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Revised TS Pages for ITS Section 3.4
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Issue date:	03/22/1999
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ENCLOSURE 3 CONSUMERS ENERGY COMPANY PALISADES PLANT DOCKET 50-255 CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS ADDITIONAL REVISIONS - ITS SECTION 3.4 REVISED PAGES SECTION 3.4, PRIMARY COOLANT SYSTEM

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PDR ADOCK 05000255 p PDR

CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION REVISED PAGES FOR SECTION 3.4 Page Change 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 COMMENT#

ATTACHMENT 1 TO ITS CONVERSION SUBMITTAL ITS. 3.4.1-2 ITS 3.4.1-2 03/18/99 Tech change ITS 3.4.4-1 ITS 3.4.4-1 03/18/99 NRC request #3 ATTACHMENT 2 TO ITS CONVERSION SUBMITTAL ITS B 3.4.1-5 ITS B 3.4.1-5 03/18/99 Tech change ITS B 3.4.4-2 ITS B 3.4.4-2 03/18/99 NRC request #3 ITS B 3.4.4-3 ITS B 3.4.4-3 03/18/99 NRC request #3 ATTACHMENT 3 TO ITS CONVERSION SUBMITTAL CTS 3.4.4, pg 3-lb CTS 3.4.4, pg 3-lb 03/18/99 NRC request #3 DOC 3.4.1, pg 4 of 6 DOC 3.4.1, pg 4 of 6 03/18/99 Tech change DOC 3.4.4, pg 1 of 3 DOC 3.4.4, pg 1 of 3 03/18/99 NRC request #3 ATTACHMENT 4 TO ITS CONVERSION SUBMITTAL No page changes.

ATTACHMENT 5 TO ITS CONVERSION SUBMITTAL NUREG 3.4.1, pg 3.4-3 NUREG 3.4.1, pg 3.4-3 03/18/99 Tech change NUREG 3.4.2, pg 3.4-4 NUREG 3.4.2, pg 3.4-4 03/18/99 NRC request #2 NUREG 3.4.4, pg 3.4-7 NUREG 3.4.4, pg 3.4-7 03/18/99 NRC request #3 NUREG B 3.4.1, pg B 3.4-5 NUREG 3.4.1, pg B 3.4-5 03/18/99 Tech change NUREG B 3.4.1, NUREG B 3.4.1, pg B 3.4-5 insert pg B 3.4-5 insert 03/18/99 Tech change NUREG B 3.4.2, pg B 3.4-8 NUREG B 3.4.2, pg B 3.4-8 03/18/99 NRC request #2 NUREG B 3.4.4, pg B 3.4-18 NUREG B 3.4.4, pg B 3.4.18 03/18/99 NRC request #3 NUREG B 3.4.4, pg B 3.4-19 NUREG B 3.4.4, pg B 3.4.19 03/18/99 NRC request #3 ATTACHMENT 6 TO ITS CONVERSION SUBMITTAL JFD 3.4.2, pg 2 of 2 JFD 3.4.2, pg 2 of 2 03/18/99 NRC request #2 JFD 3.4.4, pg 2 of 2 JFD 3.4.4, pg 1 of 1 03/18/99 NRC request #3

PCS Pressure, Temperat~, and Flow DNB Limits 3.4.1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME B. Required Action and B.1 Be in MODE 2. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time not met.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3 .4.1.1 Verify pressurizer pressure z 2010 psia and 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months

~ 2100 psia.

SR 3.4.1. 2 Verify PCS cold leg temperature 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months

~ 542.99 + 0.0580(P-2060)+ O.OOOOl(P-2060) 2 2

+ 1.125(W-138) - 0.0205(W-138)

SR 3.4.1. 3 -------------------NOTE--------------------

Not required to be performed until 31 EFPD after z 90% RTP.

Verify PCS total flow rate is 18 months z 352,000 gpm.

