2). Together, LCO 3.lf/),bWM
. r&fTement-ffssemtfT~dt'TAbbliseth+t1mm1; y LCO 3.2.4,@,L MUTifAU FOGER TILT (T.)'; and LCO 3.2. Mil' AXIAL SHAPE l h-Q d'ANDEX (ASI)," provide limits on control component operation and on monitored process variables to ensure the core operates within the linear heat rate (LCO 3.2.1, " Linear Heat. Rate (LHR)") Itoral plarnari radial peaking factor WD (( ..J F"l (continued)
CE0G STS B 3.1-39 Rev 1, 04/07/95
- n.vis.d 02/0s/99
SECTION 3.1 INSERT 1 The most limiting SDM requirements for Mode 1 and 2 conditions at (Beginning of Cycle (BOC) are determined by the requirements of several transients, e.g., Loss of Flow, etc.
Ilowever, the most limiting SDM requirements for Modes 1 and 2 at End of Cycle (EOC) come from just one transient, Main Steam Line Break (MSLB). The requirements of the MSLB event at EOC fu the full power and no load conditions are significantly larger than those of any other event at that time in cycle and, also, considerably larger than the most limiting requirements at BOC.
Although the most limiting SDM requirements at EOC are much larger than those at BOC, the available SDMs obtained via tripping the full-length control rods are substantially larger due to i the much lower boron concentration at EOC. To verify that adequate SDMs are available throughout the cycle to satisfy the changing requirements, calculations are performed at both BOC and EOC. It has been determined that calculations at these two times in cycle are sufficient since the difference between available SDMs and the limiting SDM requirements are the smallest at these times in cycle. The measurement of full-length control rod bank worth l performed as part of the Startup Testing Program demonstrates that the core has the expected shutdown capability. Consequently, adherence to LCO 3.1.5, " Shutdown and Part-Length Rod Group Insertion Limits," and LCO 3.1.6 provides assurance that the available SDM at I any time in cycle will exceed the limiting SDM requirements at that time in cycle.
INSERT 2 For a control rod group to be considered above its insertion limit, all rods in that group must l ;
be above the insertion limit. I j l
l i
B 3.1-42 Revised 02/05/99
Regulating ICEA/InsertAoniLicits IMalaa_1 (R .4 G r pPah[
- 8ASES (continued)
APPLICA8ILITY The regulating q ence overlap, and physical insertion d .
limits shall be maintained with the reactor in MODES 1 and 2. These limits must be maintained, since they preserve n the assumed power distribution, ejected C D worth, 50M, and
- L'V reactivity rate insertion assumptions. Applicability in MODES 3, 4, and 5 is not re distribution nor ejected tred,sinceneitherthepower.@
worth assumptions would be exceeded in these MODES is preserved in MODES 3, 4, and 5 by djustmen oluble boron co ynctration. -
A loco.w a - d .3 C M _< u <...r<D J This as been mod i requirement is suspended dur_ing by aSR Note3. indi ating the LCRm W (.,w. fred c oJ s verifies the freedom of the to move, and requires the ,
, d 'ML Gods)p regulating 6 to move glo@w the LCO limits, which would normally violate the LCOV IThe Note a o allows the JC0 to benotaplicableduringreactorpow cutback oper4 tion, j h " " *f bd ich i serts a selected CEA croup sually group f) during loss o load eventsl.
o.n/
ACTIONS A.1M4 ACT). 2[ A . M 1. anf A . 2. Y /
Operation beyond theltrAnst4ntiinsertion limit may result in lh a loss of SDM and excessive peaking factors. l If the regul fiigl.tA inser 'IbWTillim'Tri WOL 1, then SOM st be v rified by perf raing a reactivity alance calcul ion, c idering the li ted reactivity eff ts in Bases 5 tion f MTF4.sI51- 5 3.1.1.1. One our is sufficient me for conduc ing the c iculation onsnincing boration f the SDM is at within imits.l The sientiinsertion limit should not be ih violated dur ng normal ~ operation; this violation, however, hq- D 'O3r may controllingoccur during@the transients in responsewhen the operator to changing plant is manually l@
conditions. When the regulating groups are inserted beyond theltr(nsteotlinsertion limits, actions must be taken to either withdraw the regulating groups beyond the limits or l@
to reduce THERMAL POWER to less _than or equal to that Ga y . r r3'+ d allowed for the actual _insfrfT5fiilimit. Two hours provides a reasonable t me to accomplish this, allowing th
~
l
, operator to deal with current plant conditions while 4 limiting peaking factors to acceptable levels. !
l (continued) 1 .
CE06 STS B 3.1-43 Rev 1, 04/07/95
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i Revised e 02/0s/99
ATTACHMENT 6 JUSTIFICATION FOR DEVIATIONS SPECIFICATION 3.1.1, SHUTDOWN MARGIN, T ,i > 200 F Change Discussion
, 7. ISTS Change Traveler TSTF-136 combines ISTS 3.1.1 and ISTS 3.1.2 into a single specification in order to eliminate unnecessary and confusing duplication, and renumbers the remaining specifications in Section 3.1. The impetus for this change was the approval of TSTF-9 which allowed the values for shutdown margin to be moved to the COLR. As a result of TSTF-9, the LCO, Actions, and Surveillance Requirements ofISTS 3.1.1 and ISTS 3.1.2 were the same. Palisades has relocated the shutdown l margin values to the COLR in accordance with TSTF-9 and has consolidated l ISTS 3.1.1 and ISTS 3.1.2 into a single specification. Proposed ITS 3.1.1 address the plant conditions encompassed in MODES 3,4, and 5 as a result of this consolidation. l
- 8. The Palisades plant was designed prior to issuance of the General Design Criteria (GDC) in 10 CFR 50. Therefore, reference to the GDCs is omitted and appropriately replaced by reference to " Palisades Nuclear Plant design criteria ." The Palisades Nuclear Plant design was compared to the GDCs as they appeared in 10 CFR 50 Appendix A on July 7,1971. It was this updated discussion, including the identified exemptions, which formed the origir. plant Licensing Basis for future compliance with l the GDCs.
- 9. TSTF-9 permits relocation of the shutdown margin values specified in ISTS 3.1.1 and ISTS 3.1.2 to the COLR. Palisades has elected to exercise this option in the ITS. The l l appropriate justification for this change is provided in DOC LA.1 for ITS 3.1.1. l
- 10. Samarium is not considered in the Palisades Nuclear Plant reactivity balance due to the fact the that Palisades Nuclear Plant fuel vendor does not account for Samarium in fuel design calculations. The vendor assumes that the negative reactivity defect due to Samarium is offset by the positive reactivity of Plutonium build in. Plutonium build in and Samarium are equally competing reactivity effects that are accounted for in fuel design calculations. Therefore, including Samarium into the SDM calculation would not be correct for the Palisades Nuclear Plant.
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4 Palisades Nuclear Plant Page 2 of 3 02/05/99
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1 A'ITACHMENT 6 i JUSTIFICATION FOR DEVIATIONS SPECIFICATION 3.1.5, CONTROL ELEMENT ASSEMBLY (CEA) ALIGNMENT Chance Discussion
- 6. The Frequency for the rod position deviation alarm surveillance (ISTS SR 3.1.5.4) is I being decreased from 31 days to 18 months. Verification of that alarm's operability l involves misaligning each control rod group until the alarm actuates. This involves 'l ;
both exceeding the LCO 3.1.4 group alignment limits and moving part length rods. l Neither of these actions is desired during power operation. The CTS neither requires l this alarm to be Operable nor includes any associated surveillance requirement. Since l Palisades rods are manually controlled, and rod group alignments are verified after l moving rods, the alarm is not as significant as in a plant with automatic rod control. l
- 7. The NUREG-1432 Action A (regulating rods), and Action B (shutdown rods) requirements to restore the misaligned rod to within 7 inches of its group or to restore the group to within 7 inches of the misaligned rod (Action A only) were consolidated into one Action as a result of TSTF-143. Since the CTS does not require misaligned rods to be restored, ISTS Required Actions A.3.1 and A.3.2, as modified by TSTF-143, are not included in the ITS. In addition, neither the CTS, nor the ITS make a distinguish between misaligned shutdown rods or misaligned regulating rods.
Therefore, ISTS Condition B is not required.
- 8. NUREG-1432 LCO 3.1.5 requires that control rods must be aligned to within a certain i amount of inches "(indicated position)" of their respective group. Including the term
" indicated position" is not appropriate for the Palisades plant. The NUREG was based on plants which use magnetic jacks as the mechanism for moving the control rods.
These type of mechanisms typically have a demand position and an indicated position.
