Revision as of 14:57, 23 February 2020

Text

Forwards Responses to NRC Comments Re ITS Section 3.3 & Associated Revs to ITS Sections 1.0,3.3,3.4 & 3.9 of 990126 ITS Conversion Submittal.One Technical Change & Several Editorial Changes Unrelated to NRC Comments,Also Provided
	ML18066A499
Person / Time
Site:	Palisades
Issue date:	06/11/1999
From:	Haskell N; CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.)
To:	; NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML18066A500	List: ML18066A499; ML18066A501;
References
	NUDOCS 9906180115
	Download: ML18066A499 (769)
	v • d • e

{{#Wiki_filter:,.. . A CMS Energy Company Palisades Nuclear Plant Tel: 616 764 2276 27780 Blue Star Memorial Highway Fax: 616 764 2490 Covert, Ml 49043 Nathan L. Has/Jee/I Director, Licensing June 11, 1999 U.S. Nuclear Regulatory Commission Document Control Desk Washington, DC 20555 DOCKET 50-255- LICENSE DPR PALISADES PLANT CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS; RESPONSE TO NRC COMMENTS ON ITS SECTION 3.3, INSTRUMENTATION On January 26, 1998, Consumers Energy Company submitted.a Technical Specification Change Request (TSCR) to convert the Palisades Technical Specifications to closely emulate the Standard Technical Specifications for Combustion Engineering Plants, NUREG-1432. On October 26 and 27, 1998, representatives of the Palisades and NRC staffs met to discuss Section 3.3, Instrumentation, of the conversion submittal. The NRC requested additional information regarding that section in their January 6, 1999, letter transmitting the minutes of the October meeting. This letter provides responses to the NRC comments regarding ITS Section 3.3, and associated revisions to ITS Sections 1.0, 3.3, 3.4, and 3.9 of our January 26, 1998, ITS

~~::::;::::::~::~:::!:::::~:::::;:::::~~:::ti::::~:::: ::;::::view //,

of the allowed values for the Improved Technical Specification project, it was determined that the upper limit of 4.4 psig specified in current Technical Specificatron for Containment High Pressure (CHP) actuation was not consistent with the assumptions of the containment response analyses. Therefore, it is necessary to place a more restrictive requirement on the CHP actuation allowable values, reducing the upper limit from 4.4 to 4.3 psig. 9906180115 990611 PDR ADOCK 05000255 p PDR

The following editorial changes have been made:

1. The term "Control Rod" has been revised to read "Full Length Control Rod," in footnote (a) of ITS Table 3.3.1-1 and in the Bases, where appropriate. This is discussed in DOC A.2.

2. The term "THERMAL POWER" has been revised to read "Wide Range Power in footnotes (b) and (c) of Table 3.3.1-1.

3. The terminology used when referring to the Power Range Nuclear Instrument channels has been revised to consistently use the words "power range excore channels".

'4. ITS SR 3.3.7.1 has been revised to delete the wording "for each instrument channel that is normally energized"; all channels required by LCO 3.3. 7 are normally energized.

5. SR 3.3.8.3 was deleted. No equivalent SR appears in STS; it was added to the January 26, 1998, ITS submittal to emulate a CTS requirement, but was, in fact, redundant to SR 3.3.8.2.

6. Two settings mentioned on Bases page 3.3.5-2 have been revised to match current plant settings. In the third paragraph, "92%" has been changed to "93%", and "0.5 second" has been changed to "0.65 second". These new values will be in Revision 21 of the FSAR.

7. The Section 3.3 Bases have been revised to provide additional detail and to correct irregularities and inconsistencies.

The following Enclosures to this letter have been provided: Enclosure 1 contains: a) answers to the NRC comments and, b) markups of the previously submitted pages which were affected by the response to that comment, showing where revisions have been made. Enclosure 2 contains complete sets of the January 26, 1998, ITS submittal Technical Specifications and Bases, marked to show the changes made by this submittal. Enclosure 3 contains replacement pages for Section 1.0, and instructions for page replacement. The revised pages reflect changes resulting from our response to the NRC question designated "Definitions RAI" in Enclosure 1. A copy of that question and our response has been included in this attachment. Enclosure 4 contains an entire set of replacement pages for Section 3.3. The revised pages reflect changes resulting from our response to the NRC comments and the additional changes identified in Enclosure 2. Due to the number of pages effected, all pages for Section 3.3 of our January 26, 1998 submittal, with the exception of the STS markup pages of Attachment 5, have been replaced. With the exception of hand marked pages, the replacement pages are dated for identification, and the changed text is identified in the margin. Enclosure 5 contains replacement pages for Section 3.4, and instructions for page replacement. The revised pages reflect changes resulting from our response to the NRC question designated "Definitions RAI" in Enclosure 1. A copy of that question and our response has been included in this attachment. 2 _J

Enclosure 6 contains replacement pages for Section 3.9, and instructions for page replacement. The revised pages reflect changes resulting from our response to the NRG question designated "Definitions RAI" in Enclosure 1. A copy of that question and our response has been included in this attachment.

  - - Enclosure 7 contains examples of the instrument channel drift measurements taken in support of the extension of Channel Functional Test intervals from 31 to 92 days.

The NRG letter of January 6, 1999, referred to a date of February 15, 1999 for our response to the RAI. Subsequently, in a telephone conversation with the NRR Project Manager, Consumers Energy received permission to delay the response to allow additional time for preparation and internal review.

SUMMARY

OF COMMITMENTS This submittal contains no new commitments and no revisions to existing commitments. C Administrator, Region Ill, USNRC Project Manager, NRR, USNRC NRG Resident Inspector - Palisades Enclosures 3

~-' CONSUMERS ENERGY COMPANY RESPONSE TO JANUARY 06, 1999 NRC COMMENTS To the best of my knowledge, the content of this response to NRC comments concerning Section 3.3 of our January 26, 1998 License Amendment request for conversion to Improved Technical Specifications, is truthful and complete.

Director, Licensing Sworn and subscribed to before me this f (11-._day of ~ 1999. Lllrma . Van Bu County, Michigan My commission expires May 14, 2002 4

ENCLOSURE 1 CONSUMERS ENERGY COMPANY PALISADES PLANT DOCKET 50-255 CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO JANUARY 6, 1999 NRC COMMENTS CONCERNING ITS SECTION 3.3 RESPONSE TO NRC COMMENTS SECTION 3.3, INSTRUMENTATION

fe.wv~~ -~+k ~ ~

*                                : oG>/11/qq_._qqob(<ilOl\5 1

                                                                                                          *. ~

CONVERSION TO IMPROVED.TECHNICAL SPECIFICATIONS RESPONSE TO JANUARY 6, 1999 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.3, INSTRUilllENTATION NRC REQUEST: (Referred to as "Definitions RAI") Definitions Channel Calibration, Channel Functional Test ,*,* The ITS proposes to adopt the STS definition for channel calibration exc~pt the.tE!rrn',;di~play" is replaced with the term "interlock;" the requirement to cross calibrcite RTDs following sensor element replacement is not adopted and the exclusion of neutron detectors is added to the definition. Comment: Deviations from STS definitions require a staff approved NEltSTF~ The staff

notes that TSTF-205, Rev 2 proposes changes to the channel calibration and channel .

functional test definition which eliminates unnecessary references to specific channel ' components in the definitions and clarifies the Bases. The staff recommends adopting the final form of TSTF-205. . Consumers Energy Response: The STS Channel Functional Test and Channel Calibration definitions, and the conforming bases changes included in TSTF 205 have been adopted. The STS Channel Calibration SR note, excluding neutron detectors from calibration requirements has been added to the affected ITS SRs . The Channel Functional Test Bases addition contained in TSTF 205 has been added to.the Bases for each Channel Functional Test; the text of this addition is shown as Insert A. Insert A has not been added to the bases for SR 3.3.1.8, because the affected equipment, source range nuclear instrument channels, do not provide any automatic actuations. Revised markups of the STS bases have not been included. The bases changes associated with the definition changes are shown on the marked pages associated with this NRC comment. Additional changes have been made to the Bases as a result of other NRC comments and our own reviews. Enclosure 2 contains a marked version of the Bases in our January 26, 1998 submittal which shows all changes to the Bases. Affected Submittal Pages: Att 1 ITS 1.1, page 1.1-2

 . Att 3 CTS, page 1-1 (ITS 1.1, page 1 of 6)

Att 3 CTS, page 1-2 (ITS 1.1, page\ 2 of 6) Att 3 DOC, 1.0, page 2 of 12 Att 3 DOC, 1.0, page 8 of 12 Att 5 NUREG, page 1.1-2 Att 6 JFD *1.0, page 3 of 3

1

                                                                                           * , .~ ~ .: .., :i ~* .: . .:* '

(continued)

                                                                                                                              ,J.* ,

CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO JANUARY 6, 1999 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.3, INSTRUMENTATION Consumers Energy Response: Definitions RAI (continued) Affected Submittal Pages: Att 1 ITS 3.3.1, page 3.3.1-5 Att 1 ITS 3.3.7, page 3.3.7-3 Att 1 ITS 3.3.8, page 3.3.8-2 Att 1 ITS 3.3.9, page 3.3.9-2 Att 2 ITS B 3.3.1, page B 3.3.1-29 Att 2 ITS B 3.3.1, page B 3.3.1-30 Att 2 ITS B 3.3.1, page B 3.3.1-31 Att 2 ITS B 3.3.2, page B 3.3.2-9 Att 2 ITS B 3.3.2, page B 3.3.2-10 Att 2 ITS B 3.3.3, page B 3.3.3-18 Att 2 ITS B 3.3.4, page B 3.3.4-11 Att 2 ITS B 3.3.4, page B 3.3.4-12 Att 2 ITS B 3.3.5, page B 3.3.5-6 (former submittal) Att 2 ITS B 3.3.5, page B 3.3.5-6 Att 2 ITS B 3.3.6, page B 3.3.6-4 Att 2 ITS B 3.3.7, page B 3.3.7-14 Att 2 ITS B 3.3.8, page B 3.3.8-7 Att 2 ITS B 3.3.9, page B 3.3.9-4 Att 2 ITS B 3.3.10, page B 3.3.10-4 Att 5 NUREG, page 3.3-5 Att 5 NUREG, page 3.3-42 Att 5 NUREG, page 3.3-45 Att 5 NUREG, page 3.3-48 Att 6 JFD 3.3.1, page 3 of 4 Att 6 JFD 3.3.11, page 1 of 2 Att 6 JFD 3.3.12, page 1 of 2 Att 6 JFD 3.3.13, page 2 of 3 Att 2 ITS B 3.4.12,* page B 3.4.12-12 Att 2 ITS B 3.4.15, page B 3.4.15-5 Att 1 ITS 3.9.2, page 3.9.2-2 Att 2 ITS B 3.9.2, page B 3.9.2-4 Att 5 NUREG, page 3.9.3 Att 6 JFD 3.9.2, page 1 of 2 2

Definitions 1.1

1.1 Def1 niti ons CHANNEL CALIBRATION

                                                                                                    .1 all devices in the channel required for channel OPERABILITY and Whenever a RTD or thermocouple Calibration of instrument channels with Resistance sensing element is replaced, the        Temperature Detector (RTD) or thermocouple sensors   'DEF' next required CHANNEL CALIBRATION __ may consist of an inplace qualitative assessment         RAI shall include an mplace cross           of sensor behavior and normal calibration of. the calibration that compares the other     r~aining adjustable devices in the channel.

sensing elements with the recently 1----1~ ms tolled sensing* element.

CHANNEL CHECK A CHANNEL CHECK shall be the qualitative a-ssessment, by observation, of channel behavior during operation. This determination shall include, where possible,. comparison of the channel indication and status to other indications or status derived from independent instrument channels measuring the same parameter.

CHANNEL FUNCTIONAL TEST A CHANNEL FUNCTIONAL TEST shall be:

                                        *a. Analog and bistable channels-the injection of a simulated or actual signal into the channel as close to the sensor as racticable to verif OPERABILITY,
                          - - - - - - - - l

ter, OC I
p S of oll devices in the channel

b. Digital channels-the use of diagnostic 'DEF' required for channel OPERABILITY.

programs to test digital hardware and the RAI injection of simulated process data into the channel to verif OPERABILITY, lu ng la ip nc on * *

Palisades Nuclear Plant 1.1-2 Amendment No. 01/20/98 J- ()./

Def in it i ans

1.1 Definitions CHANNEL CALIBRATION A CHANNEL CALIBRATION shall be the adjustment, as 1.1 necessary, of the channel output such that it responds within the necessary range and accuracy to known values of the parameter that the channel monitors. The CHANNEL CALIBRATION shall encompass all devices in the channel required for channel OPERABILITY and the CHANNEL FUNCTIONAL TEST.

Calibration of instrument channels with Resistance Temperature Detector (RTD) or thermocouple sensors may consist of an inplace qualitative assessment of sensor behavior and normal calibration of the remaining adjustable devices in the channel. Whenever an RTD or thermocouple sensing element is replaced, the next required CHANNEL CALIBRATION shall include an inplace cross calibration that compares the other sensing elements with the recently installed sensing element. The CHANNEL CALIBRATION may be performed by means of any series of sequential, overlapping, or total channel steps. CHANNEL CHECK A CHANNEL CHECK shall be the qualitative assessment, by observation, of channel behavior during operation. This determination shall -i include, where possible, comparison of the channel indication and status to other indications or status derived from independent instrument channels measuring the same parameter. CHANNEL FUNCTIONAL TEST A CHANNEL FUNCTIONAL TEST shall be:

a. Analog and bistable channels-the injection of a simulated or actual signal into the channel as close to the sensor as practicable to verify OPERABILITY, of all devices in the channel required for channel OPERABILITY;

b. Digital channels-the use of diagnostic programs to test digital hardware and the injectidn of simulated process data into the channel to verify OPERABILITY, of all devices I

Palisades Nuclear Plant in the channel required for channel OPERABILITY.

                                   / - 'fJ 1.1-2"""          Amendment No. 05/30/99 I

I

Specification 1.1

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

The defined terms of this section

  ~-----< appear* in capitalized type and are applicable throughout these Techmcol Specifications and Bases
   ~---------~                    TECHNICAL SPECIFICATIONS j 1.0  Use and Application j

l.@1) DEFINITIONS The following tel'lls are defined for unifonn interpretation of these Technical Specifications. * *

         @<Insert ACTIONS DEFINITION as presentrd in the ITS>

ASSEM8LY RADIAL PEAKING FACTOR - Fr core (_in divi duo 1 f ue l ossem bl y )_ ASSEMBLY RADIAL PEAKING FACTOR shall be t e maximum ratio of theipower ener e a u ass to the average fuel assembly power. ac t se er, erm s a e integrated over~core height~* a inc u *. _the totol ) _ AYERAGE QISINTEGRATION ENERGY - E AVERAGE DISINTEGRATION ENERGY shall be the average (weighted in proportion to the concentration of each radionuclide in the reactor coolant at the ti*e of sampling) of the sum of the average beta and ganna energies per disintegration (in MEY) for isotopes, other than iodines, with half lives greater t'han 15 minutes, making up at least 9S% of the total noniodine activity in the coolant .

AXIAL OFFSET~APE

_I_ Split into separate def1mt1ons INPEX - AO 12!JASI.~~ AXIAL OFFSET or AXIAL SHAPE INDEX shall be e r io s generated in the lower half of the core nus the power generated in the upper half of the core, o the sum o the power* ose powe * "'"g_e_n_e_r-at.,...e--...-1n-~~----,-------. hol ves of the d1v1ded by CHANNEL CALIBRATION all devices in the chorinel required for channel operability ond

 'DEF' RAI

Amendment No. **

1-1

                                             .a, 54, J-c...,

5-1-, 6B, -Ha, tr4, t%8, -HT, ~. 174

Specification 1.1

b. Digi tol chonnels - the use of diagnostic programs to test digi tol hardware and the injec.tion of simulated process data into the channel to ven f y OPERABILITY of oll devices in the chonnel required for channel OPERABILITY.

                                                                                                       'DEF' RA!

1-@U) PEFINITIONS {continued) CHANNEL FUNCTIONAL TEST a.Analog and bistable channels - A CHANNEL FUNCTIONAL TEST shall b he injection of a simulated ~inal or d¥t~ :lle channel to verifZ that it is OPERABLE. f1ncyu'd1"9/any/ala anaj _i __ itij{ing,.func)ion._ Z 7 7 Z L_ L_ L_ LJ os close to the sensor as practicable 'of all devices in the channel COLP SHUTDOWN required for channel OPERABILITY; MODE 5 The COLD SHUTDOWN condition shall be when the primary coolant SHUTDOWN BORON CONCENTRATION and Tave is less than 210°F . CORE OPERATING LIMITS REPORT {COLR) The COLR is the document that provides cycle specific parameter limits for the current reload cycle. These cycle specific parameter limits

         ~hall be detennined for each reload cycle in accordance with Specification~6.5. Plant operation within these limits is addressed in individual pecifications.

5 POSE EQUIVALENT I-131 DOSE EQUIVALENT I-131 shall be that concentration of I-131 {µCi/gm) which alone would produce the same thyroid dose as the quantity and isotopic mhture of I-131, I-132, I-133, J-134 and I-135 actually present. The thyroid dose conversion factors used for this calculation shall be those listed in Table III of TID-14844,~~calculation of

Distance Factors for Power and Test Reactor Sites."