After each plugging of 10 or more steam generator tubes Palisades Nuclear Plant 3.4.1-2 Amendment No. 03/18/99

Pf Loops - MODES 1 and 2 3.4.4 3.4 PRIMARY COOLANT SYSTEM (PCS) 3.4.4 PCS Loops - MODES 1 and 2 LCD 3.4.4 Two PCS loops shall be OPERABLE and in operation.

APPLICABILITY: MODES 1 and 2.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Requirements of LCD A.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months not met.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.4.1 Verify each PCS loop is in operation. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Palisades Nuclear Plant 3.4.4-1 Amendment No. 03/18/99

PCS Pressure, Temperat~, and Flow DNB Limits B 3.4.1 BASES SURVEILLANCE SR 3.4.1.3 (continued)

REQUIREMENTS The Frequency of 18 months reflects the importance of verifying flow after a refueling outage where the core has been altered, which may have caused an alteration of flow resistance. PCS flow rate must also be verified after plugging of each 10 or more steam generator tubes since plugging 10 or more tubes could result in an increase in PCS flow resistance. Plugging less than 10 steam generator tubes will not have a significant impact on PCS flow resistance and, as such, does not require a verification of PCS flow rate.

The SR is modified by a Note that states the SR is only required to be performed 31 EFPD after ~ 90% RTP. The Note is necessary to allow measurement of the flow rate at normal operating conditions at power in MODE 1. The most common, and perhaps accurate, method used to perform the PCS total flow surveillance is by means of a primary to secondary heat balance (calorimetric) with the plant at or near full rated power. The most accurate results for such a test are obtained with the plant at or near full power when differential temperatures measured across the reactor are the greatest. Consequently, the test should not be performed until reaching near full power (i.e.,~ 90% RTP) conditions. Similarly, test accuracy is also influenced by plant stability. In order for accurate results to be obtained, steady state plant conditions must exist to permit meaningful data to be gathered during the test.

Typically, following an extended shutdown the secondary -

side of the plant will take up to several days to stabilize after power escalation. It is impracticable to perform a primary to secondary heat balance of the precision required for the PCS flow measurement until stabilization has been achieved. Furthermore, an integral part of the PCS flow heat balance involves the use of Ultrasonic Flow Measurement equipment for measuring steam generator feedwater flow. This equipment requires, stable plant operation at or near full power conditions before it can be used. As such, the Surveillance cannot be performed in MODE 2 or below, and will not yield accurate results if performed below 90% RTP.

REFERENCES 1. FSAR, Section 14.1 Palisades Nuclear Plant B 3.4.1-5 03/18/99

P~Loops - MODES 1 and 2 B 3.4.4 BASES APPLICABLE Both transient and steady state analyses have been SAFETY ANALYSES performed to establish the effect of flow on DNB. The (continued) transient or accident analysis for the plant has been performed assuming four PCPs are in operation. The majority of the plant safety analyses are based on initial conditions at high core power or zero power. The accident analyses that are of most importance to PCP operation are the Loss of Forced Primary Coolant Flow, Primary Coolant Pump Rotor Seizure and Uncontrolled Control Rod Withdrawal events (Ref. 1).

Steady state DNB analysis had been performed for the four pump combination. The steady state DNB analysis, which generates the pressure and temperature and Safety Limit (i.e., the Departure from Nucleate Boiling Ratio (DNBR) limit), assumes a maximum power level of 112% RTP. This is the design overpower condition for four pump operation.

The 112% value is the accident analysis setpoint of the trip and is based on an analysis assumption that bounds possible instrumentation errors. The DNBR limit defines a locus of pressure and temperature points that result in a minimum DNBR greater than or equal to the critical heat flux correlation limit.

PCS Loops - MODES 1 and 2 satisfy Criteria 2 and 3 of 10 CFR 50.36(c)(2).

LCO The purpose of this LCO is to require adequate forced flow for core heat removal. Flow is represented by having both PCS loops with both PCPs in each loop in operation for removal of heat by the two SGs. To meet safety analysis acceptance criteria for DNB, four pumps are required at rated power.