A " demand" is placed on the magnetic jack to move a certain amount and this is reflected in the control rod "dernand counter" whether or not the control rod actually moved. The term " indicated position" would refer to the position indication system which is actually monitoring control rod travel. The design at the Palisades plant uses a primary and secondary rod position indicating system with both systems actually indicating " actual" rod position since there is no " demand" position. Therefore, the term " indicated position" is not included in the Palisades ITS.
- 9. NUREG-1432, Condition A is modeled after plants which have an analysis which have varying amounts of rod misalignment. The Palisades CTS only assumes that a control rod is either within limits or is misaligned. There are no actions or supporting analysis for differing amounts of misalignment. Therefore, NUREG-1432 is revised, in the applicable portions, to only discuss misalignments greater than 8 inches in the Palisades Nuclear Plant proposed ITS.
Palisades Nuclear Plant Page 2 of 7 02/05/99
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ATTACHMENT 6 JUSTIFICATION FOR DEVIATIONS SPECIFICATION 3.1.6, SIIUTDOWN CEA INSERTION LIMITS Chance Disencion l
- 7. NUREG-1432 has a Note in the Applicability which modifies the LCO by stating "This i LCO is not applicable while performing SR 3.1.5.5." In NUREG-1432, SR 3.1.5.5 is the rod exercise test which corresponds to SR 3.1.4.3 in the proposed Palisades ITS l l The Bases discussion for the Applicability Note has been modified to clarify the requirement of rod testing as it relates to part-length rods. Part-length control rods do not have to tested by SR 3.1.4.3 since they are not trippable. Periodically, the part- l length rods may need to be moved to help restore the mechanical seal integrity of the control rod drive mechanism. Performing part-length rod exercising in conjunction with SR 3.1.4.3 ensures it is performed under controlled conditions. CTS 3.10.3, Part- l ;
Length Control Rods states that "The part-length control rods will be completely withdrawn from the core (except for control rod exercises and physics tests)." As such, the Applicability Note in ITS 3.1.5 as clarified by the Bases is consistent with the current licensing basis.
- 8. A discussion has been added in the Bases under the LCO section to clarify that if an individual shutdown or part-length rod does not meet the insertion limit requirement, then LCO 3.1.4, " Control Rod Alignment," may be entered as long as the remainder of )
the group is above its insertion limits. This discussion was added to help avoid confusion since LCO 3.1.5 is written to address shutdown and part-length rods on a group basis and LCO 3.1.4 addresses individual rod misalignments. This is a plant specific change to reflect the Palisades control rod design and CTS requirements.
1
- 9. The Palisades plant was designed prior to issuance of the General Design Criteria l (GDC) in 10 CFR 50. Therefore, reference to the GDCs is omitted and appropriately replaced by reference to the " Palisades Nuclear Plant design." The Palisades Nuclear Plant design was compared to the GDCs as they appeared in 10 CFR 50 Appendix A on July 7,1971. It was this updated discussion, including the identified exemptions, which formed the original plant Licensing Basis for future compliance with the GDCs.
- 10. The Palisades plant always runs with the rod control system in manual. The automatic feature of the rod control system has been disabled. Therefore, references to automatic rod control have been deleted. This is a plant specific change to reflect the Palisades design and operating practices.
Palisades Nuclear Plant Page 2 of 4 02/05/99
ATTACHMENT 6 JUSTIFICATION FOR DEVIATIONS ,
SPECIFICATION 3.1.6, SHUTDOWN CEA INSERTION LIMITS I
Change Discussion I
- 14. The NUREG-1432 Bases in the Applicability section states "In MODE 3,4,5, or 6, the shutdown CEAs are fully inserted in the core and contribute to the SDM. In the i proposed ITS, MODE 3 was deleted from this sentence and another sentence added to state "In MODE 3, the shutdown rod groups are not always fully inserted. In addition, the term " fully inserted" is changed in the proposed ITS to state "to at least the lower electrical limit." This change is made to remove confusion with respect to what
- constitutes " full inserted." For the Palisades control rod design, the lower electrical i
limit corresponds to the point where electrical rod insertion ceases, and is about 3 inches from the bottom of full rod travel. The reactivity level in this region is
- negligible. These changes are plant specific changes to provide clarification of the requirements for shutdown rod groups.
- 15. To reflect the incorporation of TSTF-136 which consolidates ISTS 3.1.1 and ISTS 3.1.2, the specification number for ISTS 3.1.6, " Shutdown CEA Insertion Limits," has been changed to ITS 3.1.5 and conforming changes have been made to the Bases. These changes are consistent with NUREG-1432 as modified by TSTF-136.
- 16. The definition of Shutdown Margin was revised in NUREG-1432 to clarify that changes l in fuel and moderator temperature are included in the determination of the Control Element Assembly Power Dependent Insertion Limits which are used to ensure adequate Shutdown Margin in MODES 1 and 2. As a result of this change, ISTS 3.1.6 Required Action A.1.1 (verify SDM) and Required Action A.1.2 (initiate boration) j have been deleted since they are no longer necessary to ensure adequate Shutdown
! Margin. Therefore, these Required Actions and associated Bases discussions are not l included in proposed ITS 3.1.5. This change is consistent with NUREG-1432 as
) modified by TSTF-67, i 17. ISTS 3.1.6 Required Action A.1 (as modified by TSTF-67) allows 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> to restore l !
i '
out-of-limit shutdown rods to within the limit of the LCO. Proposed ITS 3.1.5 l l Required Action A.1 requires out-of-limit shutdown (and part-length) rods to be l declared inoperable and the Conditions and Required Actions ofITS 3.1.4 entered l l immediately. Anytime it is discovered that a control rod can not be moved by its l operator the control rod must be considered inoperable. Since movement of the l ;
shutdown rods is typically limited to the control rod exercise test, the inability to l restore a shutdown rod to within the limits of the LCO would be indicative of an l inoperable (i.e., immovable) control rod. Therefore, the Required Actions for a l shutdown rod outside its specified limit has been changed to be consistent with the l 4
Required Actions for an inoperable control rod. l
. Palisades Nuclear Plant Page 4 of 4 02/05/99
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ATTACHMENT 6 JUSTIFICATION FOR DEVIATIONS l SPECIFICATION 3.1.7, REGULATING CEA INSERTION LIMITS i Chanoe Discuccion I
- 12. The Palisades Nuclear Plant analysis does not model separate insertion limits for transient and steady state conditions as specified in Conditions A, B and C of NUREG-1432. The Palisades Nuclear Plant PDIL limits specify the regulating rod group position limits which account for anticipated power maneuvers and transient mitigation. Therefore, the proposed Palisades ITS removes the steady state and transient insertion limit discussion, where appropriate, and provides a discussion of the Palisades Nuclear Plant insertion limits. This is a plant specific change to reflect the Palisades CTS and analysis. I
- 13. A discussion has been added in the Bases under the LCO section to clarify that for a j control rod group to be considered above its insertion limit, all rods in that group must [
be above the insertion limit. This is a plant specific change to reflect the Palisades l control rod design and CTS requirements. ]
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- 14. To reflect the incorporation of TSTF-136 which consolidates ISTS 3.1.1 and ISTS 3.1.2, the specification number for ISTS 3.1.7, " Shutdown CEA Insertion Limits," has been changed to ITS 3.1.6 and conforming changes have been made to the l Bases. These changes are consistent with NUREG-1432 as modified by TSTF-136.
- 15. The definition of Shutdown Margin was revised in NUREG-1432 to clarify that changes in fuel and moderator temperature are included in the determination of the Control Element Assembly Power Dependent Insertion Limits which are used to ensure adequate Shutdown Margin in MODES 1 and 2. As a result of this change, ISTS 3.1.7
, Required Action A.1.1 (verify SDM) and Required Action A.1.2 (initiate boration) have been deleted since they are no longer necessary to ensure adequate Shutdown Margin. Therefore, these Required Actions and associated Bases discussions are not included in proposed ITS 3.1.6. An expanded discussion has been incorporated in the Applicable Safety Analyses portion of the Bases to clarify the requirements for SDM as it applies to control rod position. These change are consistent with NUREG-1432 as modified by TSTF-67.
Palisades Nuclear Plant Page 4 of 5 02/05/99
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ENCLOSURE 5 l
CONSUMERS ENERGY COMPANY l
PALISADES PLANT DOCKET 50-255 1
CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS !
RESPONSE TO DECEMBER 4, 1998 )
REQUEST FOR ADDITIONAL INFORMATION l
1 REVISED PAGE'S FOR SECTION 3.2 I
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l CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO DECEMBER 4, 1998 REQUEST FOR ADDITIONAL INFORMATION REVISED PAGES FOR SECTION 3.2 Pace Chance Instructinni Revise the Palisades submittal for conversion to Improved Technical Specifications by removing the pages identified below and inserting the attached pages. The revised pages are identified by date and contain vertical lines in the margin indicating the areas of change.