1-2 I -J AEC, 1'162 @ Amendment No. 3-t, .a, 5'4 51, 68, -Ha, -i-2-4, tra, H-r. +6-r,-tt-'4, 184 1

A.4 ATTACHMENT 3 DISCUSSION OF CHANGES CHAPTER 1.0, USE AND APPLICATION The CTS definition of CHANNEL CHECK states "A CHANNEL CHECK shall include verification that the monitored parameter is within limits imposed by the Technical Specifications." This sentence was originally added to the CTS to address the problem wherein the TS contained requirements that various parameters be within a particular limit but there was not a corresponding surveillance requirement specified to verify the limit was being inet. By adding these words to the definition of CHANNEL CHECK in the CTS, the CHANNEL CHECK would not only verify channel operability but also that the parameter was within limits.* By adopting TS which are modeled after .NUREG-1432, there is no need to have this "surveillance requirement" specified as part of the CHANNEL CHECK requirement since there will be a separate surveillance requirement specified which requires that the parameter be verified within limit. Therefore, there is no change in requirements, only in presentation of*

requirements and this is considered to be an administrative change. This change is consistent with NUREG-1432. A.5 The CTS definition of CHANNEL FUNCTI~NAL TEST is expanded in the ITS to provide further descriptive information for Analog and bistable channels, and to add a for discussion digital channels. To address digital channels, the following wording is added to the definition for CHANNEL FUNCTIONAL TEST: "the use of diagnostic programs to test digital hardware and the injection of simulated rocess data

channel to verify OPERABILITY, i * *
s This section is added to specify the appropriate requirements for digital equipment w,hich has bee added to the original plant design. o<= "-'-' ~11.vlc...,.,,.,_
1"1 -r~c~ c::.!~t~~~ILI Ti

                                                                     *tte&vl1t*t:.,..... Fo"- C.H'~"'""'t.L U(   1 The existing CTS definition relates to Analog and bistable channe.ls and has been u~nJ expanded in the proposed ITS to further describe components of a channel by ali91Ag the wording '                       *             " The hrase 11 as close to the sensor as practicable 11 is added following the phrase 11 into the channel" to make it clear where the simulated signal is to be injected. The v.*erEI "r@ruin~8" is added f"f'ier te the phrase "alarm and trip" ta wake it .clear tbat oaly* tlie f"ertiens ef the chaooel '>'.'Rich an1 "re'}Yiretl"-auist meet the CIIANNEL FUNCTIONAL TES'F:

The phrase" The CHANNEL FUNCTIONAL TEST may be performed by means of any series of sequential, overlapping, or total channel steps so that the entire channel is tested. 11 is als.o added to the CTS to provide clarification that as long as the entire channel is tested, the testing can b u into different tests.

                                       '"J:)ti,.Fl t'o.( I Tl~~

12'.AI

Palisades Nuclear Plant CH~r.Jrs.te.t;,,

Page 2of12

                                                         /-e_

01/20/98 '

A.17 ATTACHMENT 3 DISCUSSION OF CHANGES CHAPTER 1.0, USE AND APPLICATION .

CTS definition of "OPERABLE-OPERABILITY" includes "electrical power" but does not specify whether it is normal or emergency power. In the proposed ITS definition for "OPERABLE" the words "normal or emergency" are added to clarify that either is acceptable for determining OPERABILITY. Section 3.8 of the proposed ITS addresses actions to take on loss of an off-site circuit or emergency diesel generator and any other

       . actions which must be taken for the supported systems. This structure clarifies that there is no need to declare all supported equipment from electrical power sources inoperable upon loss of either the normal or emergency power source.

The addition of the words "normal or emergency" is considered an administrative change since it simply clarifies the existing application of the CTS definition of "OPERABLE-OPERABILITY;" This change is consistent with NUREG-1432. A.18 The CTS definition of Quadrant Power Tilt - Tq states in part " ... shall be the algebraic ratio .... " The word algebraic is replaced in the proposed ITS with "maximum positive." This is an administrative change to more correctly reflect that the Quadrant Power Tilt Ratio will be expressed in terms of a positive value .

A.19 The CTS does not include explanatory material related to logical connectors, completion times, or frequencies. The proposed ITS adds a discussion of each of these topics to standardize the use and application of the TS. The proposed sections to be focluded in the IT5 are 1.2, Logical Connectors, 1.3 Completion Times, and 1.4 Frequencies. The addition of this information is considered to be an administrative change since it is simply explaining the rules which are used to develop and use the ITS. This change is consistent with NUREG-1432.

A.20 The CTS definition for "Channel Calibration" states in part " ... The CHANNEL CALIBRATION shall encompass the entire channel including the sensor, alarm interlock, .... " The proposed ITS .tis~~7i the wordS" e

d" pri~ensor" at t~pose ads " ... s ncompass th mire cham<ef includi the e rured sensor alarm, interlock .. " This change is made to clarify that only the ~~'3l~~ "Tl'/

          "required" components in the channel (meaning those                           *

aaaly-si~ must have~he channel calibration. This is an administrative change since the IA.)e~~' "(.., I'> . C.1-1~ &ii.-"'?. . fi . f .

wererr@Ej\iif@(T 1s 61'\

a&aes for clan 1cat1on o the reqmrements and does not change the requirements themselves. This change is consistent with NUREG-1432, ,,..~ ,.no;>iFl!f:."'=>

           ~1 TSTF .2C~

Palisades Nuclear Plant Page 8of12 01120/98 1--/y

TSTF- ~ Definitions 205 ~ 1.1 ALL "DE-VIC. c-s I~ T+\E:- C.'MA.tJ ~'E.-L 'R°E-QL) IK"'E'D

1.1 Definitions

 -------=-==========~=====~-

CHANNEL CALIBRATION (continued)° t=os<. C+l.A.t.ll--l'E:-L enti , ch arm, is 1 OPE-RA.B\\....ITY iW..lO a tri func ons and DE.F

                                                                                                        ~A.'I incl e the CHANNEL FUNCTIONAL TEST. Calibration of instrument channels with resistance temperature detector {RTD) or thermocouple sensors may consist of an inplace qualitative assessment of sensor
                                                                                                    @l behavior and normal calibration of the remaining lZ\D oR adjustable devices in the channel. Whenever a °THeeMocooFlE:

sensing element 1s replaced, the next required'----- CHANNEL CALIBRATION shall include an inplace cross calibration that compares the other sensing elements with the recently installed sensing element. The CHANNEL CALIBRATION may be performed by means of any series of sequential, overlapping, or total channel steps so that the entire channel is calibrated. CHANNEL CHECK A CHANNEL CHECK shall be the qualitative assessment, by observation, of channel behavior during operation. This determination shall include, where possible, comparison of the channel indication and status to other indications or status derived from independent instrument channels measuring the same parameter. CHANNEL FUNCTIONAL TEST A CHANNEL FUNCTIONAL TEST shall be:

         -rsn=-                  a.                                                 injection of 2D'5
      © OF- ALL oev1c.es Ir..!
     '"'E:   C.\.l.A.~1'.lE-L    b.

R1C:QU\RE-t:> t=oe oPe-eA6\LIT'/ JI DE:,J:- l<.6-J: The CHANNEL FUNCTIONAL TEST may be performed by means of any series of sequential, overlapping, or total channel steps so that the entire channel is

tested .
CEOG STS 1.1-2

{continued) Rev 1, 04/07/95 i-zr

14.

ATTACHMENT 6 JUSTIFICATIONS FOR DEVIATIONS CHAPTER 1.0, USE AND APPLICATION The Palisades CTS contains the term "Quadrant Power Tilt (Tq)" and this term is also included in the proposed ITS. The Quadrant Power Tilt is defined as "Tq shall be the maximum positive ratio of the power generated in any quadrant minus the average quadrant power, to the average quadrant power."

15. The wording of the Identified Leakage definition has been altered to clarify that leakage which might affect the operation of leakage detection systems must be classified as unidentified leakage. It is believed that this is the intent of the STS definition .

Palisades Nuclear Plant Page 3of3

                                             /- ~

10/10/98

INSERT Section 1.0, ATT 6, JFDs Page 3 of 3

16. The Channel Functional Test and Channel Calibration definitions have been revised to reflect the changes made by TSTF 205.

17. The words 11 RTD or thermocoupl e 11 have been added to the forth sentence of the Channel Calibration definition to assure it is understood that the requirements of that sentence only apply to those types of sensor. This is an editorial change made for clarification only .

RPS Instrumentation 3.3.1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.3.1.7 Perfonn a CHANNEL FUNCTIONAL TEST of High Once within Startup Rate and Loss of Load Functions. 7 days prior to each reactor startup SR 3.3.1.8 18 months

                  --------1\loTE..--------
                    ~E.\lTROt-l "DE.TE'-tt>l2!S. ARE'. E.~Lllt>E.1:> 'F~Ot--\

T~ C~A.U"1E:'L CAL1\3eAT\ot.,l, Palisades Nuclear Plant Amendment No. 01/20/98

PAM Instrumentation 3.3.7 SURVEILLANCE REQUIREMENTS

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

These SRs apply to each PAM instrumentation Function in Tabl~ 3.3.7-1.

 ----------------------------~-------------------------------------------------

SURVEILLANCE FREQUENCY SR 3.3.7.1 31 days ET? SR 3.3.7.2 Perform CHANNEL CALIBRATION. 18 months

                                                                              ,, II DE-i:h
                  -'-~~~oT~~~~~~~

eA"J:

~ELlir2o~ t:>E.rt=c..loi::.S A~ E')(c.LLJt>a>*

Frc!01'1\ ~e. C.J.\A~~EL CA.L\~A11otJ .

/-~

Palisades Nuclear Plant 3.3.7-3 Amendment No. 01/20/98

Alternate Shutdown System

SURVEILLANCE REQUIREMENTS SURVEILLANCE 3.3.8 FREQUENCY SR 3.3.8.1 Perform CHANNEL FUNCTIONAL TEST of the Once within Source Range Neutron Flux Function. 7 days prior to each reactor startup SR 3.3.8.2 Verify each required control circuit and 18 months transfer-switch is capable of performing the intended function.

E-D Perform CHANNEL CALIBRATION for each 18 months required instrumentation channel. II DF::F"

2. WElJTeoN D'E:TEC...1""0 ttc; Aet:: e-.(LLlH:>E*t> E',A:L

                    ~oM TI\c- CAAtJME:.L CAUBEAT'tDIJ .

/-J Amendment No. 01/20/98 Palisades Nuclear Plant 3.3.8-2

Neutron Flux Monitoring Channels 3.3.9 SURVEILLANCE REQUIREMENTS - SURVEILLANCE FREQUENCY SR 3.3.9.1 Perfonn CHANNEL CHECK. 12 hours SR 3.3.9.2~rfonn CHANNEL CALIBRATION. 18 months I ,, NolE... - -------- IJ E(.)re!> ~ t:E-.re.crce ~ A~ C:~t..UDE:b

                   'F-~OH THE- ~At.JIJE"L CAU/SK:Ai10AJ.

J-rri Palisades Nuclear Plant 3.3.9-2 Amendment No. 01/20/98
BASES INSERT A Add to each Channel Functional Test Bases.

A successful test of the required contact(s) of a channel relay may be performed by the verification of the change of state of a single contact of the relay. This clarifies what is an acceptable CHANNEL FUNCTIONAL TEST of a relay. This is acceptable because all of the other required contacts of the relay are verified by other Technical Specifications and non-Technical Specifications tests at least once per refueling interval with applicable extensions .

/-fl

RPS Instrumentation B 3.3.1 BASES SURVEILLANCE SR 3.3.1.5 REQUIREMENTS (continued) A CHANNEL FUNCTIONAL TEST is performed on each RPS instrument channel, except Loss of Load and High Startup Rate, every 92 days to ensure the entire channel will perform its intended function when needed. For the TM/LP Function, the constants associated with the thermal margin monitors must be verified to be within tolerances. IJ.(S4C.\ fl. ~ lEP*"'n~ Bistable Tests The bistable setpoint must be found to trip within the Allowable Values specified in the LCO and left set consistent with the assumptions of the setpoint analysis.

                   *A  tes~ signal is superimposed on the input in one channel at a time to verify that the bistable trips within the specified tolerance around the setpoint. This is done with the affected RPS channel (trip channel) bypassed. Any setpoint adjustment must be consistent with the assumptions of the current setpoint analysis.

The as found and as left values must also be recorded and reviewed for consistency with the assumptions of the Frequency extension analysis. The requirements for this review are outlined in Reference 5. The Frequency of 92 days is based on the reliability analysis presented in topical report CEN-327, "RPS/ESFAS Extended Test Interval Evaluation" (Ref. 5). SR 3.3.1.6 A calibration of the power range excore channels using the internal test circuitry is required every 92 days. The as found and as left values must also be recorded and reviewed for consistency with the assumptions of the frequency extension analysis. The requirements for this review are outlined in Reference 5. Palisades Nuclear Plant B 3.3.1-29 01/20/98

RPS Instrumentation B 3.3.1

BASES

 ,SURVEILLANCE      SR 3.3.1.6    (continued)

REQUIREMENTS The neutron detectors are excluded from calibration because they are passive devices with minimal drift and because of the difficulty of simulating a meaningful signal. Slow changes in detector sensitivity are compensated for by perfonning the daily calorimetric calibration (SR 3.3.1.3) and the monthly linear subchannel gain check (SR 3.3.1.4). In addition, associated control room indications are continuously monitored by the operators. The Frequency of 92 days is acceptable, based on plant operating e~perience, and takes into account indications and alanns available to the operator in the control room. SR 3.3.1.7 A CHANNEL FUNCTIONAL TEST on the Loss of Load and High Startup Rate channels is perfonned prior to a reactor

start_yp to ensure the entire channel will perfonn its intended function if require(h. The High Startup Rate trip is actuated by either of the Wide Range Nuclear Instrument Startup Rate channels. NI-01/03 sends a trip signal to RPS channels A and C; NI-02/04 to channels B and D. Sine~ each High Startup Rate channel would cause a trip on two RPS channels, the High Startup Rate trip is not tested when the reactor is critical. The High Startup Rate trip Function is required during startup operation and may be operationally bypassed when below lE-4% RTP or above 13% RTP.

The four Loss of Load Trip channels are all actuated by a single pressure switch monitoring turbine auto stop oil pressure which is not tested when the reactor is critical. Operating experience has shown that these components usually pass the Surveillance when perfonned at a Frequency of once per 7 days prior to each reactor startup .

Palisades Nuclear Plant 8 3.3.1-30 01/20/98

RPS Instrumentation B 3.3.1 BASES SURVEILLANCE SR 3.3.1.8 REQUIREMENTS (continued) SR 3.3.1.8 is the performance of a CHANNEL CALIBRATION every 18 months. CHANNEL CALIBRATION is a complete check of the instrument channel including the sensor. The Surveillance verifies that the channel responds to a measured parameter within the necessary range and accuracy. CHANNEL CALIBRATION. leaves the channel adjusted to account for instrument .drift between successive calibrations to ensure that the channel remains operational between successive tests. CHANNEL CALIBRATIONS must be performed consistent with the plant specific setpoint analysis. The as found and as left values must also be recorded and reviewed for consistency with the assumptions of the frequency extension analysis. The requirements for this review are outlined in Reference 5 . As part of the CHANNEL CALIBRATION of the wide range Nuclear Instrumentation, automatic removal of the ZPMB for the Low PCS Flow, TM/LP, and Low SG Pressure trips and the automatic (operational) bypassing of the Loss of Load and High Startup Rate trips must be verified to assure that these trips are

                    ~vailable when required.

The Frequency is based upon the assumption of an 18 month calibration interval for the determination of the magnitude of £quipment drift. The neutron et ctors are excl ded from CH NNEL CALIBRATION because the re passive ces with

imal drift because the difficu of simula
g a meanin signal.

Slow c nges in det or sensiti

y are comµ, sated for by per rming the y calorime ic calibrat* n (SR 3.3.1.3) a the mont linear sub annel gain check (SR 3.3.1.4) .
I I

Palisades Nuclear Plant B 3.3.1-31 01/20/98

Definitions RAI 3.3 BASES INSERT J Add to Bases for SR 3.3.1.

This SR is modified by a Note which states that it is not necessary to calibrate neutron detectors because they are passive devices with minimal drift and because of the difficulty of simulating a meaningful signal. Slow changes in power range excore neutron detector sensitivity are compensated for by perfonning the daily calorimetric calibration (SR 3.3.1.3) and the monthly calibration using the incore detectors (SR 3.3.1.4). Sudden changes in detector perfonnance would be noted during the required CHANNEL CHECKs (SR 3.3.1.1) .

)- )L

RPS Logic and Trip Initiation B 3.3.2 BASES ACTIONS E.1. E.2.1 and E.2.2 (continued) Required Actions E.2.1 and E.2.2 allow 6 hours to verify that no more than one control rod is capable of being withdrawn or to verify that PCS boron concentration is at REFUELING BORON CONCENTRATION. The Completion Time is reasonable to place the plant in an operating condition in which the LCO does not apply. SURVEILLANCE SR *3.3.2.1 REQUIREMENTS A CHANNEL FUNCTIONAL TEST on each RPS Logic channel is performed every 92 days to ensure the entire channel wi 11 rr perform its i.ntended function when needed. 11 _ _ _ _ _ _ ._:~

              ~        This SR addresses the two tests associated with the RPS

.___~ Logic: Matrix Logic and Trip Initiation Logic .

Matrix Logic Tests These tests are performed one matrix at a time. They verify that a coincidence in the two input channels for each Function removes power from the matrix relays. During

                      *testing. power is applied to the matrix relay test coils and prevents the matrix relay contacts from assuming their de-energized state. The Matrix Logic tests will detect any short circuits around the bistable contacts in the coincidence logic such as may be caused by faulty bistable relay or trip channel bypass contacts.

Trip Initiation Logic Tests These tests are similar to the Matrix Logic tests. except that test power is withheld from one matrix relay at a time, allowing the initiation circuit to de-energize, de-energizing the affected set of clutch power supplies. The Frequency of 92 days is based on the reliability analysis presented in'topical report CEN-327, "RPS/ESFAS Extended Test Interval Evaluation (Ref. 5) . 11 Palisades Nuclear Plant B 3.3.2-9 01/20/98

RPS Logic and Trip Initiation B 3.3.2 BASES SURVEILLANCE SR 3.3.2.2 REQUIREMENTS \ (continued) A CHANNEL FUNCTIONAL TEST on the Manual Trip channels is . perfonned prior to a reactor startup to ensure the entire INS,tll."\ Pr - channel will perfonn its intended function if reguired."The Manual Trip Function is not tested at power. However, the simplicity of this circuitry and the absence of drift

 -i;>eF*"' 'T\  o"' RA-I concern makes this Frequency adequate. Additionally, operating experience has shown that these components usually pass the Surveillance when perfonned once within 7 days prior to each reactor startup.

REFERENCES 1. 10 CFR 50, Appendix A

2. 10 CFR 100 3* FSAR,. Figure 7-1

4. FSAR, Section 7.2

5. CEN-327, June 2, 1986, including Supplement 1, March 3, 1989

J-t Palisades Nuclear Plant B 3.3.2-10 01/20/98

ESF Instrumentation B 3.3.3 BASES SURVEILLANCE SR 3.3.3.1 (continued) REQUIREMENTS Agreement criteria are detennined by the plant staff based . on a combination of the channel instrument uncertainties, including indication and readability. If a channel is outside the criteria, it may be an indication that the sensor or the signal processing equipment has drifted outside its limit. If the channels are within the criteria, it is an indication that the channels are OPERABLE. If the channels are nonnally off scale during times when Surveillance is required, the CHANNEL CHECK will only verify* that they are off scale in the same direction. Offscale low current loop channels are verified to be reading at the bottom of the range and not failed downscale.. The Frequency of about once every shift is based on operating experience that demonstrates channel failure is rare. Since the probability of two random failures in

- redundant channels in any 12 hour period is extremely low, the CHANNEL CHECK minimizes the chance of loss of protective function due to failure of redundant channels. The CHANNEL CHECK supplements less fonnal, but more frequent, checks of CHANNEL OPERABILITY during nonnal operational use of displays associated with the LCD required channels.