Each OPERABLE loop consists of two PCPs providing forced flow for heat transport to an SG that is OPERABLE in accordance with the Steam Generator Tube Surveillance Program. SG, and hence PCS loop OPERABILITY with regards to SG water level is ensured by the Reactor Protection System (RPS) in MODES 1 and 2. A reactor trip places the plant in MODE 3 if any SG water level is ~ 25.9% (narrow range) as sensed by the RPS. The minimum level to declare the SG OPERABLE is 25.9% (narrow range).

Palisades Nuclear Plant B 3.4.4-2 03/18/99

P~Loops - MODES 1 and 2 B 3.4.4 BASES APPLICABILITY In MODES 1 and 2, the reactor can be critical and thus has the potential to produce maximum THERMAL POWER. Thus, to ensure that the assumptions of the accident analyses remain valid, all PCS loops are required to be in operation in these MODES to prevent DNB and core damage.

The decay heat production rate is much lower than the full power heat rate. As such, the forced circulation flow and heat sink requirements are reduced for lower, noncritical MODES as indicated by the LCOs for MODES 3, 4, 5, and 6.

Operation in other MODES is covered by:

LCO 3.4.5, "PCS Loops-MODE 3";

LCO 3.4.6, "PCS Loops-MODE 4";

LCO 3.4.7, "PCS Loops-MODE 5, Loops Filled";

LCO 3.4.8, "PCS Loops-MODE 5, Loops Not Filled";

LCO 3.9.4, "Shutdown Cooling (SDC) and Coolant Circulation-High Water Level" (MODE 6); and LCO 3.9.5, "Shutdown Cooling (SDC) and Coolant Circulation-Low Water Level" (MODE 6).

ACTIONS If the requirements of the LCO are not met, the Required Action is to reduce power and bring the plant to MODE 3.

This lowers power level and thus reduces the core heat removal needs and minimizes the possibility of violating DNB limits. It should be noted that the reactor will trip and place the plant in MODE 3 as soon as the RPS senses less than four PCPs operating.

The Completion Time of 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months is reasonable, based on operating experience, to reach MODE 3 from full power conditions in an orderly manner and without challenging safety systems.

Palisades Nuclear Plant B 3.4.4-3 03/18/99

/'

,,/

Applies to perable status of t To sp ify certain conditio of the primary be t to assure safe rea or operation.

At least one pn'mary-coofant pump or e shutdown cooling ump with a flow rate eater than or equal t 2810 gpm shall be i operation whenever a ange is being made i the boron concentra on of the primary c lant and the plant i operating in cold s tdown or above, cept during an emerg cy loss of coolant ow situation.

Under ese circumstances, e boron concentrati may be increased wit no* rimary coolant s or shutdo_~~~ooli__9._2um s runnin LC.D.1 b. Four primary coolant pumps shall be in operation whenever the Appl1c:.. reactor is operated above hot shutdown, w1 excep ion:

Before removing a pump from service, thermal power shall be reduced as specified in Table 2.3.1 and appropriate corrective action implemented. With one pump out of service, return the pump to service within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months (return to four-pump operation) or be in hot shutdown (or below) with the reactor tripped (from the C-06 panel, opening the 42-01 and 42-02 circuit breakers) within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . Start-up (above hot shutdown) with less than four pumps is not permitted and power operation with less than three pumps is not permitted.

c.

Both steam generatoh -~~9.J.l.:~~~.£§.R~EJ!._QJ__£~'.fc~T.mi_ng their heat transfer functio w enever ~11e average temperature o t e primary coolant is a ove 300°F.

The AXIAL SHAPE INDEX (ASI) shall be maintained within the limits

specified in the COLR.

(1) When the ASI exceeds the limits specified in the COLR, within 15 minutes initiate corrective actions to restore the ASI to the acceptable region. Restore the ASI to acceptable values within one hour or be at less than 70% of rated power within the following two hours.