REMOVE PAGES INSERT PAGES REV DATE NRC COMMENT #
ATTACHMENT 1 TO ITS CONVERSION SUBMITTAL ITS 3.2.1-1 ITS 3.2.1-1 02/05/99 RAI 3.2-01 ITS 3.2.1-2 ITS 3.2.1-2 02/05/99 RAI 3.2-01 ITS 3.2.1-3 ITS 3.2.1-3 02/05/99 RAI 3.2-01 ATTACHMENT 2 TO ITS CONVERSION SUBMIIIAL ITS B 3.2.1-2 ITS B 3.2.1-2 02/05/99 RAI 3.2-01
. ITS B 3.2.1-3 ITS B 3.2.1-3 02/05/99 RAI 3.2-01 ITS B 3.2.1-5 ITS B 3.2.1-5 02/05/99 RAI 3.2-01 ITS B 3.2.1-6 ITS B 3.2.1-6 02/05/99 RAI 3.2-01 ITS B 3.2.1-7 ITS B 3.2.1-7 02/05/99 RAI 3.2-01 ITS B 3.2.1-8 ITS B 3.2.1-8 02/05/99 RAI 3.2-01 ITS B 3.2.1-9 ITS B 3.2.1-9 02/05/99 RAI 3.2-01 ITS B 3.2.3-2 ITS B 3.2.3-2 02/05/99 RAI 3.2-07
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ATTACHMENT 3 TO ITS CONVERSION SUjMITTAL DOC 3.2.1, pg 2 of 7 DOC 3.2.1, pg 2 of 7 02/05/99 RAI 3.2-01 j DOC 3.2.1, pg 4 of 7 DOC 3.2.1, pg 4 of 7 02/05/99 RAI 3.2-01 DOC 3.2.2, pg 3 of 4 DOC 3.2.2, pg 3 of 4 02/05/99 RAI 3.2-04 ATTACHMENT 4 TO ITS CONVERSION SUBMITTAL NSHC 3.2.2, pg 1 of 5 NSHC 3.2.2, pg 1 of 5 02/05/99 RAI 3.2-04 ATTACHMENT 5. TO ITS CONVERSION SUBMITTAL NUREG 3.2-1 NUREG 3.2-1 02/05/99 RAI 3.2-01 NUREG 3.2-1 insert NUREG 3.2-1 insert 02/05/99 editorial NUREG 3.2-2 insert NUREG 3.2-2 insert 02/05/99 RAI 3.2-02 i NUREG 3.2-3 NUREG 3.2-3 02/05/99 RAI 3.2-01 NUREG B 3.2-4 insert NUREG B 3.2-4 insert 02/05/99 RAI 3.2-01 NUREG B 3.2-5 insert NUREG B 3.2-5 insert 02/05/99 RAI 3.2-01 !
NUREG B 3.2-23 NUREG B 3.2-23 02/05/99 RAI 3.2-07 )
ATTACHMENT 6 10 ITS CONVERSION SUBMITTAL JFD 3.2.1, pg 1 of 5 JFD 3.2.1, pg 1 of 5 02/05/99 RAI 3.2-01 JFD 3.2.1, pg 3 of 5 JFD 3.2.1, pg 3 of 5 02/05/99 RAI 3.2-01 JFD 3.2.1, pg 4 of 5 JFD 3.2.1, pg 4 of 5 02/05/99 RAI 3.2-01
o LHR 3.2.1 3.2 POWER DISTRIBUTION LIMITS 3.2.1 Linear Heat Rate (LHR)
LCO 3.2.1 LHR shall be within the limits specified in the COLR, and the Incore Alarm System or Excore Monitoring System shall be OPERABLE to monitor LHR.
APPLICABILITY: MODE 1 with THERMAL POWER > 25% RTP.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. LHR, as determined by A.1 Restore LHR to within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> the automatic Incore limits.
Alarm System, not within limits specified in the COLR, as indicated by four or more coincident incore channels.
QB LHR, as determined by the Excore Monitoring System, not within limits specified in I the COLR.
QB LHR, as determined by manual incore detector I readings, not within limits specified in the COLR.
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l Palisades Nuclear Plant 3.2.1-1 Amendment No. 02/05/99 ,
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k LHR
- 3.2.1 ACTIONS
- CONDITION REQUIRED ACTION COMPLETION TIME B. Incore Alarm and B.1 Reduce THERMAL POWER- 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> Excore Monitoring to s 85% RTP.
Systems inoperable for monitoring LHR. 680 B.2 Verify LHR is within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> limits using manual 4 incore readings. ARQ i Once per 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> i thereafter A
C. Required Action and C.1 Reduce THERMAL POWER 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> associated Completion to s 25% RTP.
4 Time not met.
i SURVEILLANCE REQUIREMENTS
, SURVEILLANCE FREQUENCY 4
SR 3.2.1.1 -------------------NOTE--------------------
Only required when Incore Alarm System is l being used to monitor LHR.
! Verify LHR is within the limits specified 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> 4
in the COLR.
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Palisades Nuclear Plant 3.2.1-2 Amendment No. 02/05/99
LHR 3.2.1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.2.1.2 -------------------NOTE--------------------
Only required when Incore Alarm System is l being used to monitor LHR.
Adjust incore alarm setpoints based on a Prior to measured power distribution. opera. tion > 50%
RTP after each fuel loading h_E 31 EFPD thereafter SR 3.2.1.3 -------------------NOTE--------------------
Only required when Excore Monitoring System is being used to monitor LHR.
Verify measured ASI has been within 0.05 of Prior to each target ASI for last 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. initial use of Excore Monitoring System to monitor LHR SR 3.2.1.4 -------------------NOTE-------------------- )
Only required when Excore Monitoring System is being used to monitor LHR.
Verify THERMAL POWER is less than the APL. I hour Palisades Nuclear Plant 3.2.1-3 Amendment No. 02/05/99 1
i LHR )
B 3.2.1 I l
l BASES l BACKGROUND Power distribution is a product of multiple parameters, )
(continued) various combinations of which may produce acceptable power !
distributions.
I The limits on LHR, Assembly Radial Peaking Factor (F,^), j Total Radial Peaking Factor (F,'), QUADRANT POWER TILT (T,),
and AXIAL SHAPE INDEX (ASI), which are obtained directly !
from the core reload analysis, ensure compliance with the j safety limits on LHR and Departure from Nucleate Boiling l
Ratio (DNBR).
l Either of the two core power distribution monitoring !
systems, the Incore Alarm System or the Excore Monitoring I )
System, provides adequate monitoring of the core power distribution and is capable of verifying that the LHR is 1 within its limits. The Incore Alarm System performs this l l function by continuously monitoring the local power at many 1 points throughout the core and comparing the measurements to ,
predetermined setpoints above which the limit on LHR could I be exceeded. The Excore Monitoring System performs this !
function by providing comparison of the measured core ASI with predetermined ASI limits based on incore measurements.
An Excore Monitoring System Allowable Power Level (APL),
which may be less than RATED THERMAL POWER, and an additional restriction on T,, are applied when using the i Excore Monitoring System to ensure that the ASI limits adequately restrict the LHR to less than the limiting l values. )
In conjunction with the use of the Excore Monitoring System )
for monitoring LHR and in establishing ASI limits, the i following assumptions are made:
- a. The control rod insertion limits of LC0 3.1.5,
" Shutdown and Part-Length Rod Group Insertion Limits," l and LC0 3.1.6, " Regulating Rod Group Position Limits," ]
are satisfied, l
- b. The additional T, restriction of SR 3.2.1.6 is !
satisfied; and !
- c. Radial Peaking Factors, F,^ and F,', do not exceed the limits of LC0 3.2.2. i l
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Palisades Nuclear Plant- B 3.2.1-2 02/05/99 ;
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LHR B 3.2.1
! BASES
. BACKGROUND The limitations on the Radial Peaking Factors provided in (continued) the COLR ensure that the assumptions used in the analysis for establishin Settings (LSSS)g the LHR remain valid limits duringand Limiting operation Safety at the System various allowable control rod group insertion limits.
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The Incore Alarm System continuously provides a direct i 1 measure of the LHR and the Radial Peaking factors. It also
- provides alarms that have been established for the individual incore detector segments, ensuring that the peak i
LHRs are maintained within the limits specified in the COLR.
- The setpoints for these alarms include tolerances, set in' conservative directions, for
- a. A measurement calculational uncertainty factor (as identified in the COLR);
- b. An engineering uncertainty factor of 1.03; and
- c. A THERMAL POWER measurement uncertainty factor
, of 1.02.