SR 3.3.3.2 A CHANNEL FUNCTIONAL TEST is perfonned every 92 days to ensure the entire channel will perfonn its intended function A__ when needed.

               ----=;;:a-::..

This test is required to be perfonned each 92 days on ESFI input channels provided with on-line testing capability. It is not required for the SIRWT Low Level channels since they have no built in test capability. The CHANNEL FUNCTIONAL TEST for SIRWT Low Level channels is performed each 18 months as part of the required CHANNEL CALIBRATION. The CHANNEL FUNCTIONAL TEST tests the individual sensor subsystems using an analog test input to each bistable. A test signal is superimposed on the input in one channel at a time to verify that the bistable trips within the

specified tolerance around the setpoint. Any setpoint adjustment shall be consistent with the assumptions of the current plan~ specific setpoint analysis.

Palisades Nuclear Plant 8 3.3.3-18 01/20/98

ESF Logic and Manual Initiation B 3.3.4

BASES ACTIONS C.l and C.2 (continued)

Condition C is entered when one or more Functions have two Manual Initiation, Bypass Removal, or Actuation Logic channels inoperable for Functions 2 or 3, and when the Required Action and associated Completion Time of Condition A are not met for Functions 2 or 3. If Required Action A.1 cannot be met within the required Completion Time, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 6 hours and to MODE 5 within 36 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems. SURVEILLANCE Refer to Table 3.3.4-1 to determine which SR shall be REQUIREMENTS performed for each Function.

SR 3.3.4.1 A functional test must be performed both with and without offsite power; When testing the "without power" circuits, proper operation of the OBA sequence and the associated logic circuit must be verified. The test circuits are designed to block those SIS functions, such as injection of concentrated boric acid, which would interfere with plant operation.

The Frequency of 92 days is based on the reliability analysis presented in topical report CEN-327, "RPS/ESFAS Extended Test Interval Evaluation 11 (Ref. 2). SR 3.3.4.2 A CHANNEL FUNCTIONAL TEST is performed every 92 days to ensure the entire channel will perform its intended function when needecl:,.)fSensor subsystem tests are addressed in

LCO 3.3.3. This SR addresses Actuation Logic tests of the AFAS using the installed test circuits .
Palisades Nuclear Plant B 3 .3. 4-11 01/20/98

ESF Logic and Manual Ini ti ati on B 3.3.4

BASES SURVEILLANCE SR 3.3.4.2 (continued)

REQUIREMENTS This SR is modified by a Note which states that Actuation Logic tests include operation of initiation relays. The Frequency of 92 days is based on the reliability analysis presented in topical report CEN-327, "RPS/ESFAS Extended Test Interval Evaluation" (Ref. 2). SR 3.3.4.3 A CHANNEL FUNCTIONAL TEST is performed on the manual ESF actuation circuitry and Actuation Logic for certain ESF Functions, providing actuation of the Function. 1~> A ~--~.-. This Surveillance verifies that the trip push buttons of the Manual Initiation Function are ~apable of opening contacts in the Actuation Logic as designed, providing Manual Initiation of the Function. Thi~ Surveillance also verifies

that the entire channel of the Actuation Logic will perform its intended function when needed. The 18 month Frequency is based on the need to perform this Survei 11 ance under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power. Operating experience has shown these components usually pass the Surveillance when performed at a Frequency of once every 18 months.

REFERENCES 1. FSAR, Chapter 7

2. CEN-327, June 2, 1986, including Suppl~ment 1, March 3, 1989

Palisades Nuclear Plant

                                      /-GJ B 3.3.4-12                          01/20/98

DG - UV Start B 3.3.5

BASES ACTIONS A. I (continued)

Required Action A.I ensures that Required Actions for the affected DG inoperabilities are initiated. Depending upon plant MODE, the actions specified in LCO 3.8.I or LCD 3.8.2, as applicable, are required immediately. SURVEILLANCE The following SR applies to each DG - UV Start Function. REQUIREMENTS ~

                   /      SR   ~3.3.5,(i e'"',.~"*~

h,Mc..~ 0~,~ SR 3.3.5.I is the performance of a CHANNEL CALIBRATION every

 -r._.;r              '   18 months. The CHANNEL CALIBRATION verifies the accuracy of
               ,~v-       each component within the in?trument channel. This includes
   '~""'\I""    - p,,     .calibration of the undervdltage relays and demonstrates that
    * \ll(~eR.*

A'l>~*P 1 ~) ~ the equipment falls within the specified operating characteri.sti cs defined by the manufacturer.

'\t wt.

             ~e\~ft.
     ~Pl-I ~:~.
                  . f-O~The Surveillance verifies that the channel* responds to a measured parameter within the necessary range and accuracy.

CHANNEL CALIBRATION leaves the channel adjusted to account for instrument drift between successive calibrations to ensure that the channel remains operational between successive tests. CHANNEL CALIBRATIONS must be performed consistent with the setpoint analysis. REFERENCES 1. 10 CFR 50, Appendix A GDCs 17 and 21

2. FSAR, Section 8.6

3. CPCo Analysis EA-ELEC-VOLT-033

4. CPCo Analysis EA-ELEC~VOLT-034

5. CPCo Analysis EA-ELEC-VOLT-17

6. FSAR, Chapter 14

7. CPCo Analysis EA 1 ELEC-VOLT-13

8. CPCo Analysis A-NL-92-111 1-x B 3.3.5-6 01/20/98 Palisades Nuclear* Plant

DG - UV Start

BASES ACTIONS A.1 (continued)

B 3.3.5 Required Action A.1 ensures that Required Actions for the affected DG inoperabilities are initiated. Depending upon plant MODE, the actions specified in LCD 3.8.1 or LCO 3.8.2, as applicable, are required immediately. SURVEILLANCE SR 3.3.5.1 REQUIREMENTS A CHANNEL FUNCTIONAL TEST is perfonned on each UV Start logic channel every 18 months to ensure that the logic channel will

        ~'.......
                  ~

Q perform its intended function when needed. The Under Voltage I sensing relays are tested by SR 3.3.5.2. A successful test of

          ~          the required contact(s) of a channel relay may be performed by
         ?~       ~  the verification of the change of state of a single contact of t        ~  the relay. This clarifies what is an acceptable CHANNEL.
                  ~  FUNCTIONAL TEST of a relay. This is acceptable because all of
         ~

the other required contacts of the relay are verified by other t '

       ~~            Technical Specifications and non-Technical Specifications tests.

The Frequency of 18 months is based on the plant conditions necessary to perfonn the test. The fellewin~ SR applies te each DG UV Start Functien. SR 3.3.5.2+ SR 3.3.5.l is the performance of a CHANNEL CALIBRATION every 18 months. The CHANNEL CALIBRATION verifies the accuracy of each component within the instrument channel. This includes calibration of the undervoltage relays and demonstrates that the equipment falls within the specified operating characteristics defined by the manufacturer. The Surveillance verifies that the channel responds to a measured parameter within the necessary range and accuracy. CHANNEL CALIBRATION leaves the channel adjusted to account for instrument drift between successive calibrations to ensure that the channel remains operational between successive tests. CHANNEL CALIBRATIONS must be performed consistent with the setpoint analysis. The Frequency of 18 months is based on the plant conditions necessary to perform the test. Palisades Nuclear Plant B 3.3.5-6 03/25/99

Refueling CHR Instrumentation B 3.3.6

BASES SURVEILLANCE SR 3.3.6.l (continued)

REQUIREMENTS Significant deviations between the two instrument channels could be an indication of excessive instrument drift in one of the channels or of something even more serious. CHANNEL CHECK will detect gross channel failure; thus*, it is key to verifying the instrumentation continues to operate properly between each CHANNEL CALIBRATION. Agreement criteria are detennined by the plant staff, based on a combination of the channel instrument uncertainties, including indication and readability. If a channel is outside the criteria, it may be an indication that the transmitter *or the signal processing equipment has drifted outside its limits. The Frequency, about once every shift, is based on operating experience that demonstrates the.rarity of channel failure. Since the probability of two random failures in redundant channels in any 12 hour period is low, the CHANNEL CHECK minimizes the chance of loss of protective function due to failure of redundant channels. The CHANNEL CHECK supplements less fonnal, but more frequent, checks of channel OPERABILITY during nonnal operational use of the displays associated with the LCO required channels. SR . 3.3.6.2 A CHANNEL FUNCTIONAL TEST is perfonned on each Refuell ng CHR channel to ensure the entire channel will perform its intended function~Any setpoint adjustment must be consistent with 10 CFR 100 requirements. *

The Frequency of 31 days is based on plant operating experience with regard to channel OPERABILITY and drift, which demonstrates that failure of more than one channel of a given Function in any 31 day interval is a rare event .
Palisades Nuclear Plant B 3.3.6-4 01/20/98

PAM Instrumentation B 3.3.7 BASES SURVEILLANCE SR 3.3.7 .1 (continued) REQUIREMENTS Agreement criteria are determined by the plant staff based on a combination of the channel instrument uncertainties, including indication and readability. If a channel is outside the criteria, it may be an indication that the sensor or the signal processing equipment has drifted outside its limit. If the channels are within the criteria, it is an indication that the channels are OPERABLE. If the channels are normally off scale during times when surveillance is required, the CHANNEL CHECK will only verify that they are off scale in the. same direction. Off scale low current loop channels are verified to be reading at the bottom of the range and not failed downscale. As indicated in the SR, a CHANNEL CHECK is only required for those channels which are normally energized. The Frequency of 31 days is based upon plant operating experience with regard to channel OPERABILITY and drift, which demonstrates that failure of more than .one channel of a given Function in any 31 day interval is a rare event. The CHANNEL CHECK supplements less formal, but more frequent, checks of channel during normal operational use of the displays associated with this LCO's required channels. SR 3.3.7.2 A CHANNEL CALIBRATION is performed every 18 months or approximately every refueling. CHANNEL CALIBRATION is a e check of the instrument channel including the The Surveillance verifies the channel responds to the measured parameter within the necessary range and accuracy. For the core exit thermocouples, a CHANNEL CALIBRATION is performed by substituting a known voltage for the thermocouple. The Frequency is based upon operating experience and consistency with the typical industry refueling cycle and is* justified by an 18 month calibration interval for the determination of the magnitude of equipment drift .

B 3.3.7-14 01/20/98 Palisades Nuclear Plant
Definitions RAI 3.3 &3.9 BASES INSERT K Add to Bases for SR 3.3.7.2, SR 3.3.9.2, and 3.9.2.2 This SR is modified by a Note which states that it is not necessary to calibrate neutron detectors because of the difficulty of simulating a meaningful signal. Wide range and source range nuclear instrument channels are not calibrated to indicate the actual power level or the flux in the detector location. The circuitry is adjusted so that wide range and source range readings may be used to determine the approximate reactor flux level for comparative purposes .

Alternate Shutdown System B 3.3.8 BASES

                               '3 SURVEILLANCE      SR 3.3.8.A' (continued)

REQUIREMENTS Operating experience demonstrates that Alternate Shutdown System instrumentation channels seldom fail to pass the Surveillance when performed at a Frequency of once every 18 months. Therefore, the Frequency was concluded to be acceptable from a reliability standpo int. ~ft

                             .            TUC)         ,J.~ 1          'T~

This SR is modified by .a- Note'iT#t+etl statei that -S is not required for Functions 16, 17, and 1~ REFERENCES 1. FSAR, Section 7.4, "Other Safety Related Protection, Control, and Display Systems"

2. 10 CFR 50, Appendix A, GDC 19 and Appendix R.

Palisades Nuclear Plant J- l'.'.L L.,

B 3.3.8-7 01/20/98

Definition RAI 3.3 BASES INSERT L 3 Add to Bases for SR 3.3.8~

Note 2 states that it is not necessary to calibrate neutron detectors because of the difficulty of simulating a meaningful signal. Wide range and source range nuclear instrument channels are not calibrated to indicate the actual power level or the flux in the detector location. The circuitry is adjusted so that wide range and source range readings may be used to detennine the approximate reactor flux level for comparative purposes .

1-cud

Neutron Flux Monitoring Channels

BASES B 3.3.9 SURVEILLANCE SR 3.3.9.1 REQUIREMENTS SR 3.3.9.1 is the perfonnance of a CHANNEL CHECK on each required channel every 12 hours. A CHANNEL CHECK is nonnally a comparison of the parameter indicated on one channel to a similar parameter on other channels. It is based upori the assumption that instrument channels monitoring the same parameter should read approximately the same value. Significant deviations between instrument channels could be an indication of excessive instrument drift in one of the channels or of something even more serious. CHANNEL CHECK will detect gross channel failure; thus, it is key to verifying that the instrumentation continues to operate properly between each CHANNEL CALIBRATION.

Agreement criteria are detennined by the plant staff and should be based on a combination of the channel instrument uncertainties including control isolation, indication, and

readability. 1f a channel is outside the criteria, it may be an indication that the transmitter or the signal processing equipment has drifted outside its limits. If the channels are within the criteria, it is an indication that the channels are OPERABLE.

The Frequency, about once every shift, is based on operating experience that demonstrates the rarity of channel failure. Since the probability of two random failures in redundant channels in any 12 hour period is extremely low, CHANNEL CHECK minimizes the chance of loss of protective function due to failure of redundant channels. CHANNEL CHECK supplements less fonnal, but more frequent, checks of channel OPERABILITY during nonnal 'operational use of displays associated with the LCO required channels.

                   *sR  3.3.9.2 SR 3.3.9.2 is the perfonnance of a CHANNEL CALIBRATION. A CHANNEL CALIBRATION is perfonned every 18 months~ The Surveillance* is a complete check and readjustment of the neutron flux channel from the preamplifier input through to the remote indicators.
              ~                            '

This Frequency is the same as that employed for the same channels in the other applicable MODES. Palisades Nuclear Plant B 3.3.9-4 01/20/98

- Definitions RAI 3.3 & 3.9 BASES INSERT K Add to Bases for SR 3.3.7.2, SR 3.3.9.2, and 3.9.2.2 This SR is modified by a Note which states that it is not necessary to calibrate neutron detectors because of the difficulty of simulating a meaningful signal. Wide range and source range nuclear instrument channels are not calibrated to indicate the actual power level or the flux in the detector location. The circuitry is adjusted so that wide range and source range readings may be used to determine the approximate reactor flux level for comparative purposes
ESRV Instrumentation B 3.3.10
BASES SURVEILLANCE REQUIREMENTS SR 3.3.10.1 (continued)

Agreement criteria are determined by the plant staff, based on a combination of the channel instrument uncertainties, including indication and readability. If a channel is outside the criteria, it may be an indication that the transmitter or the signal processing equipment has drifted outside its limits. The Frequency, about once every shift, is based on operating experience that demonstrates the rarity of channel failure. The CHANNEL CHECK supplements less formal, but more frequent, checks of channel OPERABILITY during normal operational use of the displays associated with the LCO required channels. SR 3.3.10.2 A CHANNEL FUNCTIONAL TEST is performed on each ESRV *

                  , Instrumentation channel to ensure the entire channel will perform its intend d nction Any setpoint adjustment must e consistent with the assumptions of the current plant specific setpoint analysis.

The Frequency of 31 days is based on plant operating experience with regard to channel OPERABILITY and drift., which demonstrates that failure of more than one channel of a given Function in any 31 day interval *is a rare event. SR 3.3.10.3 CHANNEL CALIBRATION is a complete check of the instrument channel including the sensor. The Surveillance verifies that the channel responds to a measured parameter within the necessary range and accuracy. CHANNEL CALIBRATION leaves the channel adjusted to account for instrument drift between successive calibrations to ensure that the channel remains operational between successive tests. CHANNEL CALIBRATIONS must be performed consistent with the plant specific setpoint analysis. The Frequency is based upon the assumption of an 18 month calibration interval for the determination of the magnitude of equipment drift in the setpoint analysis.

Palisades Nuclear Plant B 3 .3 .10-4 01/20/98

                                                                                                                                    @1 RPS InstrumentationE;:opoera~g S4na~"[D

3. 3. ~ .

SURVEILLANCE REQUIREMENTS continued SURVEILLANCE FREQUENCY

[T 4-.11 ~1] SR 3.3.1.~ Perform

                                ~..,..,..,=-t=-=-=-::~==-=-=r<-...-~,--,,-0,.......,..-=:?.T-:'.~

A.ND ¥E:f.'.'.1PY ~~ ~~""l:l..' N\A~\~ MONITO~ C..0N~n.~i"Ci. rT4-.11-l LFoctn~+e(c.\ J SR 3.3.1.~ -~--------,-------NOTE---------,---------~ Neutron dete~s are e~cluded fros:HANN~L (..--....

                                                                                                                'J}J CALIBRATION.

_______::,. _______________ ~

                                                                              ~---------      *          ----

CAL.1l5RA...T10M c 11ec."' oF THC:- 92 days Once within 7 days prior to ~ each reactor startup

SR~ au rform a CHANNEL FUNCTIONJ\L TEST on eac~ Once wi~hin matic bypa~oval fu~. 92 days P{ior to each 17~tor art up .

SR 3 .3 .1.8 ~la@ months Veri~S RES~ TIME ~thin 11~ t:>'E:-F R'AI

CEOG STS 3.3-5 Rev 1, 04/07/95
SURVEILLANCE REQUIREMENTS PAM Instrumentation ~

3I&-G) rh;;;-5R;-;;;;;-i;-;;~h-PN1-1~;i;~~~i~;~-F~~~i1;~-1~-r;b;;-3:3:~------- SURVEILLANCE FREQUENCY 31 days Eb

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

Neutron detectors are excluded from CHANNEL CALIBRATION. Perform CHANNEL CALIBRATION

i1a~ 1110nths

(

CEOG STS 3.3-42 Rev 1, 04/07/95

SURVEILLANCE RE UIREHENTS SURVEILLANCE FREQUENCY SR 3.3. Verify each required control circuit and ~l~ months transfer switch is capable of performing the intended function. SR 3.3.~-----------------NOTE-------------------- (gJNeutron detectors are excluded from the 1 CHANNEL CALIBRATION. I \ Perform CHANNEL CALIBRATION for each ~lei months

required instrumentation channel .