3-lb I ()t: I No . J-1-, 8§ , -H-8 , -H-9 , -l-34 , W , -M , +/-6-9 ,

Revised 03/18/99

ATTACHMENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.4.1, PCS PRESSURE, TEMPERATURE & FLOW DNB LIMITS M.3 CTS 4.15 specifies the requirement for primary system flow measurement and states that the measurement shall be made "within the first 31 days of rated power operation."

Proposed SR 3 .4 .1. 3 also requires a verification of the primary system flow rate but stipulates that the SR must be performed within 31 EFPD after reaching or exceeding 90% Rated Thermal Power. SR 3.4.1.3 is more restrictive than CTS 4.15 since it establishes a lower power level (100% versus 903) associated with the performance of the test. Thus, the time the reactor may be operated near the point where DNB could be most limiting (i.e., ~ 90%), without a verification of the required primary system flow rate, is reduced. This is an additional restriction on plant operations and is consistent with NUREG-1432.

RESTRICTIVE CHANGES - REMOVAL OF DETAILS TO LICENSEE CONTROLLED DOCUMENTS (LA)

LA. l CTS 4.15 states that the primary system flow measurement shall be made with "four primary coolant pumps in operation." Proposed SR 3 .4.1. 3 does not specify the number of pumps required to be in operation since the only requirement (of this LCO) is to meet the minimum flow assumed in the analysis. The number of primary coolant pumps required to be in operation to meet the safety analysis assumption for forced flow and core heat removal (and ultimately the acceptance criteria for DNB) is provided in proposed ITS 3 .4.4, "PCS Loops-MODES 1 and 2. The Bases of ITS 3 .4.4 specify that both PCS loops with both primary coolant pumps shall be in operation. Since the details regarding the number of primary coolant pumps is adequately covered in the Bases for ITS 3 .4.4, it is not necessary to place this detail in the SR for flow measurement. Placing these details in the Bases provides adequate assurance that they will be maintained since the Bases are controlled by the Bases Control Program proposed in ITS Chapter 5.0. This change is consistent NUREG-1432.

Palisades Nuclear Plant Page 4of6 03/18/99

ATTACHMENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.4.4, PCS LOOPS MODES 1 AND 2 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 Not used.

A.3 CTS 3.1.lb requires four PCPs to be in operation "whenever the reactor is operated above hot shutdown." Proposed ITS 3.4.4requires four PCPs to be in operation in MODES 1 and 2. The CTS plant condition of "hot shutdown" translates to "MODE 3" in the ITS. As such, the CTS requirement to have four PCPs in operation above "hot shutdown" is the same as the ITS requirement to have four PCPs in operation in MODES 1and2. Thus, the difference between the CTS and the ITS can be characterized as administrative since there is no change in requirements between the CTS and ITS. This change is consistent with NUREG-1432.

Palisades Nuclear Plant Page 1of3 03/18/99

pr.&;s Pressure, Tempe-re, and Flow.{ONBf Limits 3.4.l SURVEILLANCE RE UIREMENTS continued SURVEILLANCE FREQUENCY c.-TS CD SR

!-/. t5 p

Verify c-s-y-v.-re_c_,..P...

s ,....o'_n_fjl....e-at.-.-b-al....a_n_ce-.....ittM'a-..rH~cs .{is.(months total flow rate wythin lijl'l'its sjfo~1fled/1jj)

ArJQ COL

"-rs* >- ...)*~S2. , D:>o ~f rf\-.

A4'{--er- e~

pl~i'~-0~

10" or ,.,..o rli

~+eel.-*vt

~ e~e.t""-~D f'

-h"\o t-S CEOG STS 3.4-3 Rev 1, 04/07/95

~-----*---

Revised 03/18/99

~S Minimu~mperature for Criticality 3.4.2 lo.VO-:.. Tove...

?R1mttl\V. *- .- f A.c..s '=' ~U" 3 .4 &EActo©cooLANT SYSTEM (igtS) 3.4.2 ~S Minimum Temperature for Criticality p 5~

(i) LCO 3 . 4. 2 Each ~Sloop average temperature (T.,._g.) shall be ~J[~"-tf.