I The measurement uncertainties associated with LHR, F/ and ;
F/ are based on a statistical analysis performed on power I distribution benchmarking results. The COLR includes the applicable measurement uncertainties for fresh and depleted incore detector usage. The engineering and THERMAL POWER ,
uncertainties are incorporated in the power distribution j calculation performed by the fuel vendor.
1 The excore power distribution monitoring system consists of !
Power Range Channels 5 through 8. The power range channels !
monitor neutron flux from 0 to 125 percent full power. They i are arranged symmetrically around the reactor core to provide information on the radial and axial flux distributions.
The power range detector assembly consists of two uncompensated ion chambers for each channel. One detector extends axially along the lower half of the core while the I other, which is located directly above it, monitors flux from the upper half of the core. The DC current signal from each of the ion chambers is fed directly to the control room
, drawer assembly without- pre-amplification. Each excore detector supplies data to a Thermal Margin Monitor (TMM). 1 Each TMN uses these excore signals to calculate Axial Shape i Index (ASI) on a continuous basis. !
Palisades Nuclear Plant B 3.2.1-3 02/05/99
LHR B 3.2.1 BASES APPLICABLE c. During an ejected rod accident, the fission energy SAFETY ANALYSES input to the fuel must not exceed 280 cal /gm; and (continued)
- d. The full-length control rods must be capable of I shutting down the reactor with a minimum required SDM with the highest worth control rod stuck fully withdrawn (Ref. 3).
The power density at any point in the core must be limited to maintain the fuel design criteria (Ref. 4). This is accomplished by maintaining the power distribution and primary coolant conditions so that the peak LHR and DNB parameters are within operating limits supported by accident analyses (Ref. 1), with due regard for the correlations between measured quantities, the power distribution, and uncertainties in determining the power distribution.
Fuel cladding failure during a LOCA is limited by restricting the maximum linear heat generation rate so that the peak cladding temperature does not exceed 2200 F (Ref.4). High peak cladding temperatures are assumed to cause severe cladding failure by oxidation due to a Zircaloy water reaction. l The LCOs governing LHR, ASI, and the Primary Coolant System Operation ensure that these criteria are met as long as the I core is operated within the LHR, ASI, F/, F/, and T, limits. l The latter are process variables that characterize the three dimensional power distribution of the reactor core.
Operation within the limits for these variables ensures that their actual values are within the ranges used in the accident analyses.
Fuel cladding damage does not necessarily occur while the plant is operating at conditions outside the limits of these LCOs during normal-operation. Fuel cladding damage could result, however, if an accident occurs from initial conditions outside the limits of these LCOs. The potential for fuel cladding damage exists because changes in the power distribution can cause increased power peaking and can correspondingly increase local LHR.
The Incore Alarm System provides for monitoring of LHR, I radial peaking factors, and QUADRANT POWER TILT to ensure that fuel design conditions a'nd safety analysis assumptions are maintained. The Incore Alarm System is also utilized to I determine the target AXIAL OFFSET (A0) and to determine the Allowable Power Level (APL) when using the excore detectors.
Palisades Nuclear Plant B 3.2.1-5 02/05/99
LHR B 3.2.1 BASES APPLICABLE The Excore Monitoring System provides for monitoring of ASI SAFETY ANALYSES and QUADRANT POWER TILT to ensure that fuel design (continued) conditions and safety analysis assumptions are maintained.
The LHR satisfies Criterion 2 of 10 CFR 50.36(c)(2).
LC0 The power distribution LC0 limits are based on correlations i between power peaking and certain measured variables used as ;
inputs to the LHR and DNBR operating limits. The power i distribution LC0 limits, except T , are provided in the COLR. The limitation on the LHR In the peak power fuel rod at the peak power elevation Z ensures that, in the event of 1 a LOCA, the peak temparature of the fuel cladding ooes not (xceed 2200 F.
The LC0 requires that LHR be maintained within the limits l l specified in the COLR and either the Incore Alarm System or i j Excore Monitoring System be OPERABLE to monitor LHR. When I l using the Incore Alarm System, the LHR is not considered to l be out of limits until there are four or more incore detectors simultaneously in alann. When using the Excore Monitoring System, LHR is considered within' limits when the conditions are acceptable for use of the Excore Monitoring l System and the associated ASI and T, limits specified in the SRs are met. ;
1 To be considered OPERABLE, the Incore Alarm System must have I ;
at least 160 of the 215 possible incore detectors OPERABLE !
and 2 incore detectors per axial level per core quadrant I l In addition, the plant process computer must be OPERABLE. I l OPERABLE and the required alarm setpoints entered into the I l plant computer, j l
To be considered OPERABLE, the Excore Monitoring System must have been calibrated with OPERABLE incore detectors, the ASI must not have been out of limits for the last 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, and ,
THERMAL POWER must be less than the APL. j l
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Palisades Nuclear Plant B 3.2.1-6 02/05/99 1
LHR' l B 3.2.1- i j ,
BASES $
APPLICABILITY In' MODE 1 with THERMAL' POWER > 25% RTP, power distribution : .,
must be maintained within the' limits assumed in'the accident '1 analysis to ensure that fuel damage dees not result- i following an A00. In MODE 1 with THERMAL POWER s 25% RTP, .l and in other MODES, this LC0 does not apoly because there is' j not sufficient-THERMAL POWER to require a limit on the core -
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power distribution, and because ample thermal margin exists to ensure that the fuel integrity is not jeopardized 'and' .;
safety analysis assumptions remain valid. <
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ACTIONS A.J There are three acceptable methods for verifying that LHR is
within limits. The LCO requires monitoring by either an ;
OPERABLE Incore Alarm System.or an OPERABLE Excore l- .;
Monitoring System. When both of the required systems are ;
inoperable, Condition B allows for monitoring by taking i manual readings of the incore detectors. Any of these three ;
l methods may indicate that the LHR is not within limits. :
- With the LHR exceeding its limit, excessive fuel damage I could occur following an accident. In this Condition, J prompt action must be taken to restore the LHR.to within the i
specified limits. One hour to restore the LHR to within its. ,
j specified' limits is reasonable and ensures that the core i i does not continue to operate in this Condition. The 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> ]
Completion Time also allows the operator sufficient time for -)
j .' evaluating core conditions and for initiating proper- i
- corrective actions.
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j, Palisades Nuclear Plant. B 3.2.1-7 102/05/99 l
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, a a . . , _ . .
LHR B 302.1 BASES ~
ACTIONS 8.1 and B.2 (continued)
With the Incore Alann System inoperable for monitoring LHR l and the Excore Monitoring System inoperable for monitoring-LHR, THERMAL POWER must be reduced to s 85% RTP within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. Operation at s 85% RTP ensures that ample thermal margin is maintained. A 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> Completion Time is adequate to achieve the m uired plant condition without challenging plant systems. acditionally, with the Incore Alarm and Excore Monitoring Systems inoperable, LHR must be verified to be within limits within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, and every 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> thereafter by manually collecting incore detector readings at the terminal blocks in the control room. utilizing a suitable signal detector. The manual readings shall be taken on a minimum of 10 individual detectors per quadrant (to include a total of 160 detectors in a 10 hour1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> period).
The time interval of 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and the minimum of 10 detectors per quadrant are sufficient to maintain adequate surveillance of the power distribution to detect significant changes until the monitoring systems are returned to
! service.
3 i .C.d 1
If the Required Action and associated Completion Time are I not met, THERMAL POWER must be reduced to s 25% RTP. This reduced power level ensures that the core is operating
! within its thermal limits and places the core in a
- conservative condition. The allowed Completion Time of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> is reasonable, based on operating experience, to i reach s 25% RPT from full power MODE 1 conditions in an orderly manner and without challenging plant systems.
1 Palisades Nuclear Plant B 3.2.1-8 .r 105/99
LHR B 3.2.1 BASES SURVEILLANCE SR 3,2.1.1
. REQUIREMENTS The Incore Alarm System provides continuous monitoring of I
LHR through the plant computer. The plant computer is used to generate alarm setpoints that are based on measured j margin to 61 owed LHR. As the incore detectors are read by the plant computer, they are continuously compared to the alarm setpuints. If the Incore Alarm System LHR monitoring i function is inoperable, excore detectors or manual recordings of the incere detector readings may be used to ,
monitor LHR. Periodically monitoring LHR ensures that the assumptions made in the Safety Analysis are maintained. )
This SR is modified by a Note that states that the SR is i
.only applicable when the Incore Alarm System is being used l I to monitor LHR. The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Frequency is consistent with an SR which is to be performed each shift. .) 1 t
SR 3.2.1.2 Continuous monitoring of the LHR is provided by the Incore
- Alarm System which provides adequate monitoring of the core I power distribution and is capable of verifying that the LHR does not exceed its specified limits.
j Performance of this SR verifies'the Incore Alarm System can I accurately monitor LHR by ensuring the alarm setpoints are based on a measured power distribution. Therefore, they are only applicable when the Incore Alarm System is being used I to determine the LHR.