ED

CEOG STS 3.3-45 Rev 1, 04/07/95
SURVEILLANCE REQUIREMENTS* Ccont1nuedl SURVEILLANCE FREQUENCY Cr 4. 11. !..] SR 3.3.~----------------NOTE--------------------

Neutron detectors are excluded from CHANNEL CALIBRATION. Perform CHANNEL CALIBRATION

itf@ months CJ) 3.3.IO Ne...v Po.~.

[ tJ 5 £fl T i - Spee i .(. \ l-A.. +-c' ~ 3

3
I 0 ~

11 i\.SR. Ve.~*; la..f°~CVL 'I"s.\-r~~e~~ tc....+;&v\. ( CEOG STS 3.3-48 Rev 1, 04/07/95

                                                        /-0-j__

Change ATTACHMENT 6 JUSTIFICATION FOR DEVIATIONS SPECIFICATION 3.3.1, RPS INSTRUMENTATION Discussion

17. Figures 3.3.1-1, 3.3.1-2, and 3.3.1-3 do not apply to Palisades. Table 3.3.1-2 has been included which provides the required relationship and allowable values at the Thermal Margin/Low Pressure {TM/LP) Function.

18. Table 3.3.1-1 Footnote (c) does not apply, subsequent Footnotes have been renumbered, where applicable.

19. The specific wording which discusses other plants is being deleted the bases are specific to Palisades and will not contain, where possible, only Palisades specific information.

20. The bracketed Reviewer's Note has been deleted since it is not meant to be maintained in the plant specific ITS.

21. ISTS Figure B 3.3.1-1 has been deleted since similar diagrams are already included in FSAR Figure 7-1 and 7-2 .

22.

23. Note 2 of ISTS SR 3.3.1.2 has been deleted since the PHYSICS TESTS are performed below 2 % RTP and therefore, the allowance in Note 1 suffices. Neutron de tors are specific eluded fro ION, as in ITS Section , "Definition ' STF-81 is also incorporated on t sa

24. Requirement to verify constants associated with the thermal margin monitors is added, consistent with details relocated from CTS Table 4.17.1, Functional Unit 15 (see DOC LA.l).

25. TSTF-178 is incorporated to omit "trip or bypass removal" from the ACTIONS Note.

The RPS Functions listed in Table 3.3.1-1 include trip and bypass removal features where appropriate. Referring to the trip or bypass removal features as separate Functions is incorrect and confusing. Removing the words "trip or bypass removal" satisfies the intent of the Note and eliminates the error. This change is also consistent with the CEOG Digital LCO. I

Palisades Nuclear Plant Page 3 of 4 01/20/98

- ATTACHMENT 6 JUSTIFICATION FOR DEVIATIONS .

SPECIFICATION 3.3.11, PAM INSTRUMENTATION 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.

4.

The following requirements have been renumbered, where applicable, to reflect this

deletion.

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.

7. Containment Isolation Valve (CIV) required channels is modified to reflect plant design, which includes only one position indication channel per valve, and not all penetrations are equipped with CIVs of a type that is equipped with position indication, consistent with current design and licensing basis. *

8. Details related to method of calibrating core exit thermocouples are added, consistent with details relocated from CTS Table 4.,17.4, Footnote (a) (see DOC LA.1) .

Palisades Nuclear Plant Page 1of2 01/20/98
Change SPECIFICATION 3.3.12, REMOTE SHUTDOWN SYSTEM Discussion ATTACHMENT 6 JUSTIFICATION FOR DEVIATIONS 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.

4.

The following requirements have been renumbered, where applicable, to reflect this . deletion. 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. 7.

8. The Applicable Safety Analyses section of the Bases has been modified to reference 10 CFR 50.36(c)(2), consistent with the Bases for other Specifications.

9. Alternate Shutdown System transfer switches and their location are added, consistent with details relocated from CTS Tables 3.17.5 and 4.17.5 (see DOC LA.1) .

Palisades Nuclear Plant Page 1of2 01/20/98

Change SPECIFICATION 3.3.13, POWER MONITORING CHANNELS Discussion ATTACHMENT 6 JUSTIFICATION FOR DEVIATIONS

8. The Applicability of the Specification has been modified to be consistent with the current Technical Specifications to ensure the neutron flux indicators are available whenever the plant is shutdown (MODES 3, 4, and 5, with no more than one control rod capable of being withdrawn and the Primary Coolant System (PCS) boron concentration less than the REFUELING BORON CONCENTRATION, and MODES 3, 4, and 5, with the PCS boron concentration at the REFUELING BORON CONCENTRATION). In these MODES of operation, SHUTDOWN MARGIN must be consistent with ITS 3 .1.1. A boron dilution event could occur in these conditions and, therefore, neutron flux indication is necessary so that the plant operator can take the apprqpriate actions. The Palisades design does not include reactor trip circuit breakers, however it does have individual control rod circuit breakers. Under these circumstances, the capability to withdraw one control rod provides flexibility in conducting control rod Surveillances and tests.

9. ISTS SR 3.3.13.2 for performance of a CHANNEL FUNCTIONAL TEST is not being proposed since the CHANNEL CHECK and CHANNEL CALIBRATION Frequencies are considered adequate to ensure the OPERABILITY of the equipment. The proposed Frequencies are consistent with the Surveillance Frequencies of the Post Accident Monitoring Instrumentation Frequencies in ITS 3.3. 7, which provides indication-only Functions. Therefore, this change is considered to be consistent with NUREG-1432 for similar type instrumentation functions. This change is also supported by the current licensing basis since no periodic CHANNEL FUNCTIONAL TEST Surveillance is provided in the CTS. A CHANNEL FUNCTIONAL TEST is required just prior to each startup but its purpose is to verify monitoring capability for the startup, not for monitoring during the shutdown conditions. Therefore, the CTS CHANNEL FUNCTIONAL TEST is appropriately addressed in ITS 3.3.1.

10. e Note cludin~ n~n detectors fro HANNEL~IBRA~~ted.

is lusion ha~n incorporat

to the defin~of C~L _...

A IBRATION

11. Text referring to other plant designs has been deleted to make the ITS specific to Palisades .

Palisades Nuclear Plant Page 2 of 3 1-a.o 01/20/98

LTOP System B 3.4.12

BASES SURVEILLANCE REQUIREMENTS SR 3.4.12.3 (continued)

The 72 'hour Frequency considers operating experience with accidental movement of valves having remote control and position indication capabilities available where easily monitored. These considerations include the administrative controls over main control room access and equipment co.ntrol.

SR 3.4.12.4 Performance of a CHANNEL FUNCTIONAL TEST is required every 31 days tg v9rify eRd, ilS RQ~Qsss.ry, s.dablst U1e PQRV e~eR set~ei Rts, Hie Glol/UINEL i;YNGTlQNAL T~~T wi 11 vm*i fy eA a AIQRthly basis that the PQR'l lift set~eiRtS aFe HitRiR tR@

LGQ lilRH:,.APORV actuation could depressurize the PCS and is not required. The 31 day Frequency considers experience with equipment reliability. A Note has been added indicating this SR is required to be performed 12 hours after decreasing any PCS cold leg temperature to< 430°F. This Note allows a discrete period of time to perform the required test without delaying entry into the MODE of Applicability for LTOP. This option may be exercised in cases where an unplanned shutdown below 430°F is necessary as a result of a Required Acfion specifying a plant shutdown, or other plant evolutions requiring an expedited cooldown of the plant. The test must be performed within 12 hours after entering the LTOP MODES. SR 3.4.12~5 Performance of a CHANNEL CALIBRATION on each required PORV J actuation channel is required every 18 months to adjust the entire channel so that it responds and the valve opens within the required LTOP range and with accuracy to known input. The 18 month Frequency considers operating experience with equipment reliability and is consistent with the typical refueling outage schedule .

Palisades Nuclear Plant B 3.4.12-'12 01/20/98

                                          /-a_p

PCS Leakage Detection Instrumentation B 3.4.15

BASES ACTIONS (continued) k..t.1 If all required monitors are inoperable, no automatic means of monitoring leakage are available and immediate plant shutdown in accordance with LCO 3.0.3 is required.

SURVEILLANCE SR 3.4.15.1. SR 3.4.15.2. and SR 3.4.15.3 REQUIREMENTS These SRs require the perfonnance of a CHANNEL CHECK for each required containment sump level indicator, containment atmosphere gaseous activity monitor, and containment atmosphere humidity monitor. The check gives reasonable confidence the channel is operating properly. The Frequency of 12 hours is based on instrument reliability and is reasonable for detecting off nonnal cond it i ans. SR 3.4.15.4

~ _

SR 3.4.15.4 requires the perfonnance of a CHANNEL FUNCTIONAL TEST of the required containment air cooler condensate level switch. Since this instrumentation does not include control room indication of flow rate, a CHANNEL CHECK is not possible. The test ensures that the 1tJ.)tCT ,..,~evel switch can perfonn its function in the desired manner. Hie test verifies tl:le al arm set~gi Rt aREI rehti VQ

                     -~ee~fiaex gf tl:le iRltrwleR~ &i[iB9_,. The Frequency of 18
  'llt~t~\T\o...\     man s 1s a typica re ue ing cycle (perfonnance of the "t.        test is only practical during a plant outage) and
      ~~              considers instrument reliability. Operating experience has shown this Frequency is acceptable for detecting degradation.

SR 3.4.15.5, SR 3.4.15.6. and SR 3.4.15.7 These SRs require the perfonnance of a CHANNEL CALIBRATION for each required containment sump level, containment atmosphere gaseous activity, and containment atmosphere humidity channel. The calibration verifies the accuracy of the instrument string, including the instruments located inside containment. The Frequency of 18 months is a typical refueling'cycle and considers channel reliability. Operating experience has shown this

Frequericy is acceptable .

Palisades Nuclear Plant B 3.4.15-5 01/20/98

                                         /-~A

Nuclear Instrumentation 3.9.2

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.9.2.1 Perfonn CHANNEL C~ECK. 12 hours SR 3.9.2.2~erfonn CHANNEL CALIBRATION. 18 months

     ~   -   ~E1J~kl
                                   ~OTE:

1E--re"C.rof!S AlZE F-=-1LW~ l=-e'OM it\E:' C~M~EL CAL\6~T\Ol-J.* II

Palisades Nuclear Plant 3.9.2-2 Amendment No. 01/20/98 l-0-s

Nuclear Instrumentation B 3.9.2 BASES

 'SURVEILLANCE      SR 3.9.2.2 REQUIREMENTS (continued)

REFERENCES 1. FSAR, Section 7.6

2. FSAR, Section 14.3

Palisades Nuclear Plant B 3.9.2-4 01/20/98 1-o_f
Defi nit i ans RAI 3.3 &3.9 BASES INSERT K Add to Bases for SR 3.3.7.2, SR 3.3.9.2, and 3.9.2.2 This SR is modified by a Note which states that it is not necessary to calibrate neutron detectors because of the difficulty of simulating a meaningful signal. Wide range and source range nuclear instrument channels are not calibrated to indicate the actual power level or the flux in the detector location. The circuitry is adjusted so that wide range and source range readings may be used to determine the approximate reactor flux level for comparative purposes .
/-au

Ruclear Instrumentation 3.9.2

- SURVEILLANCE REQUIREMENTS SURVEILLANCE

                                                                  --     FREQUENCY SR 3.9.2.l     Perform CHANNEL CHECK.                        12 hours SR 3.9.2.2     -----*******---~---NOTE*-------------------

Neutron detectors are excluded from CHANNEL CALIBRATION. Perform CHANNEL CALIBRATION. (j) CEOG STS 3.9-3 Rev 1, 04/07/95

!-av

- ATTACHMENT 6 JUSTIFICATION FOR DEVIATIONS SPECIFICATION 3.9.2, NUCLEAR INSTRlTh-fENTATION 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 infonnation 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 initial perfonnance of SR 3.9.1.1 within 4 hours of entry into Condition B has been deleted. The accelerated perfonnance of this SR is not warranted based on routine perfonnances of this SR (every 72 hours), and knowledge of stable conditions prior to the loss of the source range channel. Secondarily, PCS dilution events are recognizable through other means such as uncontrolled increases in pool water level. This change is consistent with NUREG-1432 as modified by TSTF-96.

7. fro

Palisades Nuclear Plant Page 1of2 01720798

CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO JANUARY 6, 1999 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.3, INSTRUMENTATION NRC REQUEST: (Referred to as Bases RAI) Bases Instrumentation LCOs Bases review RAls will be provided following discussion and agreement on ITS LCOs. Consumers Energy Response: The ITS Bases have been revised to both incorporate changes necessitated by the LCO changes included in this response, and to incorporate corrections and clarifying changes found to be desirable during preparation of the RAI related changes. The formerly submitted ITS bases has been marked up to show these changes. Changes related to an. RAI question have been marked accordingly. The marked up Bases is provided in Enclosure 2. Affected Submittal Pages: Section 3.3, Attachment 2, ITS Bases, All pages (See Enclosure 2)

3

CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO JANUARY 6, 1999 REQUEST FOR ADDITIONAL INFORMATION

NRC REQUEST:

3.3.1-1 SECTION 3.3, INSTRUMENTATION ITS Applicability JFD 8 JFD 8 references changes which are approved by TSTF-85, Rev. 1. This TSTF reformats the applicability by moving Mode and applicable conditions to Table 3.3.1-1. These TSTF changes were not included as stated. Comment: Provide a revised JFD 8. Also, revise JFD 8 to include discussion justifying deviation from the STS Applicable Conditions "any RTCB's closed and any control element assemblies capable of being withdrawn." Consumers Energy Response: The following changes to our January 26, 1998 conversion submittal have been made: JFD 8 has been revised. ITS 3.3.1 and its bases have been revised to reflect the changes contained in TSTF-85. The Applicability of the High Startup Rate Trip Function was revised from "when. there is fuel in the reactor, more than one CONTROL ROD is capable of being withdrawn, and the PCS is less than REFUELING BORON CONCENTRATION" to "MODES 1and2". The Completion Time for the Required Actions associated with that Function were revised from "Prior to entering MODE 3 from Mode 4" to "Prior to entering MODE 2 from MODE 3", to agree with the change in Applicability. The 3.3.1 Action table was re-ordered to place all Conditions addressing "one channel inoperable" first, then "one or more ZPM bypass removal channels inoperable" next, and those addressing "two channels inoperable" last.

Former ITS Conditions G and H, and their Actions G.1 and H.1 have been deleted; the general "two channels inoperable" Condition E, and its Action E.1 provide equivalent requirements. Former ITS Actions G.2 and H.2 are unnecessary, since equivalent requirements are provided by Actions B.1 and C.1 respectively.

The Note stating that former Condition B did not apply to the High Startup Rate and Loss of Load trip Functions has been applied to Action E.2. These changes in the Actions Table result in a more logical sequence of Conditions, and a less complicated table. The rearrangement of the Conditions and Actions does not alter the requirements. *

Conforming changes have been made to the ITS Bases .

DOCs M.1 and M.6 have been revised accordingly. 4 (continued)

CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO JANUARY 6, 1999 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.3, INSTRUMENTATION NRC REQUEST: 3.3.1-1 Consumers Energy Response: (continued) The Applicabilities and footnotes listed in Table 3.3.-1-1 are plant specific. They are based on the Palisades Safety Analyses assumptions and the current licensing bases. Affected Submittal Pages: Att 1, ITS 3.3.1, page 3.3.1-1 Att 1, ITS 3.3.1, page 3.3.1-2 Att 1, ITS 3.3.1, page 3.3.1-3 Att 1, ITS 3.3.1, page 3.3.1-6 Att 1, ITS 3.3.1, page 3.3.1-7 Att 2, ITS B 3.3.1, page B 3.3.1-19 (former submittal) Att 2, ITS B 3.3.1, page B 3.3.1-25 (replacement) Att 3, CTS, page 3-64 Att 3, CTS, page 3-65 Att 3, DOC 3.3.1, page 5 of 9 (former submittal) Att 3, DOC 3.3.1, page 7 of 9 (former submittal) Att 3, DOC 3. 3. 1, page 5 of 1O (replacement) Att 3, DOC 3. 3.1, page 6 of 10 (replacement) Att 3, DOC 3.3.1, page 7 of 10 (replacement) Att 5, NUREG, page 3.3-1 (and Insert) Att 5, NUREG, page 3.3-2 (and Insert) Att 5, NUREG, page 3.3-3 Att 5, NUREG, page 3.3-6 Insert Att 5, NUREG, page 3.3-7 Insert Att 6, JFD 3.3.1, page 2 of 4 (former submittal) Att 6, JFD 3.3.1, page 3 of 4 (replacement)

5

RPS Instrumentatiori

3. 3 .1

3.3 3.3.1 INSTRUMENTATION Reactor Protective .system (RPS) Instrumentation C~/l..~\..\e.LS, ~t-lD AS~oc..1A.-ri=t> 1:.E:Ro Pow~ Mooe- cPM LCO 3.3.1 Four RPS trip units,~ associated instrument~ 8ypass removal channels for each Function in Table 3.3.1-1 shall be OPERABLE.

APP LI CAB I LI TY: ACTIONS

 ---------------~---------------------NOTE-------------------------------------

Separate Condition entry is allowed for each function. CONDITION REQUIRED ACT ION COMP LET ION TIME

A. --------NOTE---------- A.1 Not applicable to High Startup Rate>~ Loss of Load~unctions.

Place affected trip uni t in t ri p.

                                         ~OR ~Pt-A. SY'PA':f':, R&MD'/6.L.)

7 days One or more Functions '!. with one RPS trip unit or associated instrument channel inoperable. . .

Palisades Nuclear Plant 3.3.1-1 S-c.__

Amendment No. 01/20/98

                             - - - - - -~oTE:-- - - - - -

- ACTIONS CONDITION "10\ ~?'?L\C.0.61..E- 1"o ~='~

STA.RTL>P RA.IE:- oR Lo~b Ol=- LOO..D HJ~C..llO~S. REQUIRED ACTION RPS Instrumentation COMPLETION TIME

3. 3 .1 KAI.

33.1-1 One or more Functions 3.~.1-fo with two RPS trip units or associated instrument channels Restore one trip unit 7 days inoperable. and associated instrument channel to OPERABLE status. Two power range Restrict THERMAL 2 hours channels inoperable. POWER to ~ 70% RTP .