APPLICABILITY: MOOE 1 with ava in one more RCS lo ps < [535]"F MOOE 2 wit~ T.,. in one r more RCS l ops <,[535]" and K.tt ~ 1.0.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME TS1 F *f.4; A. T_. in one or more ~S A. l Be in MOOE@ 2. 30 minutes loops not within w 1"\'t\ I'\ c...f~ '- I. 6 limit.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY p 525 lSTf*'ll ~.~ (D. SR ~.4.2.1 Verify ~s T*...j in each loop~ .:£mPF.

------------------=--====-=====-=-==-----==-

CEOG STS 3.4-4 Rev 1, 04/07/95

Revised 03/18/99

and 2 3.4.4 3.4 - cooiANT SYSTEM (lcs)

3. 4. 4 ~S Loops-MODES l and 2 crs LCO 3.4.4 TwoJcs loops shall be OPERl\BLE and in operation.

3.1.16

3. 1.1 d APPLICABILITY: MODES l and 2.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Requirements of LCO A.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months not met.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.4.1 Verify each ~S loop is in operation. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months CEOG STS 3.4-7 Rev 1, 04/07/95

Revised 03/18/99

~S Pressure, Tempe~re, and Flow.J'oN~Limits 8 3.4.1 BASES SURVEILLANCE SR 3 .4.1. 2 REQUIREMENTS (continued) srnce e l re c on . a ows a omp on 1 me 0 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months o restore parameters that are ot within limits, the 12 our Surveillance Frequency for. cold leg temperature is suf icient to ensure that the RCS oolant temperature can be re tored to a normal operation, eady state condition foll ing load changes and other e ected transient ope tions. The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months interval as been shown by op ating practice to be suffici t to regularly assess for p ential degradation and to ver fy operation is within fet anal sis assum tions.

@ The 12 h ur Surveillance Fr uency for RCS total flo rate is perf rmed using the inst lled flow instrumentati n. The 12 hou Frequency has been shown by operating expe ience to be z:u icient to assess f r potential degradation and to veri operation is withjn safety analysis assum ions.

. SR is modified by a Note that only requir performance of this SR in MODE 1. The Note is necessary allow

.asurement of RCS f w rate at normal opera ng conditions t ower with all R SR 3. 4.1 a5 l a ow e ins a w s r me ca brated nd verifies that the actual within the bounds of the anal ses.

The Frequency of ;(1~months reflects the importance of verifying flow after a refueling outage where the core has been altered, which may have caused an alteration of flow

@> [-klf\ u resistance. ')" . 3/ EfPD The SR is modified by a ~ tha~states the SR is only

required to be performed 14.f"hoYifs(after ~..9(90,1% RTP. The

(mm 2/ (continued)

CEOG STS 8 3.4-5 Rev 1, 04/07/95

~,

Revised 03/18/99

SECTION 3.4 INSERT 1 PCS flow rate must also be verified after plugging 10 or more steam generator tubes since plugging 10 or more tubes could result in an increase in PCS flow resistance. Plugging less than 10 steam generator tubes will not have a significant impact on PCS flow resistance and, as such, does not require a verification of PCS flow rate.

INSERT 2 The most common, and perhaps accurate, method used to perfonn the PCS total flow surveillance is by means of a primary to secondary heat balance (calorimetric) with the plant at or near full rated power. The most accurate results for such a test are obtained with the plant at or near full power when differential temperatures measured across the reactor are the greatest. Consequently, the test should not be performed until reaching near full power (i.e., 2.. 90% RTP) conditions. Similarly, test accuracy is also influenced by plant stability.