The alarm setpoints must be initially adjusted following each fuel loading prior to operation above 50% RTP, and periodically adjusted every 31 Effective Full Power Days (EFPD) thereafter. A 31 EFPD Frequency is consistent with the historical testing frequency of the reactor monitoring system. The SR is modified by a Note which allows the SR to i be performed only when the Incore Alarm System is being used I to determine LHR.
l Palisades Nuclear Plant B 3.2.1-9 02/05/99
T, B 3.2.3 BASES ACTIONS M If the measured T is > 0.05, T must be restored within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or F ^ and, F,' must be de,termined to be within the limits of LC0 3.2.2, and determined to be within these limits every 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> thereafter, as long as T, is out of limits. Two hours is sufficient time to allow the operator I to reposition control rods, and significant radial xenon redistribution cannot occur within this time. The 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> Completion Time ensures changes in F," and F,' can be identified before the limits of LC0 3.2.2 are exceeded.
M With the measured T > 0.10, power must be reduced to i
< 50% RTP within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, and F," and F,' must be within I their specified limits to ensure that acceptable flux )
peaking factors are maintained as required by Condition A (which continues to be applicable). Based on operating l experience, 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> is sufficient time for evaluation of these factors. If F," and F,' are within limits, operation i may proceed while attempts are made to restore T, to within its limit. If the tilt is generated due to a control rod I ,
misalignment, continued operation at < 50% RTP allows for I realignment; if the cause is other than control rod I misalignment, continued operation may be necessary to I discover the cause of the tilt. Reducing THERMAL POWER to l
< 50% RTP, and the more frequent measurement of peaking l factors required by ACTION A.1, provide conservative i protection from potential increased peaking due to xenon I redistribution. I fo_1 If T is > 0.15, or if Required Actions and associated CompletionTimesarenotmet,THERMALPOWERmustbereduced to s 25% RTP. This requirement ensures that the core is operating within its thermal limits and places the core in a conservative condition. Four hours is a reasonable time to reach 25% RTP in an orderly manner and without challenging plant systems.
Palisades Nuclear Plant B 3.2.3-2 02/05/99
ATTACHMENT 3 .
DISCUSSION OF CHANGES SPECIFICATION 3.2.1, LINEAR HEAT RATE A.4 CTS 3.23.1 provides actions when the LHR is being monitored by the excore i monitoring system but the system is no longer appropriate for monitoring LHR as indicated by an Axial Offset (AO) of more than 0.05 (ACTION 2). The actions include ,
both " discontinue using the excore monitoring system for monitoring LHR" and
" follow the procedure in ACTION 3 below." Inherent in entry into CTS 3.23.1 ACTION 2 is that the normally used Incore Alarm System is inoperable. Therefore, l this situation is one with both the Incore Alarm System and the excore monitoring l
- system inoperable for the purpose of monitoring LHR. This is included as ITS 3.2.1 1 Condition B. The specific direction to enter this Condition is not included in ITS since j this is the normal use and application of the improved STS format. Therefore, this ,
j omission is considered an administrative change. J
! A'. 5 CTS 3.23.1 provides actions when the LHR is indicated as not within the limits ]
l specified in the COLR by four or more coincident incore alarms (ACTION 1), and
]
! vhen the manually recorded incore readings indicate a local power level greater than 3
alarm setpoints (ACTION 3). .However, no specific action is provided in the CTS fi when the LHR is not within limits as monitored by the excore monitoring system.
l P ITS includes a second entry condition for ITS 3.2.1 Condition A specifically for j ..en the LHR is determined to be not within limits using the excore monitoring ;
j system, Since the appropriate action is the same regardless of the method used to )
j determine that LHR is not within limits, the addition of a specific Required Action, j entry condition for "LHR, as determined by the Excore Monitoring System, not within j 4
limits specified in the COLR" is cansidered an admuustrative change. 1
- I t
- A.6 CTS 3.23.1 ACTION 3 indicates that when the LHR is indicated as not within the l
! limits specified in the COLR by the manually recorded incore readings "the action i specified in ACTION 1 above shall be taken." The ITS includes a third entry condition for ITS 3.2.1 Condition A specifically for when the LHR is determined to be not within limits using the manual incore readings, Since these are only different formats to
- require the same action, the addition of a specific Required Action, entry condition for
- "LHR, as determined by manual incore readings, not within limits specified in the j COLR" is considered an administrative change.
i I
1 Palisades Noclear Plant Page 2 of 7 '02/05/99
ATTACHMENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.2.I. LINEAR HEAT RATE M CTS does not include specific surveillance requirements to verify that LHR remains within limits. Such an SR is included as ITS SR 3.2.1.1. This SR is necessary to provide direct verification that the LCO requirements are met when using the Incore Alarm System for monitoring LHR. Consistent with the NUREG, verification that an l OPERABLE Incore Alarm System does not indicate LHR out of limits is sufficient to l fulfill this SR. This is an additional restriction on plant operation.
LESS RESTRICTIVE CHANGES - REMOVAL OF DETAILS TO LICENSEE CONTROLLED DOCUMENTS (LA)
LA.1 CTS 3.23.1 contains specific details regarding the requirements for monitoring of the LHR, i.e., "in the peak powe- fuel rod at the peak power elevation Z." This information is not required to De provided in NUREG LCO 3.2.1. These details describe elements of the LHR which are addressed by the methodology for determining LHR and are not directly a part of the actual requirement, i.e., Limiting Condition for Operation. Since these details are not necessary to adequately describe the actual regulatory requirement, they can be moved to a licensee controlled document without a significant impact on safety. Placing these details in the LCO Bases ofITS 3.2.1 provides adequate assurance that they will be maintained. The Bases are controlled by the Bases Control Program in Chapter 5 of the proposed Technical Specifications. This change is consistent with NUREG-1432.
LA.2 CTS 3.23.1 ACTION 3 contains specific details regarding the requirements for monitoring of LHR by manual readings of the incore detection system when the incore LHR alarm system is inoperable, i.e., " readings shall be taken on a minimum of 10 individual detectors per quadrant (to include a total number of 160 detectors in a 10-hour period)." This information is not provided in NUREG LCO 3.2.1. These
' details describe elements of the incor.e detection system requirements which are addressed by the methodology for proper use of the system and are not directly a part of the actual requirement, i.e., Limiting Condition for Operation. Since these details are not necessary to adequately describe the actual regulatory requirement, they can be moved to a licensee controlled document without a significant impact on safety.
Placing these details in the Bases ofITS 3.2.1 provides adequate assurance that they will be maintained. The Bases are controlled by the Bases Control Program in Chapter 5 of the proposed Technical Specifications. This change is consistent with NUREG-1432.
Palisades Nuclear Plant Page 4 of 7 02/05/99
AT'1 ACHMENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.2.2, RADIAL PEAKING FACTORS LA.2 CTS 4.19.2.1 provides Surveillance Requirements (SRs) for the Radial Peaking Factors. However, it contains specific details for monitoring of the peaking factors, j i.e., that the SR is performed by verifying the " measured" radial peaking factors "obtained by using the incore detection system." This information is not provided in NUREG SR 3.2.2.1. These details describe elements of the radial peaking factor verification which are addressed by the methodology and are not directly a part of the actual requirement, i.e., Surveillance Requirement. Since these details are not necessary to adequately describe the actual regulatory requirement, they can be moved to a licensee controlled document without a significant impact on safety. Placing these details in the Bases of ITS SR 3.2.2.1 provides adequate assurance that they will be maintained. The Bases are controlled by the Bases Control Program in Chapter 5 of the proposed Technical Specifications. This change is consistent with NUREG-1432.