D.

(J:tit\ME.D\~T1::\.."' ) IY tmef I

~A!

QR _ _ _ _. _ 3;s.1-eo D.2 ~,......, ~~1flb!i:L.0 One High Startup Rat-e Restore trip unit and. Prior to ~ Trip unit or associated instrument entering~~ associated instrument channel to OPERABLE from MODE~ g~ channel inoperable. status. ~.'3.1-1 One Loss of Load trip Restore trip unit and unit or associated associated instrument instrument channel channel to OPERABLE inoperable. status. 5-b Palisades Nuclear Plant 3.3.1-2 Amendment No. 01/20/98

ACTIONS CONDITION REQUIRED ACT ION RPS Instrumentation 3.3.1 COMPLETION TIME

                                                                       . RA'I.

Required Action and Be in MODE 3. 6 hours 3,3,1-1 associated Completion Time not met. E-D Verif no more than 6 hours one control rod is Control room ambient capable of being air temperature withdrawn.

        > 90°F.

Verify PCS boron 6 hours concentration is at REFUELING BORON CONCENTRATION . 5-c_ Palisades Nuclear Plant 3.3.1-3 Amendment No. 01/20/98

RPS Instrumentation

3. 3 .1 Table 3.3.1-1 (page 1 of 2)

Reactor Protective System Instrumentation A?PL\C.A8L£. SURVEILLANCE FUNCTION 'MODES REQUIREMENTS ALLOWABLE VALUE

1. Variable High Power Trip SR 3. J .1.1 ~ 15% RTP above current SR J.J.1.2 THERMAL POWER wfth a SR J.J.1.3 *fni*1111 of ~ 30% RTP SR J.J.1.4 and a 11axi*u* of ~

SR 3. J .1. 5 106.5% RTP SR 3.J.1.6 SR J. 3 .1. 8

2.

High Startup RatJlJT~1'P(b) SR J.J.1.1 HA SR J. J .1. 7 SR J.J.1.8 g~

              '                                          "") .,(~ (~ ~(Q.J                                                      !'.3.\-1

3. Low Pri111ary Coolant Syste111 Flowli \1 '-1;;1 1 Y ,? SR J.J.1.1 , 95%

                                     '            ~1p(c.)                       SR J. J. l."S SR J.J.1.8
        '                          '                         (a..) (2.) ~(a.)

4.. Low Stea111 Generator A Level 11<1? I, '2 1!. 1 1.-\ SR J. J .1.1 . , 25.9% narrow range 1 -::i SR J.J.1.5 SR J. 3 .1. 8

5. Low Steam Generator B Level~I? 1 1 '2. 1 ~(~~~ '5'L~) SR 3. J .1.1 , 25.9% narrow range SR 3. 3 .1. 5 SR 3. 3 .1. 8 SR 3. 3 .1.1 , 500 psfa SR 3. 3 .1. 5 SR 3.3.1.8 SR 3.3.1.1 ' , 500 ps fa SR J. J .1. 5 SR 3.3.1.8

8. High Pressurizer Pressure""\~'? SR J.J.1.1 ~ 2255 psia SR 3.3.1.5 SR 3.3.1.8 (a..'> w1~ "-\ol?'i~~ oue. FUw..-1..E=-~G.~ co~Teo'- E?Ol:' CA.PA'SL.e- or=- se-1...u.

WITMt>~tJ l\tlD PGS eoear-.t COklC.-t::::f..IT1CA.lfON LESS T-HAtJ REFUELING:.

         "Boi:::'DIJ CDNcEN 71?.AI~.

5~d Palisades Nuclear Plant 3.3.1-6 Amendment No. 01/20/98

RPS Instrumentation

Table 3.3.1-1 (page 2 of 2)

Reactor Protective System Instrumentation 3.3.1 A??l..\CA.~ SURVEILLANCE FUNCTION l°'t\Cl:E$ REQUIREMENTS ALLOWABLE VALUE

                                  .              (o.)      (o.)    - (o.)

9. Ther1141 Hargfn/Low Pr1ssur1tsl \ 1Z,'; 1 L.l ,? SR 3.3.1.1

Table 3.3.1-2 T~1 f" (c) SR 3. 3 .1. 2 SR 3.3.1'.3 SR 3.3.1.4 SR J.3.1.5 SR J.3.1.6 SR 3. J .1. 8

10. L~ss of Loadl?IT~1? SR 3.J.1.7 NA SR 3. 3 .1. 8

                                               ' '7 (11.)       (a.) '5 (o.)

11. Contafn11entHfghPressurel~1'F" 1,2,J ,Y 1 SR 3.3.1.S ~ 3.70 psfg i2., lE!i?'o Po"'-'EiZ. }-1\ot:E. 8vf'Ac:;s z '5("") '-l (GL) 5 (~) SR 3.3.1.8 11 Au101-1t1.."T 1c.. i(EMol/AL. ' ' '

Bypass s ha 11

. ~

n THERMAL POWER fS'l2J17% RTP. le;1ai'l

                                                                                                                                       ~I
                                                                                                                                   . 3,~.1-1
                                                                                                                                     ~.~.I-~

5-e__

Amendment No. 01/20/98 Palisades Nuclear Plant 3.3.1-7

RPS Instrumentation B 3.3.1 BASES APP LI CAB IL ITY This LCO is applicable in MODES 1 and 2, and in MODES 3, a 5 when more than one control rod is capable of bei

   ~~I    J."S.1-/  wit~   awn and PCS boron concentration is less than R BORON     CENTRATION. As indicated in Note (c) of Table 3.3. -1, the Loss of Load trip is not re red to be OPERABLE wit HERMAL POWER < 17% RTP. Belo 17% RTP, the ADVs are capabl of relieving the pressur due to a Loss of Load event without hallenging other o rpressure protection.

If PCS boron conce ation is at REFU NG BORON CONCENTRATION, if no more than one con al rod is capable of being wit awn, the RPS Functirin is alre y fulfilled (the safet analyses and the SHUTDOWN MARGIN d 'nition both use the ssumption that the highest worth withdra control rod

l fail to insert on a trip) and the safety an ses
ACTIONS a . mpti ans and SHUTDOWN MARGIN requirements wi 11 be me ithout the RPS trip Function.

The most common causes of channel inoperability are failure or drift of the bistable or process module sufficient to* exceed the Allowable Value. Typically, the drift is found to be small and results in a delay of actuation rather than a total loss of function. This determination is generally made during the performance of a CHANNEL FUNCTIONAL TEST when the process instrument i~ set up for adjustment to bring it to within specification. If the trip setpoint is less conservative than the Allowable Value in Table 3.3.1-1, the channel is declared inoperable immediately, and the appropriate Condition(s) must be ~ntered immediately. In the event a channel's trip setpoint is found nonconservative with respect to the Allowable Value, or the transmitter, instrument loop, signal processing electronics, or RPS bistable trip unit is found inoperable, then all affected Functions provided by that chan~el must be declared inoperable, and the plant must enter the Condition for the particular protection.Functions affected .

.Palisades Nuclear Plant B 3.3.1-19 01/20/98 5--iJ

RPS Instrumentation B 3.3.1

BASES APPLICABILITY This LCO requires all safety related trip functions to be OPERABLE in accordance with Table 3.3.1-1. While in MODES 3, 4, or 5, if PCS boron concentration is at REFUELING BORON CONCENTRATION, or if no more than one full-length control rod is capable of being withdrawn, the RPS Function is already fulfilled (the safety analyses and the SHUTDOWN MARGIN definition both use the assumption that the highest worth withdrawn full-length control rod will fail to insert on a trip) and the safety analyses assumptions and SHUTDOWN MARGIN requirements will be met without the RPS trip Function.

The High Startup Rate Trip Function is required to be OPERABLE in MODES 1 and 2, but may be bypassed when the associated wide range NI channel indicates below lE-4% power, when poor countfng statistics may lead to erroneous indication. It may also be bypassed when THERMAL POWER is above 13% RTP, where moderator temperature coefficient and fuel temperatu*re coefficient make high rate of change of power unlikely. In MODES 3, 4, 5, and 6, the High Startup Rate trip is not required to be OPERABLE. Wide range channels are required to be OPERABLE in MODES 3, 4, and 5, by LCO 3.3.9, "Neutron Flux Monitoring Channels," and in MODE 6, by LCO 3.9.2, "Nuclear Instrumentation." The Loss of Load trip is required to be OPERABLE with THERMAL POWER at or above 17% RTP. Below 17% RTP, the ADVs are capable of relieving the pressure due to a Loss of Load event without challenging other overpressure protection. The trips are designed to take the reactor subcritical, maintaining the SLs during AOOs and assisting the ESF in providing acceptable consequences during accidents .

Palisades Nuclear Plant B 3.3.1-22 05/30/99 5-~

3. 3.1

3 .~

3.3.1 Reoctoi- Protective System mPSl Instrumentot1on INSTRUMENTATION lS1SUMS I RAI I3.3.1-6 RAI Tob le 3.3.1-1 3.3.1-1 ond Footnote o 1£'17}:2 I Wi.th one RPS trip unit or associated instrument channel inoperable for ( Cond A J . Ofle or more functions: [RA A.l J a)* Place the affected tr1p unit in the tripped condition within 7 days. 1)(17.)':31 With two RPS trip units or associated instrument channels inoperable for [ Cond. E J one or more functions: < Add RA E NOTE >--@ [RA E.1 J a) Place one inoperable trip unit in the tripped condition within 1 hour. and RAI 3.3.1-1 [ Cond.F J b) If two Power Range Nuclear Instrument channels are inoperable. [RA F.1 J limit power to s 70% RATED POWER within 2 hours. and [RA E.2 J c)* Restore one RPS trip unit and associated instrument channel to OPERABLE status within 7 days. SEE 3.3.2> 1£'11 .J':6I ( Cond. G l [ RA G.1 J [ RA G.2.1 J RAI 3.3.1-1 Amendment No. 1:6-e. 186 Ver1f y PCS boron concentrot1on is ot REFUELING BORON CONCENTRATION. Add Cond. B ~nd C >-@ -< Add Cond. 0

                                                                                         'vM,p
      <   Add Required Ac t1ons B.1, ond C.l               .

M.6 1 5- L Add Requ1r ed Ac t1ons 0.1, ond 0.2 ~ Page 1 of 7

3.3.l

3.17 INSTBUJf;NfAJIQN SYSTEMS Tablelt16Qi.1-1 )

InstCU11ent1t1on Operating Requirements for Reactor protective System A. 5 Required Pen1issible RPS Operational

Euoctiao1l Uoit CblDDt=li Byp&Sses

1. [)!{nuefrrjj 7 / 7 72 7

              "----"--"--'----1--..___....__..__  _ _ _.________.'--....../H'--o. njt{""""._i=(sEE
                                                                                  ..               ,3.3.2)

[1] 2. Variable High Power(Trip ) 4 None. [2)3. H1gh Start Up Rate 4 Below 10-4%(cJ or fT3.3.1-1 1 above 13% RATED POWER. ~ootnote lbij cq14. Thenna 1 Margin/ Low Pressure (4 (b) rn@ 3~~i1-1 A.5 3.3.1-6 [8] s. H1gh*Pressur1zer 4 None. Pressure . [3] *6. LQW PCS Flow 4 (b)~

- [1017 *

[4] 8. (5) g. Loss of Load 1d1 Low *A* Steam Generator Level Low *a* Steam 4 4 4 Below 17% RATED POWER. None. None. Generator Level [6] 10. Low 1 A1 Steam 4 (b)rn--@ Generator Pressure [7Jl 1. Low *e* Steam 4 (b) [!Jill-@ Generator Pressure [11] 12. 4 None. 13. SEE 3.3.2> 14 .. [12]~ [Ccl]~ RAI 3.3.1-1 3.3.1-6

        <Add Footnotes A & B>                                   Amendment No. 118, 139, 136, 162 3-65 5-~                                          Poge 2 of 7

Se=e ~i=.111se."?:::. ti/./

K: Pt I .,. 1. -s .1 - I f 3. 3. / - f, TECHNICAL CHANGES - MORE RESTRICTIVE (M) ATTACHMENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.3.1, RPS INSTRUMENTATION I M.1 CTS 3 .17 .1 allows some Functions to not be OPERABLE, as allowed by the rmissible operational bypasses. CTS 3 .17 .1 does not contain any specific r uirements for inoperable automatic bypass removal instrument channels a ociated wiili ertain RPS Instrument Functions. CTS Table 3.17.1Functions3, ( 1gh Start Up Rate), unction 4 (Thermal Margin/Low Pressure), Function 6 (Low P Flow), Function (Loss of Load), Function 10 (Low "A" Steam Generator P. essure), and Function 1 (Low "B" Steam Generator Pressure) all have at least e operational bypass featur associated with them. ITS 3.3.1 places a require nt in the LCO that applicable bypa removal channels must be OPERABLE and e blishes Required Actions specifi for when one or two bypass removal ch els are inoperable (ITS Condition D). e Required Actions are essentially t same as when one or two. RPS trip units or assoc ted instrument channels are inop ble, except that the bypass channel is allowed to be *sabled or removed. ITS Table 3.3.1-1, Functions , 3, 6, 7, 9, and 10 e associated with bypass removal channels as indicated i the Table 3.3.1- Footnotes (a), (b) and (c) . The automatic bypass removal fea res are r red for each of these safety related trips to ensure the RPS Fune* ns ar ot operationally bypassed when the safety analysis assumes the Functio e OPERABLE. The specific RPS operating bypasses are discussed in the sed Applicable Safety Analyses Bases for ITS 3. 3 .1. The addition of this new require nt is conside more restrictive. This change will have no adverse im ct on safety beca se the action to remove the bypass channel when the auto atic bypass removal 1 inoperable ensures that the trip Function is OPE LE. This change is consi ent with NUREG-1432. New Conditions E and and new Required Actions G.2 an H.2 have been

added for the Loss of ad *and High Startup Rate Functions, d a Note has been added to the uired Actions specified in ITS 3.3.1 Con
ions A and B which states that onditions A and B are not applicable to the Los of Load and High Startup te Functions. These actions require restoration of tli inoperable c nel(s) prior to entering operational conditions during w *ch the Functions ay be needed. These Required Actions and Completion Tim are consiste with current practice though not required by CTS for these Fune ns for o and two inoperable channels. This Required Actions and Completion Ti s are acceptable since these Functions are not credited in the plant safety
Palisades Nuclear Plant Page 5of9

                                                   .5-r   .

01/20/98 I

M.6 ATTACHMENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.3.1, RPS INSTRUMENTATION S Trip Functions are This-**~..-......

 - M. 7   The CTS Table 3 .17. 6 operating requirements for the Flux-Ll T Comparator are modified by a footnote which states that Specification 4.0.4 is not applicable. This allowance is not permitted in ITS 3.3.1. Deletion of the allowance to enter a MODE or other specified condition of the applicability without having met the associated surveillance requirements is a reduction jn flexibility and is considered more restrictive.

Deletion of this allowance provides additional assurance that the required Function will be OPERABLE when it is required. This change is consistent with NUREG-1432 .

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

LA.1 CTS Table 4.17.1, Functional Unit 15, "Thermal Margiil Monitor" requires that the constants be verified every 92 days. In the ITS, this surveillance is considered to be a part of SR 3.3.1.5, the CHANNEL FUNCTIONAL TEST for the Thermal Margin/Low Power (TM/LP) Function and other Functions. The constants of the Thermal Margin Monitor are specified in CTS Table 2.3.1 (ITS Table 3.3.1-2). These constants must be verified to ensure the trip setpoint of the TM/LP Function is set correctly. The details of the CHANNEL FUNCTIONAL TEST have been relocated to the Bases. The Bases for ITS SR 3.3.1.5 states that the thermal margin monitor constants must also be checked to be within tolerances. Removing the details of the CHANNEL FUNCTIONAL TEST from the CTS and placing them in the Bases of the ITS is acceptable since these details are not pertinent to the actual requirements. 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 chan~e is consistent with NUREG-1432 .

Palisades Nuclear Plant Page 7 of 9 5-A_'

01/20/98

ATTACHMENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.3.1, RPS INSTRUMENTATION TECHNICAL CHANGES - MORE RESTRICTIVE (M) M.1 The allowances for channels to be bypassed and the requirements for automatic removal I of these bypasses is treated differently in CTS and ITS. Both treatments allow some I trip Functions' to be bypassed during conditions when the LCO requires them to be I Operable. Both treatments require the bypasses on the safety related trips (those I assumed to function by the safety analyses) to be automatically removed. I I Operability: I I In CTS, the automatic bypass removal channels (for safety related channel bypasses) I are required to be Operable by footnote (a) of Table 3 .17 .1. (When the Wide Range I Nuclear Instrument channels are Operable, if the indicated power level increases above I the setpoint, the permissive signal is removed, automatically removing the bypass, .I regardless of the position of the manual bypass switch. This bypass is called the Zero . I -.} Power Mode or "ZPM" bypass.) In ITS, the bypass removal channels are required to be Operable as part of the LCO statement and are specifically listed in Table 3. 3 .1-:-1 as required Functions. The ITS LCO and Actions are worded to differentiate between

                                                                                                         -I "trip Functions" and "bypass removal Funct~ons." These treatments are equivalent, with the ITS treatment being more explicit.

Applicability:

       . In CTS, the automatic bypass removal channels. are required to be Operable "When there is fuel in the reactor, more than one CONTROL ROD is capable of being withdrawn, and PCS boron concentration is less than REFUELING BORON CONCENTRATION, if the Zero Power Mode bypass is used"; in ITS, they are required to .be Operable whenever the associated trips are required to be Operable,*

i.e., in Modes 1and2, and in Modes 3, 4, and 5 with more than one foll-length control rod capable of being withdrawn and PCS boron concentration less than REFUELING BORON CONCENTRATION. The ITS, therefore, require the bypass removal channels to be Operable over a broader range of plant conditions than do the CTS, making this change More Restrictive .