In order for accurate results to be obtained, steady state plant conditions must exist to permit meaningful data to be gathered during the test. Typically, following an extended shutdown the secondary side of the plant will take up to several days to stabilize after power escalation. It is impracticable to perform a primary to secondary heat balance of the precision required for the PCS flow measurement until stabilization has been achieved. Furthermore, an integral part of the PCS flow heat balance involves the use of Ultrasonic Flow Measurement equipment for measuring steam generator feedwater flow. This equipment requires,* stable plant operation at or near full power conditions before it can be used. As such, ...

B 3.4-5

---~

Revised 03/18/99

f~s Minim~emperature for Criticality B 3.4.2 BASES 1"r,'(. ho+be.ro fct.v~ value.. of 5.32..°F a."cl LCO The LCO fts inly ijlpl~a/f~e 0

~eww [535) ~provides a (continued) reasonable distance e im t of;t'52.fi . This allows a equa e ime o rend its approach and 5 take corrective actions prior to exceeding the limit.

APPLICABILITY a.."d C4.n mt bt rc.stoo.J 1n .Jo m11111rv ACTIONS A:l ~~ ~rtr. Kc.~ 1.l.n If T8 --'-** ~~ ~ the plant must e brought to a is below~F, ..

MODE vfn which the LCO does not apply. To achieve this status, the pl ant must be brought to MODE. within 30 minutes. Rapid reactor shutdown can be readily and practically achieved within a 30 minute period. The allowed time reflects the ability to perform this action and to maintain the plant w~thin the analyzed range.

SURVEILLANCE SR 3.4.2.1 REQUIREMENTS m nu es.

IS' Tf .z 7 R.3 7.J\JSl..RT---==,

@ . is is REF~RENCES 1. FSAR, Section m' fL.f.1. 3 .

CEOG STS B 3.4-8 Rev 1, 04/07/95 Revised 03/18/99

A RCS Loops-HODES 1 and 2 W' B 3.4.4 BASES APPLICABLE aspect for this LCO is the @C:fQrf cog l ant forced fl ow rate, SAFETY ANALYSES which is represented by the number off'<BCS loops in service.

(continued) f$li5licyLoo§z-M°2EI J1 and ateQLD 2 satisfy Criteri~ 2 and 3 of(ft(e ijBt)

ID c.FP. so.~lP Cc.I (2.1. .

LCO The purpose of this LCO is to require adequate forced flow for core heat removal. Flow is represented by having both f '8CS loops with bothfftCPs in each loop in operation for removal of heat by the two SGs. To meet safety analysis acceptance criteria for ONB, four pumps are required at rated power. p

' Each OPERABLE loop consists of two @;Ps providing forced flow for heat transport to an SG that is OPERABLE in accordance with the Steam Generator Tube Surveillance Program. SG, and hence RCS loop, OPERABILITY with regard to SG water*level is ensured by the Reactor Protection System (RPS) in HODES 1 and 2. A reactor trip places the plant in (continued)

CEOG STS B 3.4-lS Rev 1, 04/07/95 Revised 03/18/99

9f ~s Loops-MODES B1 3.4.4 and 2 BASES

~s, ~ 0/-o (~q,rrOloel ,.,,..~,

CD LCO MODE 3 if any SG level is ~ [_.?$]% as sensed by the RPS. The (continued) minimum water level to. declare the SG OPERABLE is~]%. __)

~.ct o/o fb,rr~ flol' W 04-r ~(..

APPLICABILITY In MODES 1 and 2, the reactor <l::S)critical and thus has the potential to produce maximum THERMAL POWER. Thus, to ensure that the assumptions of the accident analyses remain valid, allP~s loops are required to be OPERABLE and in operation in tnese MODES to prevent DNB and core damage.

The decay heat production rate is much lower than the full power heat rate. As such, the forced circulation flow and heat sink requirements are reduced for lower, noncritical MODES as indicated by the LCOs for MODES 3, 4, 5, and 6.