LESS RESTRICTIVE CHANGES (L)
L.1 CTS 3.23.2 provides actions for peaking factors exceeding their limits based on power l level. The first of these actions is for P (power) < 50%, and requires the plant to be l in at least hot shutdown (i.e., subcritical) within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. ITS 3.2.2 Required Action l A.1 provides 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> to attempt restoration of the peaking factors to within limits, and l if the Required Action and its associated Completion Time is not met, then Required l Action B.1 requires that THERMAL POWER be reduced to s 25% RTP. This change l is less restrictive in two ways. First,6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> is provided to attempt restoration of the l peaking factors to within limits that is not provided in the CTS. Second, the default l action requires only that the plant to be reduced to s 25% RTP, rather than subcritical, l in the subsequent 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. l The ITS Required Action to restore the radial peaking factors to the within limits specified in the COLR assure the plant will not operate for an extended period with the l peaking factors not within limits. The Completion Time of 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> provides a l reasonable time for evaluating core conditions, calculating a reduced power level at l which the peaking factors would be within limits, determining the proper method for l the power reduction (e.g., rod positioning and/or boration) and, completing the l reduction in power. In the event the peaking factors are not restored to within limits, l an additional 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> is provided to remove the plant from the mode of applicability. l Although CTS 3.23.2 requires the plant to be placed in hot shutdown, terrninating the l power reduction anywhere below 25% is permissible since CTS LCO 3.0.1 only l requires compliance with an LCO during the plant condition specified in that LCO. l Thus, the default action of proposed ITS Required Action B.1 is consistent with the l
- shutdown action for CTS 3.23.2. A Completion Time of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> is reasonable to l
! reduce thermal power below 25% in an orderly manner and without challenging plant l-
[ systems. l l
Palisades Nuclear Plant Page 3 of 4 02/05/99 4
ATTACHMENT 4 NO SIGNIFICANT HAZARDS CONSIDERATION SPECIFICATION 3.2.2, RADIAL PEAKING LESS RESTRICTIVE CHANGE L,1 CTS 3.23.2 provides actions for peaking factors exceeding their limits based on power level. l The first of these actions is for P (power) < 50%, and requires the plant to be in at least hot l shutdown (i.e., subcritical) within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. ITS 3.2.2 Required Action A.1 provides 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> l to attempt restoration of the peaking factors to within limits, and if the Required Action and its l associated Completion Time is not met, then Required Action B.1 requires that THERMAL l POWER be reduced to s 25% RTP. This change is less restrictive in two ways. First, l 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> is provided to attempt restoration of the peaking factors to within limits that is not l provided in the CTS. Second, the default action requires only that the plant to be reduced to l c 25% RTP, rather than subcritical, in the subsequent 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. l l
1 The ITS Required Action to restore the radial peaking factors to the within limits specified in l the COLR assure the plant will not operate for an extended period with the peaking factors not l within limits. The Completion Time of 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> provides a reasonable time for evaluating core l conditions, calculating a reduced power level at which the peaking factors would be within l l
limits, determining the proper method for the power reduction (e.g., rod positioning and/or l boration) and, completing the reduction in power. In the event the peaking factors are not l restored to within limits, an additional 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> is provided to remove the plant from the mode l of applicability. Although CTS 3.23.2 requires the plant to be placed in hot shutdown, l terminating the power reduction anywhere below 25% is permissible since CTS LCO 3.0.1 l l only requires compliance with an LCO during the plant condition specified in that LCO. Thus, l j the default action of proposed ITS Required Action B.1 is consistent with the shutdown action l for CTS 3.23.2. A Completion Time of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> is reasonable to reduce thermal power below l 25% in an orderly manner and without challenging plant systems. l Palisades Nuclear Plant Page I of 5 02/05/99
l
. 1 l
\
l l
h LHR f(Ani1 # 1 3 .'z . . ,
i 3.2 POWER DISTRIBUTION LlHITS
@ 3.2.1 Linear Heat Rate (LHR) [(kfalsgf}
h 3,11.1 uo LCO 3.2.1 (be wiAM LHR.shall mer merc0 the limits specified in the COLR.
end % TNorr M.a.em Syoke e e c % O rt. Imaukhy
@. ifJi+ DloII & OfGhbll +s thss n+v- QfA, v 3 23.1 p g APPLICABILITY: MODE 1 v
3[.idT_{CfetALPoW6E *2f(4R.TP]
l ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME
@ A. LHR, as determined by A.1 Restore LHR to within I hour 3.23 i
/tet I the Incore (URTErm limits.
Am System h {mG;EspeAnw Eng)1imits sciNelk l
- e Iseed.El-r cil t e - -
h- COLR, as indicated by four or more coincident incore channels.
E d nw LHR, as determined by LHR , as de+erminedlag g the Excore M Monitoring System a i, erM;+,
h {fw;& % n as i icat by iimi as no t wikin lieuiff I l
AS 37,,;f;,f g g e ggtyp,,
outsi the ower e dent ontr 4
lim ts nafsoecified in 3,0.1 (;nurn.2/1-2M the NT3 GULR.
h.TA)SCETM 1 Required Atticn and .1 Firyf00r hours ,
s,21.1 associated Completion f M3 Hme not met. ( Powc2 gg,,4gggg 5 :Lf% ETP. l
{ _
\
CEOG STS 3.2-1 Rev 1, 04/07/95 02/05/99
SECTION 3.2 INSERT B. Incore Alarm and Excore B.1 Reduce THERMAL POWER 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> Monitoring Systems to s 85% RTP.
SMI inoperable for monitoring Adt LHR. M B.2 Verify LHR is within limits 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> l using manual incore readings.
3.15l M Nc13 Once per 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> thereafter 3.2-1 Revised 02/05/99
l SECTION 3.2 INSERT SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.2.1.1 --NOTE-----
Only required when Incore Alarm System is being used l to monitor LIIR. l
}
Verify LHR is within the limits specified in the COLR. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> l
?
3.2-2 a.vi 4 02/05).7,
4 ... k LHRT5AfiyYog()
3.Z.I I
SURVEILLANCE REQUIREMENTS (continued)
SURVEILLANCE FREQUENCY-SR b
- 3. 2.1.q) ----------------
A OT ------------------
. l0niv anolicabi when the Incore pr;n %
y,gg,g,g Gete/ tor MynTuoring) System is being
- oPrdim l'usedtodetermineCHR. 250% ETP
. to perfopdbelo a +c e e 1ncore Rder,k6r %caVrw6r dM ) 5TAD 31 6F?b
( ,
(llf4*f) fueob4er
(.TtJSE2 --
Lbased o_s c~__uyaufeb power ditHMQQ r .\
[ 4/,19. l. 2. A !
i 4,19,l . 7, b 4,14.1,2.c.
419.l. L d Revised 02/05/99 CEOG STS 3.2-3 Rev 1, 04/07/95
SECTION 3.2 1
INSERT A
- Tle Incore Alarm System provides for monitoring of LHR, radial peaking factors, and QUADRANT POWER TILT to ensure that fuel design conditions and safety analysis assumptions are maintained. The Incore Alarm System is also utilized to determine the target AXIAL OFFSET (AO) and to determine the Allowable Power Level (APL) when using the excore detectors.
The Excore Monitoring System provides for monitoring of ASI and QUADRANT POWER TILT to ensure that fuel design conditions and safety analysis assumptions are maintained.
i INSERT B The LCO requires that LHR be maintained within the limits specified in the COLR and either l the Incore Alarm System or Excore Monitoring System be OPERABLE to monitor LHR. l When using the Incore Alarm System, the LHR is not considered to be out of limits until there are four or more incore detectors simultaneously in alarm. When using the Excore Monitoring System, LHR is considered within limits when the conditions are acceptable for use of the Excore Monitoring System and the associated ASI and T, limits specified in the SRs are met.
To be considered OPERABLE, the Incore Alarm System must have at least 160 of the 215 l possible incore detectors OPERABLE and 2 incore detectors per axial level per core quadrant l ;
OPERABLE. In addition, the plant process computer must be OPERABLE and the required l l alann setpoints must be entered into the plant computer.
l To be considered OPERABLE, the Excore Monitoring System must have been calibrated with OPERABLE incore detectors, the ASI must not have been out of limits for the last 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, 1 and THERMAL POWER must be less than the APL. ;
l l
l l
I i
l B 3.2-4 n.wi..a 02/05/99
'SECTION 3.2 9
INSERT A B.1 and B.2.
F
. With the Incore Alarm System inoperable for monitoring LHR and the Excore Monitoring
. System inoperable for monitoring LHR, THERMAL POWER must be reduced to s 85% RTP within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. Operation at s 85% RTP ensures that ample thermal margin is maintained. A -
-2 hour Completion Time is adequate to achieve the required unit condition without challenging plant systems. Additionally, with the Incore Alarm and Excore Monitoring Systems inoperable, LHR must be verified to be within limits within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, and every 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> thereafter by manually collecting incore detector readings at the terminal blocks in the control room utilizing a suitable signal detector. The manual readings shall be taken on a minimum of 10 individual detectors per quadrant (to include a total of 160 detectors in a 10 hour1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> period).
The time interval of 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and the minimum of 10 detectors per quadrant are sufficient to -
naintain adequate surveillance of the power distribution to detect significant changes until the monbring systems are returned to service.
INSERT B The Incore Alarm System provides continuous monitoring of LHR through the plant computer. l The plant computer is used to generate alarm setpoints that are based on measured margin to allowed LHR. As the incore detectors are read by the plant computer, they are continuously compared to the alarm setpoints. If the Incore Alarm System LHR monitoring function is inoperable, excore detectors or manual recordings of the incore detector readings may be used to monitor LHR. Periodically monitoring LHR ensures that the assumptions made in the Safety Analysis are maintained. This SR is a lified by a Note that states that the SR is only applicable when the Incore Alarm System is bemg used to monitor LHR.