Palisades Nuclear Plant
Page 5of11 05/30/99

                                               .!J-J

ATTACHl\fENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.3.1, RPS INSTRUMENTATION M.1 (continued) I I Bypass Allowance: I I In CTS, these bypass allowances are provided by footnote (b) of Table 3 .17 .1 which I apply to the TM/LP trip, Low PCS Flow trip, and Low SG Pressure trip functions. I That footnote allows these trips to be manually bypassed if the associated wide range I nuclear instrument ch~nnels indicate below the specified power level. In ITS, the I allowance to bypass the associated trip channels is provided in Table 3.3.1-1, footnote I (c). These treatments are equivalent, with the ITS treatment b_eing more explicit. I I Automatic Bypass Removal: I I ..._,. While not explicitly stated, CTS requires automatic removal of the ZPM bypasses by I ' requiring the associated instrument channels to be Operable: footnote (a) of I~ Table 3 .17 .1 requires both wide range channels to be Operable if the subject bypass is I \() used. Since the wide range channels, which supply the input for the High Startup Rate I rri trip, are subject to requirements for Channel Calibration (SR 4.17 .1, item 3), which 1-

explicitly requires verification of the bypass removal function. In ITS,* the requirement for automatic removal of the ZPM bypass is provided as part of Table 3. 3 .1-1 footnote (c). These. treatments are equivalent; with the ITS treatment being more ~xplicit.

Actions: 1 ct lcL 141 In CTS, the Required Actions for the bypass removal (i.e., wide range nuclear instrument) channels are those provided for inoperable RPS trip units. If a channel were bypassed (other than as allowed by the footnotes), it could not perform its specified function and would be inoperable. Therefore, if a trip channel were ZPM bypassed, when a bypass removal (wide range NI) channel became inoperable, either the bypass would have to be removed (which places the plant in a condition outside the applicability), or the bypassed trip channel(s) would have to be declared inoperable. In ITS, those actions have been preserved as ITS actions D.1 and D.2. These treatments are equivalent, with the ITS treatment being more explicit .

Palisades Nuclear Plant Page 6of11 05/30/99 6-rr,

ATTACHMENT 3

M.2 DISCUSSION.OF CHANGES SPECIFICATION 3.3.1, RPS INSTRUMENTATION CTS 3 .17 .1. 6 requires specific actions be taken when the number of OPERABLE channels is less than specified in the "Minimum OPERABLE Channels" column of Table 3 .17 .1. The actions are to place the reactor in HOT SHUTDOWN within 12 hours; and place the reactor in a condition where the affected equipment is not required, within 48 hours. In the ITS, the actions when the "number of OPERABLE ,

Channels': is less than the CTS minimum required is to enter LCO 3 ..0.3. The actions of ITS LCO 3.0.3 are to initiate action within 1 hour to place the plant, as applicable,

        . in MODE 3 within 7 hours; in MODE 4 within 31 hours; and in MODE 5 within 37 hours. Since this change will require the plant to be in a lower MODE in a shorter time frame, this change is considered more restrictive. This change continues to assure that a plant shutdown can be achieved in a controlled manner without challenging plant systems. This change is an additional restriction on plant operations and is consistent with NUREG-1432.

M.3 CTS 3.17.1.6b requires the reactor to be in a condition where the affected equipment is not required wiaTin 48 hours. ITS 3. 3 .1, Condition G Actions require the same condition be achieved within 6 hours. Since the plant is required to be in a lower MODE in a shorter time frame, this change is considered more restrictive. With the plant required to be in MODE 3 in 6 hours, de-energizing the clutch power supplies or

       . borating to the REFUELING BORON CONCENTRATION can also be performed within the same time frame. These actions will not challenge plant systems, since the reactor is subcritical in MODE 3. This change is consistent with NUREG-1432.

M.4 Not Used M.5
CTS Table 3.17.1 Footnote (c) provides an allowance to change the setpoint of the low power bypass setpoint from 104 % RTP to 10-1 % RTP during LOW POWER PHYSICS TESTING. Since this allowance is not used, it is not included in the ITS. Since .this.

change deletes an allowance and the plant will*no longer be able* to change this low power setpoint during the ITS PHYSICS TEST (see Discussion of Changes for ITS Section 1.0), this change is considered more restrictive. This change is consistent with NUREG-1432. M.6 New Conditions Band C and new Required Actions B.1 and C.l have been added for I

~~

the Loss of Load and High Startup Rate Functions. These actions require restoration of the inoperable channel(s) prior to re-:entering operational conditions during which the Functions are required to be Operable. CTS Actions applicable to single channel I I inoperability for other RPS trip Functions are not required for High Startup Rate and J -. rn Loss of Load. In ITS, some Action must' be specified to avoid an LCO 3.0.3 entry I ff! r---@ ACTIONS _- - - - - - - - - - - - - - - - - - - - - *- - - - - - - - - - - - - - - NOTE- - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - Separ~te Condition entry is allowed for each P'S tjrip Of'bySjl"ss #movflf----125' Funct1on. * ' ~

-~----------------------------------------------------------------------------
REQUIRED ACTION COMPLETION TIME RAI [3.17.l.2JA. I3.3.1-6 Not opphcoble to High Stortup Rate, Loss of Lood or ZPM
~ Place ~ffected unit in tr,ip.
trip
]
Bypass Removal Functions.
< Insert 1 > © (continued) I RAI 3.3.1-1
(Poli sades Nuclear Plant) lftoGjts I . 3. 3-1 5-p

SECTION 3.3 INSERT 1 KAI.
3,3,\-1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME B. One High Startup Rate trip B.l Restore trip unit and Prior to entering unit or associated associated instrument MODE 2 from instrument channel channel to OPERABLE MODE3 inoperable. status. C. One Loss of Load trip unit C .1. Restore trip unit and Prior to increasing or associated instrument associated instrument THERL\1AL POWER to channel inoperable. channel to OPERABLE ~ 17% RTP following status. entry into MODE 3
3.3-1 5-rr
---------NOTE----------
Not applicable to ZPM Bypass Removal RPS Instrument at i onb?'Opej?4t i nj' (Anf og
. 3.
p Function. CONDITION REQUIRED ACTION COMPLETION TIME
------------NOTE-------------
LCO 3.0.4 is not applicable. (3.17 .1.3.o] 1 hour RAI 3.3.1-1 3.3.1-6
~hours (3.17.1.2) fu}1mmed1otely)
o. 0.1 IYnou
[3.17.1.3] QR 0.21J] (continued) RAI
I3.3.1-1
< Insert 1 > @
( Pohsodes Nucleor Plont) [#oG/8fs I 3.3-2 5-A !Afv 1/ 04/A!/9fil
SECTION 3.3 INSERT 1 RA-r.
3.~.t-1 G. Required Action and G.1 Be in MODE 3. 6 hours associated Completion Time not met. ANl2 QR G.2.1 Verify no more than 6 hours one full-length Control room ambient air control rod is capable temperature > 90°F. of being withdrawn. *' QR G.2.2 Verify PCS boron 6 hours concentration is at REFUELINQ BORON CONCENTRATION
3.3-2
. @t RPS InstrumentationE:O)!;.ribing 'b\na~
ACTIONS CONDITION REQUIRED ACTION
. 3. .1 COMPLETION TIME Restore bypass removal channel a fected trip unit to PERABLE status.
QR
.2.2.2Place af ted trip units in t E. One or more unctions ------------NOTE-----------~-
with two aut at1c LCO 3.0.4 is no applicable.
~ bypass removal channels 1nopera le.
E.l Disable bypa channels
Pl ace one affected

trjp unit in bypass and*place the other in lri p for each affected trip unction.
one bypass [48] hours nd the associat trip unit to OPERABL status for each af cted trip Function. F. Required F.l 6 hours associated mpletion Time not met . CEOG STS 3.3-3 Rev 1, 04/07/95 5-J-
SECTION 3.3
INSERT 1 Table 3.3.1-1 (page 1 of 2)
Reactor Protective System Instrumentation APPLICABLE SURVEILLANCE
}2A1:..
j.~.1-1 3.3,\-(o FUNCTION MODES
REQUIREMENTS ALLOWABLE VALUE
1. Variable High Power Trip SR 3. 3 .1.1 ~ 15% RTP above SR 3. 3 .1. 2 current THERMAL SR 3 .3 .1. 3 POWER with a SR 3.3.1.4 minimum of ~ 30%
SR 3.3.1.5 RTP and a maximum SR 3.3.1.6 of ~ 106.5% RTP SR 3.3.1.8
2. High Startup Rate Tri p<bl 1,2 SR 3. 3 .1.1 NA SR 3. 3 .1. 7 SR 3.3.1.8

3. Low Primary Coolant System Fl ow Tri p(cl SR 3. 3 .1.1 ~ 95%
SR 3. 3 .1. 5
\_ SR 3. 3 .1.8
4.
5. Low Steam Generator A Level Trip Low Steam Generator B 1,2,3<*l,4<*J,5(al SR 3. 3 .1.1 SR 3. 3 .1.5 SR 3.3.1.8
~ 25.9% narrow range Level Trip 1,2,3(*l,4(al,5(al SR 3. 3 .1.1 ~ 25.9% narrow SR 3 .3 .1.5 range SR 3.3.1.8
6. Low Steam Generator A Pressure Tri p<cl 1,2,3<* 1,4<al ,5C*l SR 3. 3 .1.1 :<: 500 psia SR 3.3.1.5 SR 3 .3 .1.8 -

7. Low Steam Generator B Pressure Trip<cl 1,2,3<*l ,4<*l ,5(*l SR 3. 3 .1.1 :<: 500 psia SR 3. 3 .1. 5 SR 3.3.1.8

8. High Pressurizer Pressure Trip 1,2,3<*l,4(al,5Cal SR 3. 3 .1.1 ~* 2255 psi a SR 3 .3 .1.5 SR 3.3.1.8 (a) With more than one full-length control rod capable of being withdrawn and PCS boron concentration less than REFUELING BORON CONCENTRATION.
(b) Trip may be bypassed when Wide Range Power is < lE-4% RTP or when THERMAL POWER is
> 13% RTP.
(c) Trips may be bypassed when Wide Range Power is< lE-4% RTP. Bypass shall be automatically removed when Wide Range Power is ~ lE-4% RTP. 3.3-6 5-u
SECTION 3.3
INSERT 1 Table 3.3.1-1 (page 2 of 2)
Reactor Protective System Instrumentation
*eu:.
3.'!.l-1
~.'!.\-~
APPLICABLE SURVEILLANCE I. FUNCTION MODES REQUIREMENTS ALLOWABLE VALUE I
9. Thennal Margin/
Low Pressure Trip<~ 1, 2' 3(*) '4 (*) '5 (*) SR 3.3 .1.1 Table 3.3.1-2 SR 3.3.1.2 SR 3.3.1.3 SR 3.3.1.4 SR 3.3 .1.5 SR 3.3 .1.6 SR 3.3.1.8
10. Loss of Load Trip 1(d) SR 3. 3.1. 7 NA SR 3.3.1.8

11. Containment High Pressure Trip 1, 2, 3(a) , 4(at, 5(a) SR 3.3.1.5 ~ 3~70 psig SR 3.3.1.8

12. Zero Power Mode
Bypass Automatic Removal l,2,3(a) ,4Cal ,5Ca) SR 3.3.1.8 *NA (a) With more than one full-length control rod capable of being withdrawn and PCS boron concentration less than REFUELING BORON CONCENTRATION.
(c) Trips may be bypassed when Wide Range Power is < lE-4% RTP. Bypass shall be automatically removed when Wide Range Power is ~ lE-4% RTP. (d) When THERMAL POWER is ~ 17% RTP .
.b-v 3.3-7
ATTACIDIE~T 6 JUSTIFICATION FOR DEVIA TIO!'IS SPECIFICATION 3.3.1, RPS INSTRUNIE:\i'TATIO~
Discussion
~Al -s.~.1-f. ~::~bility has been *sed to b~stent with t~en~basis.
f~e flltA) Ir-t:> ~ syos, Rev. 1, is i tporated onifiis necessary tG-'fetlect ttreCurrent li~ensing as1s.
9. The Palisades RPS design does not include Reactor Trip Circuit Breakers (RTCBs).
therefore any reference to these in ITS 3. 3. 1 has been deleted.
10. ISTS 3.3. l Required Action A. l and A.2.1 have been deleted since the remaining OPERABLE channels can provide the required trip even with another channel failure.
The requirement to place the channel in trip within 7 days is considered adequate. These. allowance were approved in Amendment 162 of the Palisades Operating license.
11. The Palisades RPS design does not include Core Protection Calculators (CPCs).
therefore any reference to ~hese in ITS 3.3. l has been deleted .
I
\
12. 13. The requirement to place one trip unit in bypass has been deleted since this action does not restore the trip capability of the affected Function. The Palisades RPS design does not include Axial Power Distribution (APD) Trips, therefore any reference to these in ITS 3.3. l has been deleted.
14. ISTS SR 3.3. l. 7 has been delgted since the 18 month CHANNEL CALIBRATION surveillance (ITS SR 3. 3. l .T; 1s considered adequate to ensure the bypass removal channels are functioning properly. This change is consistent with the plant current licensing basis. Subsequent surveillances have been renumbered, where applicable.

15. ISTS SR 3.3.1.9, the RPS RESPONSE TIME test is not included in the ITS. This test is not required by the current licensing basis since the conclusions of NUREG-082, "Integrated Plant Safety Assessment Systematic Program Palisades Plant" - Final Report USNRC 10/82, concluded that backfitting this testing into the current program was not required.

16. ISTS 3. 4 .17, "RCS Loops - Test Exceptions" is not being proposed therefore reference to this specification is deleted.
\
Palisades Nuclear Plant Page 5-w 2 of 4 01120798
ATTAC1'ENT 6
Change Discussion JUSTIFICATION FOR DEVIATIONS SPECIFICATION 3.3.1, RPS INSTRUMENTATION
8. The Applicability has been revised to be consistent with the current licensing basis.
TSTF-85, Rev. 1, is incorporated, but revised as necessary to reflect the current licensing basis. The CTS applic~bility reflects the Palisades safety analyses assumptions with respect to RPS operation. The CTS RPS LCO applicability was issued as part of Amendment 162. The broader ITS applicability encompasses the
*applicable conditions of STS LCOs 3.3.1and3.3.2. The ITS (andCTS) wording "when more than one [full length] control rod is capable of being withdrawn" replaces the STS wording "with any RTCB's closed and any control element assemblies capable of being withdrawn".
I 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. The part length control rods have no clutches, remain fully withdrawn during operation, and are unaffected by c;;t
RPS functions . Q::'
r Palisades is not equipped with ,RTCBs; power to the CRDM clutches is interrupted, to 1-initiate a scram, by de-energization of normally energized contactors. These contactors 1 are addressed in ITS LCO 3:3.2. Their functioning is explained in the Bases. I
9. The Pallisades RPS design does not include Reactor Trip Circuit Breakers (RTCBs)~
therefore any reference to these in ITS 3.3.l has been deleted.
10. ISTS 3.3.1 Required Action A.1 and A.2.1 have been deleted since the remaining OPERABLE channels can provide the required trip even with another channel failure.
The requirement to place the channel in trip within 7 days is considered adequate. These allowance were approved in Amendment 162 of the Palisades Operating license.
11. The Palisades RPS design does not include Core Protection Calculators (CPCs),
therefore any reference to these in ITS 3.3.1 has been deleted.
12. The requirement to place one trip unit in bypass has been deleted since this action does not restore the trip capability of the affected Function. t

13. The Palisades RPS design does not include Axial Power Distribution (APD) Trips, therefore any reference to these in ITS 3 .3 .1 has been deleted .
5-Palisades Nuclear Plant Page 2 of 4 04/30/99
,; CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO JANUARY 6, 1999 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.3, INSTRUMENTATION NRC REQUEST: 3.3.1-1a JFD 6 Comment: Revise JFD 6 to give a design basis justification for not adopting the STS RPS Instrumentation - Shutdown LCO. Consumers Energy Response: JFD 6 for LCO 3.3.1 does not deal with omission of LCO 3.3.2. It is assumed that the NRC's intent was to revise JFD 6 for LCO 3.3.2, which deals with it explicitly. JFD 6 for LCO 3.3.2 has been revised. Affected Submittal Pages: Att 6, JFD 3.3.2, page 1 of 1 (former submittal) Att 6, JFD 3.3.2, page 1 of 1
6
, r
Change SPECIFICATION 3.3.2, RPS INSTRUMENTATION - SHUTDOWN Discussion ATTACHMENT 6 JUSTIFICATION FOR DEVIATIONS 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. Deviatio.ns 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.
4.
The following requirements have been renumbered, where applicable, to reflect this deletion. 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.
Palisades Nuclear Plant Page 1of1

01/20/98
ATTACI'ENT 6
Change JUSTIFICATION FOR DEVIATIONS SPECIFICATION 3.3.2, RPS INSTRUMENTATION - SHUTDOWN Discussion Note: This attachment provides a brief discussion of the deviations from NUREG-1432 that wen~ 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
4.
deletion . 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. ISTS LCO 3.3.2, "RPS Instrumentation- Shutdown," is not used. The ISTS I LCO 3.3.2 addresses the Power Rate of Change - High trip. The Palisades design I does not credit the comparable trip, High Startup Rate, in any of the Safety Analyses and that trip is not required to be Operable by the CTS. This lack of reliance is part of ld the plant design; the High Startup Rate trips are not installed* as safety grade I -;-
c~mponents. Their design is discussed in some detail in the Bases for LCO 3.3.1. Therefore, this specification is not applicable to the Palisades Nuclear Plant. Instead, II cv) the CTS LCO 3.17.1 and ITS LCO 3.3.1 utilize an applicability includes all conditions I r0 when the Palisades analyses credit functioning of the RPS to trip the reactor. That 1_ applicability was approved as part of Amendment 162 to the Palisades operating I (£_ license. TSTF-82, Rev. 1, and TSTF-180 are also not incorporated on the same basis . I O:::"
Palisades Nuclear Plant Page 1of1 05/30/99
CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO JANUARY 6, 1999 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.3, INSTRUMENTATION NRC REQUEST: 3.3.1-1b Comment: Verify ITS surveillances in LCO 3.3.1 require testing Variable High Power Trip Function reset feature for power ascension and decension. Provide a Bases discussion giving the basis for the TS required testing. Consumers Energy Response: The Variable High Power Trip reset function is tested during the Channel Calibration, CTS Table 4.17.1, footnote "d" or ITS SR 3.3.1.8 .. The surveillance procedure verifies,appropriate resetting of the trip setpoint at numerous points during a simulated power increase and decrease. The bases for SR 3.3.1.8 has been revised to discuss this testing. Affected Submittal Pages: Att 2, ITS B 3.3.1, page B 3.3.1-31
7
RPS Instrumentation B 3.3.1
BASES SURVEILLANCE SR 3.3.1.8 REQUIREMENTS (continued) SR 3.3.1.8 is the performance of a CHANNEL CALIBRATION every 18 months.
CHANNEL CALIBRATION is a complete check of the instrument channel including the sensor. The Surveillance verifies that the channel responds to a measured parameter within the necessary range and accuracy. CHANNEL CALIBRATION leaves the channel adjusted to account for instrument drift between successive calibrations to ensure that the channel remains operational between successive tests. CHANNEL CALIBRATIONS must be performed consistent with the plant specific setpoint analysis.
~The as found and as left values must also be recorded reviewed for consistency with the assumptions of the and frequency extension analysis. The requirements for this review are outlined in Reference 5.
As part of the CHANNEL CALIBRATION of the wide range Nuclear Instrumentation, automatic removal of the ZPMB for the Low PCS Flow, TM/LP, and Low SG Pressure trips and the automatic (operational) bypassing of the Loss 9f Load and High Startup Rate trips must be verified to assure that these trips are available when required. The Frequency is based upon the assumption of an 18 month calibration interval for the determination of the magnitude of equipment drift. The neutron detectors are excluded from CHANNEL CALIBRATION because they are passive devices with minimal drift and because of the difficulty of simulating a meaningful signal. Slow changes in detector sensitivity are compensated for by performing the daily calorimetric calibration (SR 3.3.1.3) and the monthly linear subchannel gain check (SR 3.3.1.4) .
Palisades Nuclear Plant B 3.3.1-31 01/20/98 7-_)__