Operation in other MODES is covered by:

Cf LCO 3.4.5, 'IS Loops-MOOE 3";

LCO 3.4.6, ' S Loops-MODE 4";

LCO 3.4.7, ' CS Loops-MODE 5, Loops Filled";

LCO 3.4.8, ' S Loops-MODE 5, Loops Not Filled";

LCO 3.9.4, "Shutdown Cooling (SOC) and Coolant Circulation-High Water Level" (MODE 6); and LCO 3.9.5, "Shutdown Cooling (SOC) and Coolant Circulation-Low Water Level" (MODE 6).

ACTIONS If the requirements of the LCO are not met, the Required Action is to reduce power and bring the plant to MODE 3.

This lowers power level and thus reduces the core heat removal needs and minimizes the possibility of violating DNB limits. It should be noted that the reactor will trip and place the plant in MODE 3 as soon as the RPS senses less than four RCPs operaiing.

' The Completion Time of 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months is reasonable, based on operating experience, to reach MODE 3 from full power conditions in an orderly manner and without challenging safety srtems.

(continued)

CEOG STS B 3.4-19 Rev 1, 04/07/95 Revised 03/18/99

ATTACHMENT 6 JUSTIFICATION FOR DEVIATIONS SPECIFICATION 3.4.2, RCS MINIMUM TEMPERATURE FOR CRITICALITY Change Discussion

6. The Applicability of ITS LCO 3.4.2 and the Frequency of SR 3.4.2.1, as well as their associated Bases discussions, have been revised. The LCO now applies whenever the plant is in MODE 1 or MODE 2 with ~ff ~ 1.0 regardless of PCS temperature. The SR Frequency has been changed from "30 minutes thereafter" to "12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ."

Although the change is effectively more restrictive (i.e., the Applicability includes all of MODE 1 and the performance of SR 3 .4.2.1 must continue even above 535 °F) the intent of the change is to correct the current presentation which could lead to an inadvertent violation of the Frequency for SR 3 .4.2.1. Restating the Applicability such that an unexpected decrease in PCS temperature (e.g., during a plant startup) would not result in an SR 3.0.1 violation, and establishing the Frequency of SR 3.4.2.1 such that it does not divert operators during the performance of critical plant evolutions to fulfill data logging requirements, is considered to be a benefit to overall plant safety.

The change in Applicability for ISTS 3.4.2 in NUREG-1432 is consistent with the Applicabilities stated for ISTS 3.4.2 in NUREG-1430 (B&W Plants) and NUREG-1431 (Westinghouse Plants). The change to ISTS SR 3.4.2.1 is consistent with NUREG-1432 as modified by proposed TSTF-27, Rev.3. An additional change has been made in the ITS to the Bases of SR 3 .4.2.1. The phrase which states "is frequent enough to prevent the inadvertent violation of the LCO" has been deleted since this statement is not entirely true.

7. Required Action A.1 and its associated Bases discussion have been revised to maintain consistency with the Applicability. The Applicability of ISTS 3.4.2 is MODE 1 and MODE 2 with ~ff ~ 1.0. ISTS Required Action A.1 requires the plant to be placed in MODE 3 which is outside the Applicability requirement. As such, to maintain consistency with the Applicability, Required Action A.1 has been revised to place the plant in MODE 2 with !\:ff~ 1.0. This change is consistent with NUREG-1432 as modified by TSTF-26.

8. The LCO Bases discussion has been modified to be consistent with the Applicability by eliminating the upper plant condition of < 535 °F.

Palisades Nuclear Plant Page 2 of 2 03/18/99

ATTACHMENT 6 JUSTIFICATION FOR DEVIATIONS SPECIFICATION 3.4.4, RCS LOOPS - MODES 1 AND 2 Change 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.

6. The Bases of ITS SR 3.4.4.l have been modified to reflect that installed plant instrumentation is not capable of displaying primary coolant pump flow rate. Total PCS flow is determined by summing the differential pressures across the primary coolant pumps. Therefore, indication of individual pump differential pressure may be used as a method of determining the pump's relative flow as compared to the other operating primary coolant pumps.

Palisades Nuclear Plant Page 1of1 03/18/99

ML18066A425

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