B 3.2-5 n.,ina '
02/05/99
T, UnaMW l 5 3.2 I
BASES (continued) l ACTIONS A . l lam _k2)
- if the measured Tds > DKOF ad <A.1d N2ic61Mibnl i f (o tWBan '
n p lurg) T, must be restored within !
A 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or and fr must be determined to be within the r
Timits of L 3.2.2 aortfar 3ETI. and determined to be Ogh EC " within these limits every 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> thereafter, as long as T,
N .' i'Ap> is out of limits. Gd) hours is sufficient time to allow the 4 operator to reposit on @ , and significant radial xenon redistribution cannot occur within this ting. The 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> Completion Time ensures changes in and F, can be -
O7 identified before the limits of LC .2.2 MWiCFR2Z1 -
(twptgtvelm are exceeded.
& Th. IS > 0. I[, or i If Required Actions and associated Completion Times @
rwntMn Ju are not met, THERMAL POWER must be reduced to l
Of M s. RTP, This requirement ensures that the core is l
l operating within its thermal limits and places the core in a l conservative condition. Four hours is a reasonable time to l reach (49% RTP in an orderly manner and without challenging j plant systems. f powc r met be reduced to < Soo/, MP (E.ladBO wih 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, ata Fl--
- - = -
V rA n6 CfD O Lours) i 3 Q q hC"f^#h,WithYT,>0.10,I and Fi must be within their specified 5' limits to nsure hat acceptable flux peakin factors are maintaine . Based on operating experience, o of 4*78@
. e VM"d'* #h-- sufHcient gime forwonernor toavif uate th rs.
l pAr IflQ)and Fe are within limits, operation @ay proceed 6
/ / uwl X immar4f e.T.htM1 Lim iswieriif'; whi1e l
/ attempts are made to restore T, to within its limit.(
If T. s 0 cannot be a eved, power mu be reduced to s 50% within 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> , if the tilt generated due to CEA salignment, op ating at s 50% allows for the I j . re very of the CE . Except as a re it of CEA /
salignment, T. 0.10 is not expp ted; if it occurs f i continued oper ion of the reactpr may be necessary o
\ discover the ause of the tilt./ If this procedure s I >
followed, operation is restricted to only those nditions l required to identify the caus'e of.the tilt. I t, s necessary to accourit explicitly for power asymmetries because the .
(continued)
CEOG STS B 3.2-23 Rev 1, 04/07/95 a ....u - - s -
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ATTACHMENT 6 JUSTIFICATION FOR DEVIATIONS SPECIFICATION 3.2.1, LINEAR HEAT RATE (LHR)
Chance Discussion Note: This attachment provides a brief discussion of the deviations from NUREG-1432 that were made to support the development of the Palisades Nuclear Plant ITS. The Change Numbers correspond to the respective deviation shown on the "NUREG MARKUPS." The first five justifications were used generically throughout the markup of the NUREG. Not all generic justifications are used in each specification.
- 1. The brackets have been removed and the proper plant specific information or value has been provided.
- 2. Deviations have been made for clarity, grammatical preference, or to establish consistency within the Improved Technical Specifications. These deviations are editorial in nature and do not involve technical changes or changes of intent.
- 3. The requirement / statement has been deleted since it is not applicable to this facility. The following requirements have been renumbered, where applicable, to reflect this deletion.
- 4. Changes have been made (additions, deletions, and/or changes to the NUREG) to reflect the facility specific nomenclature, number, reference, system description, or analysis description.
- 5. This change reflects the current licensing basis / technical specification. These include an ITS 3.2.1 Applicability less restrictive than the NUREG and the addition of an ACTION for determination of LHR using manual readings when both the Incore Alarm System and the excore monitoring system are inoperable l for determining LHR. With power reduced to below 85% RTP (per ITS 3.2.1, Required Action B.1), the manual readings of the incore monitors provide an adequate indication that LHR is within limits. This is consistent with CTS as approved in Amendment 68. Additionally, the proposed Applicability for ITS 3.2.1 is actually more restrictive than CTS 3.23.1 which is applicable only above 50% RTP. An ITS 3.2.1 Applicability of " MODE 1 > 25% RTP" is consistent with the Applicability for the other Power Distribution Limit specifications, and provides for incore adjustments based on power distribution maps prior to exceeding 25% which is consistent with Quadrant Power Tilt needs for incore adjustments.
Palisades Nuclear Plant Page 1 of 5 02/05/99
ATTACHMENT 6 JUSTIFICATION FOR DEVIATIONS SPECIFICATION 3.2.1, LINEAR HEAT RATE (LHR)
Chanae Discuaclan
- 8. An addition to the LCO in incorporated which requires that the LHR be determined by an OPERABLE Incore Alarm System or by an OPERABLE l excore monitoring system. Such an LCO requirement is consistent with the NUREG SR Note which requires that the LHR be determined by either the incore detector monitoring system or the excore detector monitoring system.
However, incorporating the requirement into the LCO provides a more direct indication that the LCO is not met when both the incore LHR alarm function and the excore LHR monitoring function are inoperable (which results in entry into ITS Condition B, as discussed in JFD 5).
- 9. The Surveillance Requirements (SRs) for LHR are revised consistent with the current licensing basis. The NUREG SR Note is inappropriate for Palisades Nuclear Plant because manual reading of the incore monitors is also allowed for determining LHR to be within limits. This is corrected by incorporating the SR Note requirements directly into the LCO (see JFD 8) and adding an ACTION for use of the manual incore readings (see JFDs 5 and 7). The NUREG SRs are also inappropriate for all plams since failure of the alarms or setpoints to be properly set does not mean that the LHR is not within limits. However, SR 3.0.1 would require that the LCO be considered not met when any of these ,
SRs are not met . This is not consistent with the format and content intent of the improved STS NUREGs, is considered overly conservative, and is not adopted.
ITS SR 3.2.1.1 specifically requires the verification that LHR is within the limits specified in the COLR. This SR is a direct verification that the LCO is being met (which is missing from the NUREG). However, since the LHR is normally automatically monitored and alarmed by the incore power distribution monitoring system, the SR is only required to be performed when the Incore Alarm System is being used to determine LHR, and is met by administrative l verification that the Incore Alarm System is OPERABLE for monitoring LHR, l and that the Incore Alarm System does not indicate LHR is not within limits. l Palisades Nuclear Plant Page 3 of 5 02/05/99
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ATTACHMENT 6 JUSTIFICATION FOR DEVIATIONS !
SPECIFICATION 3.2.1, LINEAR HEAT RATE (LHR) '
Chance Discussion
- 9. (continued) i NUREG SR 3.2.1.2 and SR 3.2.1.3 requirements for incore alarms are combined and revised to reflect CTS 4.19.1. ITS SR 3.2.1.2 requires that the incore alarm setpoints be adjusted (i.e., the alarms be set) based on a measured power distribution. This Surveillance provides adequate assurance that the Incore Alarm System is providing accurate monitoring of the LHR. This l change is consistent with CTS 4.19.1 requirements for adjustments of incore alarm settings.
ITS SR 3.2.1.3, SR 3.2.1.4, SR 3.2.1.5, and SR 3.2.1.6 require the verification of parameters that similarly indicate the LHR is within the limits specified in the COLR when using the excore monitoring system. These SRs also provide verification that the parameters are appropriate for use of the excore mor ' ring system to monitor LHR and that the LCO is being met (which is missing from the NUREG). However, since the LHR is normally automatically monitored and alarmed by the Incore Alarm System, these SRs are only required to be met l when the excore monitoring system is being used to determine LHR. These SRs I are generally consistent with the requirements of CTS 4.19.1.2a, b, c, and d.
- 10. The periodic Frequency of NUREG SR 3.2.1.3 is revised to 31 EFPD.
CTS 4.19.1.1 provides requirements to adjust the incore alarm settings based on a measured power distribution on a periodic Frequency of "7 days of power operation." Although the CTS Frequency is based on days of power operation, this is inconsistent with the Frequency of ITS Section 3.1 SRs which are based .
J on EFPD, inconsistent with NUREGs for other vendors (e.g., NUREG-1430 and NUREG-1431) for Power Distribution Limit SRs which are based on EFPD, and inconsistent with preferred methods for tracking this Frequency since EFPD is already required to be tracked to for numerous calculations j related to burnup and other fuel status parameters. When the plant is operatmg steadily at full power there is no difference in the NUREG SR 3.2.1.3 periodic Frequency of "31 days" and the proposed "31 EFPD." However, when the 31 days includes operation at less than full power the "31 EFPD" is longer than the NUREG would allow. Still, the revision to the SR Frequency is acceptable since the Frequency continues to be sufficient to assure the incore alarm settings are appropriately since any change is a slow process.