Insert for SR 3.3.l.8 Bases:
The bistable setpoints must be found to trip within the Allowable Values specified in the LCD and left set consistent with the assumptions of the setpoint analysis. The Variable High Power Trip setpoint shall be verified to
' I .
reset properly at several power levels during indicated power increases and power decreases . 1-b
CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO JANUARY 6, 1999 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.3, INSTRUMENTATION NRC REQUEST: 3.3.1-2 ITS Required Action B.1 JFD 12 . The ITS markup eliminates the Action to place one channel in bypass whereas the Bases markup includes discussion that the 3.0.4 provisions are suspended for this Condition because one inoperable channel is tripped and the other is bypassed. Comment: Provide discussion to support the 3.0.4 exception without relying on channel bypass features. Consumers Energy Response: The Bases for Actions B.1 and B.2 have been revised to discuss having one inoperable channel bypassed as an allowance, rather than as a requirement. The bypassing of an inoperable channel provides no safety benefit, since a bypassed channel cannot provide a trip function. The LCO 3.0.4 exception is based upon having one inoperable channel in trip, which places the plant in a condition where any additional channel of that parameter reaching a trip setting would cause a reactor trip. Actions B.1 and B.2 of our January 26, 1998 submittal have .been renumbered as Actions E.1 and E.2 as part of the changes made in response to RAI 3.3.1-1. Bases pages, marked to
show the changes are attached .
Revised markups of the STS bases have not been included. Affected Submittal Pages: Att 2 ITS B 3.3.1, page B 3.3.1-29 (Mark-up) Att 2 ITS B 3.3.1, page B 3.3.1-30 (Mark-up)
8
RPS Instrumentation B 3.3.1
BASES ACTIONS (continued)
D.l and D.2 Condition D applies when one or more automatic ZPM Bypass removal channels are inoperable. If the ZPM Bypass removal channel cannot be restored to OPERABLE status, the affected ZPM Bypasses must be immediately removed, or the bypassed RPS trip Function channels must be immediately declared to be inoperable. Unl~ss additional circuit failures exist, the ZPM Bypass may be removed by placing the associated "Zero Power Mode Bypass" key operated switch in the nonnal position. A trip channel which is actually bypassed, other than as allowed by the Table 3~3.1-1 footnotes, cannot perfonn its specified safety function and must immediately be declared to be inoperable. D.1 and D.2 Cenditien D applies te ene er twe autematic (eperating) bypass remeval channels ineperable. If the bypass remeval channel fer any eperating bypass cannet be restered te OPERABLE status, the asseciated RPS instrument channel may be censidered OPERABLE enly if the bypass is net in effect. Otherwise, the affected RPS channel must be declared ineperable, and the bypass either remeved er the bypass remeval channel repaired. This is addressed by requiring entry inte the apprepriate Cenditien fer the channels rendered ineperable by the bypass channel failure. EB.1 and EB.2 Condition EB applies to the failure of two channels in any RPS automatic trip Function, except lligh Startup Rate and Less ef Lead. Cenditien B is medified by a Nete stating that this Cenditien dees net apply te the lligh Startup Rate and Less ef Lead Functiens. The Required Actions are ts-;nodified by a Note stating that LCO 3.0.4 is not applicable. The Note was added to allow the changing of MODES even though two channels are inoperable, with one channel (trip channel) bypassed and ene tripped. MODE changes in this configuration are allowed because two trip channels for the affected function remain OPERABLE. A trip occurring in either or both of those
channels would ~ause a reactor trip .
Palisades Nuclear Plant 03/25/99
t RPS Instrumentation B 3.3.1 BASES ACTIONS EB.1 and EB.2 (continued} While it is conceptually possible that, if the two operable channels were those that do not have total channel separation in their cable routings, a single failure could disable both from tripping, in reality, such failures are extremely unlikely. Most failures involving a common cable fault would cause the affected channel(s) to fail in the de-energized condition, thereby initiating a reactor trip not preventing one. to p'ermit mai flteflaf!ce afl a testi fig of! of!e of the il'loperaBle chaf!flels. In this configuration, the protection system is in a one-out-of-two logic, and the probability of a common cause failure affecting both of the OPERABLE channels during the 7 days permitted is remote. Required Action EB.1 provides for placing one inoperable channel in trip within the Completion Time of 1 hour. Though not explicitly required. the other inoperable channel may shoula be (trip chaflflel) bypassed. If it is !'lot Bypassed, leaviflg ofle il'loperaBle trip Fuflctiofl ifl aH ufltrippea coflaitiofl, aamiflistrative cof!trols are proviaea to preveflt iflaaverteflt Bypass of the same Fuflctiofl ifl af!other chaflflel. Such iflaaverteflt Bypass coula aefeat three of the four RPS chaflflels, ref!aeriflg the RPS ifloperaBle. This Completion Time is sufficient to allow the operator to take all appropriate actions for the failed channels while ensuring that the risk involved in operating with the failed channels is acceptable. With one channel of protective instrumentation bypassed or inoperable in an untripped condition, the RPS is in a two-out-of-three logic for that function; but with another channel failed, the RPS may be operating in a two-out-of-two logic. This is outside the assumptions made in the analyses and should be corrected. To correct the problem, one of the inoperable channels is placed in trip. This places the RPS in a one-out-of-two for that function logic. If any of the other unbypassedOPERABLE channels for that function receives a trip signal, the reactor will trip. Action E.2 is modified by a Note stating that this Action does not apply to the High Startup Rate and Loss of Load Functions . Palisades Nuclear Plant B 3.3.1-30 03/25/99
CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO JANUARY 6, 1999 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.3, INSTRUMENTATION NRC REQUEST: 3.3.1-3 Resolved during 10/27/98 meeting, Consumers Energy not required to respond. NRC REQUEST: 3.3.1-4 CTS 4.17.1, footnote (b) ITS SR 3.3.1.3 CTS Table 4.17.1, footnote b, requires calibration of the Variable High Power Function through comparison with Heat Balance every 24 hours when RTP is greater than 15%. ITS SR 3.3.1.3 adds an allowance where this surveillance is only required when the absolute difference between the calorimetric calculation is 2.0%. STS SR 3.3.1.3 provides this allowance below 1.5% absolute difference. Comment: There is no discussion for the Less Restrictive Change to CTS requirements or justification for deviation from the STS for this change. Provide missing documentation. Consumers Energy Response: The formerly proposed 2.0% was based on our former procedural acceptance criterion. Actual practice has been to readjust the instrumentation if deviation from calorimetric results by approximately 1%. The value in ITS SR 3.3.1.3 and the associated Bases has been revised from 2% to 1.5% to agree with the STS. DOC A.8 has been revised to discuss the administrative nature of discussing, within the Technical Specifications, when an adjustment must be made. Affected Submittal Pages: Att 1ITS3.3.1, page 3.3.1-4 Att 3 3.3.1 DOCS, page 3 of 11 Att 5 NUREG, page 3.3-4
9
RPS Instrumentation 3.3.1 SURVEILLANCE REQUIREMENTS
-------------------------------------NOTE---~---------------------------------
Refer to Table 3.3.1-1 to detennine which SR shall be perfonned for each Function. SURVEILLANCE FREQUENCY SR 3.3.1.1 Perf onn a CHANNEL CHECK.
12 hours SR 3.3.1.2 Verify control room temperature is~ 90°F. 12 hours SR 3.3.1.3 -------------------NOTE--------------------
Not required to be performed until 12 hours after THERMAL POWER is ~ 15% RTP. ,. E.~c.o~ E.D Perfonn calibration (hea balance only) and 24 hours adjust the power range and ~T power channels to agree with calorimetric calculation if the absolute difference is I
~ ~AJ:. .
3.'3.\-4 SR 3.3.1.4 31 days SR 3.3.1.5 92 days SR 92 days
Palisades Nuclear Plant 9-()_/
3.3.1-4 Amendment No. 01/20/98
A.7 5 ee..
te\)) :)e c1 /J,,. 7 ATTACHMENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.3.1, RPS INSTRUMENTATION I CTS Table 3 .17 .~Fa tnote (b) provides a restriction that the low power -byp.ass --- (i.e., Zero Power ode Bypass) cannot be enabled unless the SHUTDOWN BORON CONCENTRAT N for the COLD SHUTDOWN condition is achieved. This . provision is noy1ncluded in the ~ssociated ITS Table 3. 3 .1-1 Footnote (b) s~rice the SHUTDOWN/MARGIN (SDM) requirements are m:et through ITS SectioIY'3 .1. The CTS definitj6'n of SHUTDOWN BORON CONCENTRATION required j~ > 2% ~p with all c9tltrol rods insert~d in the core and~the 6ighest worth control r-Od fully withdraw'n. In ITS Specification 3 .1.1, the m*
um SDM required dUring any cooldjJ~n must be >2r/ ~p and is increased o > 3.53 ~p when*~ PCS average

temp'erature is reduce below 525°F. Tterfore, since the CTS SHUTDOWN BORON C_¢NCENTRATIO requirements are m 'ntained throughout the plant cooldown, the 1
removal of this p, vision can be conside ed administrative. This change is consistent i with NUREG-1432. ----- --**-~---------------*----f A.8 CTS ~a e ~.17 .1 Foo note (b) requires a calibratio~-of fu~ Variabl~- --i;h P~~er. . . --) Funct10 with a heat alance to be pe ormed when p wer is > 15% TP. In add1t1on Foot te (c) requir the excore ch els to be calib ated with ate signal. In ITS Tab 3.3.1-1, th e surveillances SR 3.3.1.3 and R 3.3.1.6) a also associated with th Thermal M gin/Low Press e (TM/LP) Tri unction. T s association is necessary sine there is no ad tional testing re ired. This c ange is considered l administrative, and is consist nt with NUREill_'!l~'.__________ __ ---- --------- - A.9 CTS 2.3 specifies the requirements for the RPS Limiting Safety System Settings. The Applicability of this Specification is when the associated RPS channels are required to be OPERABLE by Specification 3.17.1. In addition, CTS 2.3.1 specifies that if the* RPS instrument setting is not within the allowable settings of Table 2 .3 .1, the instrument mus_t be declared inoperable and complete corrective action as directed by Specification 3 .17 .1. In the ITS the RPS Allowable Values are listed in ITS Table 3.3.1-1 and the ITS 3.3.1 Actions provide identical protection. This change essentially moves requirements from one Technical Specification to another, and therefore, is considered administrative. This change is consistent with NUREG-1432 .
Palisades Nuclear Plant Page 3of9 1-h 01/20/98
ATTACHMENT 3
A.7 DISCUSSION OF CHANGES SPECIFICATION 3.3.1, RPS INSTRUMENTATION CTS Table 3 .17 .1 Footnote (b) provides a restriction that the low power bypass (i.e., Zero Power Mode Bypass) cannot be enabled unless the SHUTDOWN BORON CONCENTRATION for the COLD SHUTDOWN condition is achieved. This provision is not included in the associated ITS Table 3.3.1-1 Footnote (b) since the SHUTDOWN MARGIN (SDM) requirements are met through.ITS Section 3.1. The CTS definition of SHUTDOWN BORON CONCENTRATION required is > 2 % .t!.p with all control rods inserted in the core and the highest worth control rod fully withdrawn. In ITS Specification 3.1.1, the minimum SDM required during any cooldown is specified in the COLR, and the COLR requirements must be calculated in 1
accordance with approved methodology. Changes to the COLR must be performed in accordance with the requirements of ITS 5.6.5. The COLR SDM requirements provide the same assurance that the reactor will remain shutdown as do those in CTS. Therefore, since appropriate SDM must still by maintained throughout plant operation and cooldown, the removal of this provision can be considered administrative. This change essentially moves requirements from one Technical Specification to another, and therefore,- is considered administrative. This change is consistent with NUREG-1432. A.8 CTS Table 4.17.1 Footnotes (b) and (c) require, a calibration of the Variable High Power trip Function channels (excore nuclear power and llT power) with a heat balance, and calibration of the excore power channels with a test signal. In ITS, a requirement to adjust the instrumentation if the absolute difference between the instrument readings and the results of the heat balance exceeds 1.5 %. That requirement corresponds to "the necessary range and accuracy" requirement of the Channel Calibration definition, and does not constitute a change in requirements. Also
-J .
in ITS, Table 3. 3 .1-1, these surveillances (SR 3. 3 .1. 3 and SR 3. 3 .1. 6) are also associated with the Thermal Margin/Low Pressure (TM/LP) Trip Function. In CTS, this association is provided by a table in the bases. This association is necessary since the power signal from the excore power range is an input to the TM/LP trip function setpoint calculator. There is no additional testing required. This change is considered
administrative, and is consistent with NUREG-1432.
A.9 CTS 2.3 specifies the requirements for the RPS Limiting Safety System Settings. The Applicability of this Specification is when the associated RPS channels are required to be OPERABLE by Specification 3.17.1. In addition, CTS 2.3.1 specifies that if the RPS instrument setting is not within the allowable settings of Table 2. 3 .1, the instrument must be declared inoperable and complete corrective action as directed by Specification 3.17.1. In the ITS the RPS Allowable Values are listed in ITS Table 3.3.1-1 and the ITS 3.3.1 Actions provide identical protection. This change essentially moves requirements from one Technical Specification to another, and
therefore, is considered administrative. This change is consistent with NUREG-1432.
Palisades Nuclear Plant Page 3of11 05/30/99
~*
RPS Instrumentat1o~~erabijlg J.4,na~p
SURVEILLANCE REQUIREMENTS
3. .
-------------------------------------NOTE-------------------------------------
Refer to Table 3.3.1-1 to determine which SR shall be performed for each~ ~ i
\ . .,,
Function.
SURVEILLANCE FREQUENCY
[11\.11.1] SR 3.3.1.1 12 hours
CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO JANUARY 6, 1999 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.3, INSTRUMENTATION NRC REQUEST: 3.3.1-5 CTS 4.18.2.1 (b) CTS 4.18.2.1 (b) requires individual excore channel measured Axial Offset (AO) compared to the total core AO measured by the incore channels to determine if the excore monitoring system must be recalibrated. ITS SR 3.3.1.4 (Insert 2) introduces the acronym ASI for the excore channel comparison. Comment: It is not clear that ASI and Axial Offset (AO) are equivalent terms. Provide clarification and appropriate change documentation. Consumers Energy Response: CTS Section 1.0 contains the following definition: AXIAL OFFSET or AXIAL SHAPE INDEX shall be the ratio of the power generated in the lower half of the core minus the power generated in the upper half of the core, to the sum of those powers. Section 1.0 of the ITS conversion request proposes two definitions, one for Axial Offset, and one for Axial Shape Index to accommodate terminology used with newer RPS components. They differ only in that ASI is measured with the power range excore detectors while AO is
measured with the incore detectors.
The ITS definitions are: AXIAL OFFSET (AO): AO shall be the power generated in the lower half of the core less the power generated in the upper half of the core, divided by the sum of the power generated in the lower and upper halves of the core (determined using the incore monitoring system). AXIAL SHAPE INDEX (AS/): AS/ shall be the power generated in the lower half of the core less the power generated in the upper half of the core, divided by the sum of the power generated in the lower and upper halves of the core (determined using the ex<;ore monitoring system).
Affected Submittal Pages:
No page changes .
10
CONVERSION TO IMPROVED TE:CHNICAL SPECIFICATIONS RESPONSE TO JANUARY 6, 1999 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.3, INSTRUMENTATION NRC REQUEST: 3.3.1-6 CTS 3.17.1, M.1 ITS Required Action D, JFD 5 DOC M.1 states that CTS 3.17.1 does not provide specific requirements for inoperable automatic bypass removal channels that are associated with certain RPS instrument functions. ITS 3.3.1 places a requirement in the LCO that applicable automatic bypass removal channels must be operable. ITS Condition D applies to one or two inoperable bypass removal channels. Required Action D.1 requires the inoperable bypass removal channel to be disabled within 1 hour or Required Action D.2 provides the option to declare affected trip units inoperable within one hour from time of discovery of the inoperable condition which allows another 7 days to place a'trip unit in trip. The ISTS Condition D specifies requirements for one inoperable automatic bypass removal channel and Condition E specifies requirements for two inoperable automatic bypass removal channels. ISTS Required Action D.1 'requires the inoperable channel to be disabled within 1 hour. Required Action D.2.1 provides the option to place the trip units in bypass or trip and then enter D.2.2 to restore the bypass removal channel. Thus, the ISTS requires an action within one hour place the automatic bypass removal channel in a safe configuration. JFD 5 states the ITS reflects the current licensing basis. Comment: The staff notes, that DOC M.1 states that CTS do not provide specific requirements for inoperable automatic bypass removal channels. Thus, the proposed ITS requirements to declare the trip units inoperable represents a generic change to the ISTS and a . change to the CTS. Revise'the ITS and associated DOCs to adopt the ISTS Conditions D and E. Comment: The staff also notes it is unnecessary for ITS required actions to state ,"declare the affected channel trip units inoperable and enter the appropriate Condition". The provisions of LCO 3.0.2 require TS actions to be met upon discovery of failure to meet an LCO. Consumers Energy Response:
\ '
The CTS, do not explicitly address inoperable bypass channels in a separate Action Statement, however, they do require selected bypass removal channels to be operable, and provide actions if they are not. The ITS requirements have been revised to preserve both the CTS operability requirements and the associated actions. The applicability for the requirement to have the bypass removal channels operable has been expanded to be the same as the associated trip functions. DOC 3.3.1 M.1 has been revised to more clearly explain the differences in presentation and applicability between the CTS'and the ITS. The ITS Actions differ from the STS because both the Palisades safety analyses and the Palisades RPS differ significantly from those discussed in STS Bases. ITS Action D.2, "Declare affected trip units inoperable and enter the appropriate Conditions", is necessary as an alternative action to "removing the bypass" in the unlikely event that the
manual switch could not be used to remove the bypass. The Completion Times for Actions D.1 and D.2 were changed to "Immediately" to remove the unintended 1 hour extension of CTS actions.
11 (continued)
CONVERSION TO IMPROVED TE-CHNICAL SPECIFICATIONS RESPONSE TO JANUARY 6, 1999 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.3, INSTRUMENTATION NRC REQUEST: 3.3.1-6 (continued) Consumers Energy Response: In developing the response to this NRC comment and the associated ITS revisions, a review vvas made of the effects on the plant circuitry and trip capability made by an inoperable bypass removal channel. These effects were compared to the assumptions of the safety analyses. As a result of these reviews, ITS LCO 3.3.1 bypass removal requirements, Condition D, and Action D.1 were revised to address only the Zero ~ower Mode (ZPM) Bypass removal channels. The ZPM bypass removal for the low PCS flow, low SG pressure and TM/LP trips are the only bypass removals assumed to function by the safety analyses. The other bypass removal channels affect only RPS trip functions for High Startup Rate and Loss of Load. As discussed in the ITS LCO 3.3.1 Bases, these two trips are installed only for equipment protection purposes, and are not assumed to function by any safety analyses. Table 3.3.1-1 was revised to add the ZPM Bypass Removal Function as a required Function with SR 3.3.1.8, an 18 month Channel Calibration (which must include a Channel Functional Test), as the required surveillance and an applicability the same as the associated trip functions. In addition, the parts of Section 3.3 Bases which discuss the RPS bypasses were revised and rearranged to add more detail.
Affected Submittal Pages:
Att 1, Att 1, Att 1, ITS 3.3.1, page 3.3.1-1 ITS 3.3.1, page 3.3.1-2 ITS 3.3.1, page 3.3.1-7 Att 1, ITS 3.3.1, page 3.3.1-8 Att 3, CTS, page 3-64 Att 3, CTS, page 3-65 Att 3, CTS, page 4-76 Att 3, DOC 3.3.1, page 2 of 9 (former submittal) Att 3, DOC 3.3.1, page 2 of 10 Att 3, DOC 3.3.1, page 5 of 9 (former submittal) Att 3, DOC 3.3.1, page 5 of 10 Att 3, DOC 3.3.1, page 6 of 10 Att 3, DOC 3.3.1, page 7 of 9 (former submittal) Att 3, DOC 3.3.1, page 7of10 Att 3, DOC 3.3.1, page 8 of 10 Att 5, NUREG, page 3.3-1 Att 5, NUREG, page 3.3-2 Att 5, NUREG, page 3.3-5 Att 5, NUREG, page 3.3-6 insert ,, Att 5, NUREG, page 3.3-7 insert Att 5, NUREG, page 3.3-8 insert
Att 6, Att 6, JFD 3.3.1, page 4 of 4 (former submittal)
JFD 3.3.1, page 4 of 4 12
RPS Instrumentation
3. 3. 1
3.3 3.3.1 INSTRUMENTATION Reactor Protective System (RPS) Instrumentation Cl-\ta.~'-le.LS, ti..~P Assoc..1 A.'T'Et:> t.C:Ro Pow~ Mooe (c?M LCO 3.3.1 Four RPS trip uni ts,~ associated instrument !MiQ} Bypass removal channels for each Function in Table 3.3.1-1 shall be OPERABLE.
APPL ICAB IL ITV: ACTIONS
-------------------------------------NOTE---------------*----------------------
Separate Condition entry 1s allowed for each Function. CONDITION REQUIRED ACTION COMPLETION TIME
A. -------~NOTE----------
Not applicable to High Startup Rate,IM@ Loss of Load,,_Functions. _____________________ ..:-"'I A~ 1 ( . Place affected trip unit in trip. (OR ~t-'M SY'PA~ l<\::-MOV!i.L) 7 days One or more Functions '!. with one RPS trip unit or associated instrument channel inoperable .
Palisades Nuclear Plant 3.3.1-1
/d)..-o-Amendment No. 01/20/98
- - - - - -~on=- - - - - - "10f A.??L\C.A6L.E- 1"o ~1(:,µ, RPS Instrumentation STA.RTuP RA."\E:- oR Los$ 3.3.1 o~ LOC..O hJt-l'-llO~S.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME
------------NOTE------------- ~-~~~~~-~~-~~:-~~~:~=~~:~~- ,__________ Place one trip unit in trip. KAI --- One or more Functions with two RPS trip 3.3.1-1 3.~.1-fD units or associated instrument channels Restore one trip unit 7 days inoperable. and associated instrument channel to OPERABLE status.
Two power range Restrict THERMAL 2 hours K'A.I channels inoperable. POWER to ~ 70% RTP . !.~. ,_'
D.
(l:>~MEt:">\~-ret..'{ ) IY~itl
~A!
QR _ _ _ _. - 3:5.1-" D.2
~~if !..i"l::L.0 One High Startup Rat-e Restore trip unit and. Prior to ~
Trip unit or associated instrument entering~~ associated instrument channel to OPERABLE from MODE~ e"1: channel inoperable. status. ~.1.1-1 One Loss of Load trip Restore trip unit and unit or associated associated instrument instrument channel channel to OPERABLE inoperable. status.
/di-6 Palisades Nuclear Plant 3.3.1-2 Amendment No. 01/20/98
RPS Instrumentation 3.3.1 Table 3.3.1-1 (page 2 of 2) Reactor Protective System Instrumentation A??l..\CA'&.E'" SURVEILLANCE FUNCTION Mo~s REQUIREMENTS ALLOWABLE VALUE
9. TheMMll Margin/Low Pressur1lill \ 1 Z 1 ~(~1 Y (o.~ '5 (o.) SR J.J.1.1 Table J.J.1-2 ii?'.1 P (c) SR J. J .1. 2 SR J.J.l.J SR J.J.1.4 SR J.J.1.5 SR J.J.1.6 SR J. J .1. 8