Palisades Nuclear Plant Page 4 of 5 02/05/99
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ENCLOSURE 6 4
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3 PALISADES PLANT DOCKET 50-255 i
CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS i i
RESPONSE TO DECEMBER 4, 1998 REQUEST FOR ADDITIONAL INFORMATION ,
- l REVISED PAGES FOR CHAPTER 5.0
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CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO DECEMBER 4, 1998 REQUEST FOR ADDITIONAL INFORMATION REVISED PAGES FOR CHAPTER 5.0 Pace Chance Instructions Revise the Palisades submittal for conversion to Improved Technical Specifications by removing the pages identified below and inserting the attached pages. The revised pages are identified by date and contain vertical lines in the margin indicating the areas of change.
BIMOVE PAGES INSERT PAGES REV DATE NRC COMMENT #
ATTACHMENT 1 TO ITS CONVERSION SUBMITTAL ITS 5.0-25 ITS 5.0-25 02/05/99 RAI 3.1-01 ATTACHMENT 2 J_Q_JTS CONVERSION SUBMITTAL No page changes ATTACHMENT 3 TO ITS CONVERSION SUBMITTAL
- CTS 5.0, pg 6-20 CTS 5.0, pg 6-20 02/05/99 RAI 3.1-01 ATTACHMENT 4 TO ITS CONVERSION SUBMITTAL No page changes l ATTACHMENT 510 ITS CONVERSION SUBMITTAL l NUREG 5.6, pg 5.0-21 NUREG 5.6, pg 5.0-21 02/05/99 RAI 3.1-01 l ATTACHMENT 6 TO ITS CONVERSION SUBMITTAL No page changes l
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7 -- - - - - - _ - _ . _ _ _
Reporting Requirements 5.6
,5. 6 Reporting Requirements 5.6.4 Monthly Goeratino Reoort Routine reports of operating statistics and shutdown experience shall be submitted on a monthly basis to the NRC no later than the fifteenth of each month following the calendar month covered by ,
the report. !
1 5.6.5 CORE OPERATING LIMITS REPORT (COLR)
- a. Core operating limits shall be established prior to each reload cycle, or prior to any remaining portion of a reload cycle, and shall be documented in the COLR for the following:
3.1.1 Shutdown Margin 3.1.6 Regulating Rod Group Position Limits 3.2.1 Linear Heat Rate Limits 3.2.2 Radial Peaking Factor Limits 3.2.4 ASI Limits
- b. The analytical methods used to determine the core operating limits shall be those approved by the NRC, specifically those described in the latest approved revision of the following documents:
- 1. XN-75-27(A), " Exxon Nuclear Neutronics Design Methods for Pressurized Water Reactors," and Supplements 1(A),
2(A), 3(P)(A), 4(P)(A), and 5(P)(A); Exxon Nuclear Company. (LCOs 3.1.1, 3.1.6, 3.2.1, 3.2.2, & 3.2.4) l
- 2. ANF-84-73(P)(A), " Advanced Nuclear Fuels Methodology for Pressurized Water Reactors: Analysis of Chapter 15 Events," and Appendix B(P)(A) and Supplements 1(P)(A), 2(P)(A); Advanced Nuclear Fuels Corporation.
(LCOs 3.1.1, 3.1.6, 3.2.1, 3.2.2, & 3.2.4) l
- 3. XN-NF-82-21(P)(A), " Application of Exxon Nuclear Company PWR Thermal Margin Methodology to Mixed Core Configurations," Exxon Nuclear Company.
(LCOs 3.2.1, 3.2.2, & 3.2.4)
- 4. ANF-84-093(P)(A), "Stea:..line Break Methodology for PWRs," and Supplement 1(P)(A); Advanced Nuclear Fuels Corporation. (LCOs3.1.1,3.1.6,3.2.1,3.2.2,& l 3.2.4)
- 5. XN-75-32(P)(A), " Computational Procedure for Evaluating Fuel Rod Bowing," and Supplements 1(P)(A),
2(P)(A), 3(P)(A), and 4(P)(A); Exxon Nuclear Company.
(LCOs 3.1.6. 3.2.1, 3.2.2, & 3.2.4)
Palisades Nuclear Plant 5.0-25 Amendment No. 02/05/99
k.O I ShN 8.0 40MINf1TRAffYE CONTR6tf 5, t,,5 I .6,4-6 Core omaratina timits assert (COLR)
- a. Core operating limits shall be established artor to each reload 4' *Ql l
cycle, or prior to any remaining portion of a reload cycle, and shall be documented in the COLR for the following: g g 3 l.1 Shen f()Wn/ v 3.L9 A51 Limits, t.d ***Y 1.n fo Regulating rc e t - Limits 4.1 3.1.1 Linear Heat Rate M Limits J i
1.1. 't. Radial Peaking Factor Listts l b. The analytical methods used to determine the core operating limits shall be those approved by the NRC, specifically those described in the latest approved revision of the following documents:
l 1. XN 75 27(A), ' Exxon Nuclear Neutronics Design Methods for Pressurized Water Reactors,' and Supplements 1(A), 2(A),
C y p$
3.z M s 1 u. J. .I 3, t. t ANF.84 73(P)(A)3,9' Advanced Nuclear Fuels Methodology for
[
- 2. ,
Pressurized Water Reactors: Analysis of Chapter 15 Events ' .
and Appendix B(P) A and Supplements 1(P)(A )
ed M u la Corporation. (LC 3, IN NF.82 21(P)(A), ' Application of Exxon Nuclear Company PWR Thornel Margin Methodology to Mix Ce fig ons,' -
Exxon Nuclear Company. (LCOs , & ,
l t.s .i u.t 4, ANF 84 093(P)(A), 'Stenaline Break Methodology for PWRs,' and ,
l P d Nc als Corporatton,
- 5. XM.75 32(P){A), ' Computational Procedure for Evaluating Fuel Rod Bowine, and Supplements 1(P)(A), 2 P A), 3 P A 4P A luon Nuclear Company. (LCOs ,
& .< ) 3 1.9 1.1.b 3.t.
- 3. L. 4'.
- 6. 1 as defined by:
EIEM (L (LCOs , A& .
J . .t. \
a) XM.NF.82 20(A) ' Exxon Nuclear Company Evaluation Model EIEM/PWit ECCS Model Updates,' and Supplements 1(P)(A),
2(P)(A), *(P)(A), and 4(P)(A); Exxon Nuclear Company, b) IM.NF.82 07(P)(A), ' Exxon Nuclear Company (CCS Cladding Swelling and Rupture Mode),' Enon Nuclear Company.
c) IM.NF.8154(A), 'R00EX2 Fuel Aod Thersal Mechanical I
Response Evaluation Model,' and Supplements 1(P)(A),
, 2(P)(A), 3(P)(A), and 4(P)(A); Exxon Nuclear Company.
6 20 Amendment No, Me,114
- October 31, 1996 M Of M r
Reporting Requirements
-5.6 5.6 Reporting Requirements c35 Monthly Ooeratino Reoorts gq 5.6.4 (continued)
Jpower operged relief valvepor pressurizer safety / valves,]j shall De submitted on a monthly basis no later than the 15th of each month following the calendar month covered by the report, 6.4.5 5.6.5 CORE OPERATING LIMITS REPORT (COLR) g/
3.P0
- a. Core operating limits shall be established prior to each ue u. < *SA a rpy reload cycle, or prior to any remaining portion of a reload )(
u co uk g* cycle, and shall be documented in the COLR for the
- rl '* '
- lowing:
- u. L L ' **' Y y The indifidual specitncations th!t address co M* limits must be referinced here.
operating ~
f tco 1. L. t LA: .4.
b'5 b. The analytical methods used to deternine the core operating Y f3 L limits shall be those previously reviewed and approved by .
the NRC, specifically those described i following i eco m \~[r u 43 documents: g r % ,, v, , y , ,= y Identify e Topical Repo t(s) by number, title, date,fand 9 pJ1ERT/ y ' NRC staff approval docum t, or identify he staff Saf4ty F Evaluati n Report for a lant specific thodologyby/NRC 1 3
letter d date. / / _
l
- c. The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling
- Systems (ECCS) limits, nuclear limits such as SDM, transient analysis limits, and accident analysis limits) of the safety analysis are met.
' d. The COLR, including any mid cycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.
5.6.6 Rggior Cdlant System (RCcI PRESSURE AND TEM RATURE LIMITS REPORT t#TLR)
- a. 5 pressure and mperature limits f r heatup, cooldown, low temperature eration, critical , and hydrostatic (continued) 5.0 21 Rev 1, 04/07/95 CEOG STS