10. Loss of Load~TR1? SR 3.3.1.7 HA SR 3. 3 .1. 8 7(0..')' i{<t) '5 (o.)

11. Conhim1ent High Pressurel~1'P 1,2,"J ,..., 1 SR 3.3.1.5 ~ 3.70 psig

12. eE~o Fo>0E.iZ_ >Aot:E 8vl'A<;S I 'Z. !>(... } '-1(.:t.), 5(~) SR 3.3.1.8 Au10~0...,.1i:.. K'i::Mol/AL ' '~ ' * -i:.~........~--------.~-
.;R 3.~.l.b Bypass shal 1
n THERMAL POWER i~l7% RTP. ls)?ai"J eA"I
~3'.1-1 ~.~.I-~
ld.-c_
Palisades Nuclear Plant 3.3.1-7 Amendment No. 01/20/98
RPS Instrumentation 3.3.1 The Allowable Value for the Thennal Margin/Low Pressure Trip, P1 , 1, , is the higher of two values, P.1, and P" both in psia: P01 ,
1750
. P.,.
2012(QA) (QR 1) + 17 .O(T 1,) - 9493 Where:
QA* - 0.720(ASI) + 1.028; when - 0.628 ~ ASI < - 0.100 QA * - 0.333(ASI) + 1.067; when - 0.100 ~ ASI < + 0.200 QA * + 0.375(ASI) + 0.925; when+ 0.200 ~ AS! ~ + 0.565 ASI
Measured ASI when Q 2 0.0625 ASI = 0.0 when Q < 0.0625 QR 1

0.412(Q) + 0.588; when Q ~ 1.0 QR,

Q; when Q > 1. 0 Q

THERMAL POWER/RATED THERMAL POWER T.,

Maximum primary coolant inlet temperature, in °F AS!, T.,, and Qare the existing values as measured by the associated instrument channel.
/CJ.-d Palisades Nuclear Plant 3.3.1-8 Amendment No. 01/20/98

3 .~
3.3.1 Reoctor Protective System (RPS! Instrumentot1on INSTRUMENTATION l~ST~MS I RAI I3.3.1-6 RAI Toble' 3.3.1-1 3.3.1-1 ond Footnote o [l(17J';21 With one RPS trip unit or associated instrument channel inoperable for [ Cond A l one or more functions: [ RA A.1 J a)i* Pl ace the affected trip unit in the tripped condition within 7 days. IY17.)":31 With two RPS trip units or associated instrument channels inoperable for
*c Cond. E J one or more functions: < Add RA E NOTE >-@
[RA E.1 J a) Place one inoperable trip unit in the tripped condition within 1 hour, and RAI 3.3.1-1 [ Cond.F J b) If two Power Range Nuclear Instrument channels are inoperable, [ RA F.1 J limit power to s 70% RATED POWER within 2 hours. and [RA E.2 l c)* Restore one RPS trip unit and associated instrument channel to OPERABLE status within 7 days. SEE 3.3.2> l¥17 .)':6 I [ Cond. G ] [ RA G.1 J [ RA G.2.1 J RAI 3.3.1-1 Amendment No. -%-6r, 186 [ RA G.2.2 J Verify PCS boron concentrotion is ot REFUELING BORON CONCENTRATION. Add Cond. B and C >--@ -< Add Cond.O 0.1,~
< Add Required Actions B.l, *
ond C.1 .
M.6 Jd _e_ Add Required Actions ond 0.2 fC:Y Page 1 of 7
3.3.l
3.17 INSTRUf£NTATIQN SYSTEM$
Table Lg16@.1-1 ) Instrumentat1gn Operating Requirements Reactor protect1 Ye System A. 5 Required Penn1ss1ble RPS Operational EMD't1 QDI] llD it tblDDe]S Bypasses
1. (J!{nuefrrj( 7
..____....__.....,___....__..____.__.._____.__..___.....L..-----J.'---'/j'---o nJ?_._KsEE 3.3.2)
(l] 2. Variable H1gh Power(Trip) 4 None. [2)3. H1gh Start Up Rate 4 Below 1o°"<cl or fT3.3.t-1 1 above 13% RATED POWER. l£ootnote !b!.J [CJ]4. Thennal Marg1*n/ Low Pressure 4 (b) rn@ 3~~l-1 A.5 3.3.1-6 (8) s. H1gh *Pressurizer 4 None. Pressure , [3]~. LQW PCS Flow 4 (b)lE@-@ [10) 7
Loss of Load<dJ 4 Below 17% RATED POWER.
[4J a. Low *A* Steam 4 *None. Generator Level [5] g. Low *B* Steam 4 None. Generator Level [6)10. Low *A* Steam 4 (b)~ Generator Pressure [7)11. Low *B* Steam 4 (b) C!Jill-@ Generator Pressure [11] 12. 4 None. 13. 14. [121[@] [(ell~ RAI 3.3.1-1 3.3.1-6
<Add Footnotes A & B> Amendment No. 118, 139, 136, 162 3-65 /~-f Poge 2 of 7
3.3.1
4.17 INSTRUMENTATION SYSTEMS TESTS ~----"----
Table Lf<'l~(_E_.1-1 ) Instrymentation Surveillance Regyirements for Reactor protective System . [SR 3.3.1.11 [S~HKNRiL.51 [SR 3.3.t.81 CHANNEL FUNCTIONAL CHANNEL Functj onal Unit CHECK TEST CALIBRATION
1. l>fanuaf Trjj} Z Z Z /NA Z Z /{a)/ Z 7 NAJ--\sEE 3.3.2)
[1] 2.. Variable High Power 12 hours days (b, c, & d)
C2J 3. High Start Up Rate 12 hours (a) 18 months<el
[Cl] 4. Thermal Margin/ 12 hours 18 months Low Pressure C8J 5. High Pressurizer 12 hours days 18 months Pressure RAI [3] 6. Low PCS Flow 12.hours days 18 months 3.3.1-6 [10]7 . Loss of Load NA (a) 18 months
[4] 8.
[5] 9. Low '!A" SG Level Low "B" SG Level [6] 10. Low "A" SG Pressure [7] 11. Low 6 SG Pressure 11 11 12 hours 12 hours 12 hours 12 hours days days days days 18 months 18 months 18 months 18 months [11] 12. High Containment Pressure NA days 18 months
~: IT11~
CSR 3.3.l.5J 15 .. Thermal Margin Mani tor; Verify. constants each 92 days. A. 12 1 3 .~~~ 8 [SR 3.3.1.2] 16. erm Ma in oni r: Verify Control Room Temperature s; 90°F each 12 hours. CSR 3.3.1.7J (a) ~ Once wi th i n 7 days pri r to each reactor start up * <Add SR 3.3.!.3 Note~ . CSR 3.3.l.3J ( b) Calibrate with Heat Balance each 24 hours, when> 15% RATED POWER.~ CSR 3.3.l.6J Calibrate Excores channels with test signal each~ . CSR 3.3.1.Sl CHANNEL CALIBRATION each 18 months. [SR 3.3.l.8J ( e) Include verification of automatic Zero Power Mode Bypas.s removal. Apply to ITS Function cir\~ '--"".
Amendment No. -rTS, ~. 36, -t-5G, ~. ~.-rrl. 186 Poge 4 of 7
/di-~ 4~76
ATTACHMENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.3.1, RPS INSTRUMENTATION A.4 CTS 3 .17 .1. 6b requires the reactor to be placed in a condition where the affected equipment is not required. In the ITS, this requirement is presented by specifying the actual actions necessary to satisfy this requirement. ITS 3.3.1 Required Actions I.2.1 and I.2.2 require that no more than one control rod be capable of being withdrawn or that the PCS boron concentration be of the REFUELING BORON CONCENTRATION. These actions place the plant in a condition where the affected equipment is not required. Since this change only provides more specific actions to .be taken, it is considered administrative. A.5 CTS Table 3.17.1 contains a "Minimum Operable Channels" column. This column is deleted in the ITS because the Actions in the ITS are based on the number of channels inoperable, from the total number of channels, which is specified in LCO 3.3.1. When the number of inoperable channels in a trip Function exceeds that specified in any related Condition associated with the same trip Function, the plant is outside the safety analysis, and therefore, entry into ITS LCO 3.0.3 is required. This change only involves a change in presentation format, and is considered administrative. This change is consistent with the presentation format in NUREG-1432 .
A.6 Table 3.17.1 Footnote (a) requires two OPERABLE wide range nuclear instrum associated w1 channels if the Zero Power Mode (ZPM) bypass is used. This F e High Startup Rate Function. ITS 3.3.1 does not*
Footnote. There are o wide range nuclear instrumentation c nels which provide de this ote is input into the ZPM bypass. ch instrument channel
es input into two High -
Startup Rate channels and also pr
es the per * *ve signal for the ZPM bypass. The ZPM Bypass associated with these Fune 1s automatically removed when the wide range instruments indicate ~ 10 %4
. Since ITS Condition D provides adequate guidance when bypass chan s are inoperable, this ote is not necessary as the bypass channels will
er be disabled or the associated Hig up Rate channels will be plac * 'p. In addition, the footnote is associated with tti

h Startup Rat Functi , and this bypass is automatically enabled. Therefore, retaining t ote does provide any restriction to the operator .

Palisades Nuclear Plant Page 2 of 9 I c?<.-A.
01/20/98
l ATTACHMENT 3
A.3
, DISCUSSION OF CHANGES SPECIFICATION 3.3.1, RPS INSTRUMENTATION A Note was added to the Actions of CTS 3 .17 .1 which allows separate Condition entry for each RPS Function. The Note in ITS 3.3.1 provides explicit instructions for proper application of the Actions for Technical Specification compliance. In conjunction with the proposed Specification 1.3 - "Completion Times, " this Note provides direction consistent with the intent of the existing Actions for the RPS instrumentation. This change is consistent with NUREG-1432 (TSTF-178).
A.4 CTS 3 .17 .1. 6b requires the reactor to be placed in a condition where the affected equipment is not required. In the ITS, this requirement is presented by specifying the actual actions necessary to satisfy this requirement.* ITS 3. 3 .1 Required Actions I. 2 .1 and 1.2.2 require that no more than one cpntrol rod be capable of being withdrawn or that the PCS boron concentration be of the REFUELING BORON CONCENTRATION. These actions place the plant' in a condition where the affected equipment is not required. Since this change only provides more specific actions to be taken, it is considered administrative. A.5 CTS Table 3.17.1 contains a "Minimum Operable Channels" column. This column is deleted in the ITS because the Actions in the ITS are based on the number of channels
inoperable, from the total number of channels, which is specified in LCO 3.3.1. The total number .of channels required to be Operable is unchanged from CTS. ITS conditions do not depend on a specified minimum operable channels for entry into actions. The CTS use the specified minimum number of channels to initiate entry into a shutdown action statement. In ITS, when the number of inoperable channels in a trip 0
16 I Function exceeds that specified in any related Condition, the plant does not meet the LCO and no associated Action is provided, therefore, a shutdown is required in accordance with LCO 3.0.3. is required. The discussion of the change in Required Actions is discussed in DOC M.2. This change' only involves a change in presentation - format, and is considered administrative. This change is consistent with the presentation format in NUREG-1432. A.6
CTS Table 3.17.1 Footnote (a) requires two OPERABLE wide range nuclear I instrument channels if the Zero Power Mode (ZPM) bypass is used. This Footnote is I associated with the automatic ZPM bypass removal Function. ITS 3. 3 .1 does not I include this Footnote, instead, the automatic ZPM bypass removal function has been I specified as Function 12 in Table 3.3.1-1. ITS Condition D provides guidance when ,~
bypass channel(s) are inoperable, allowing indefinite continued operation with an I ,._:_ inoperable ZPM bypass removal channel only if the ZPM bypass is not used. These I l'<) two ITS features are the equivalent of the, CTS footnote (a) requirement that these I I<") bypass removal channels be operable when the bypass is used. Since the ITS
I ct:
requirements are equivalent to the CTS requirements, this change is considered to be administrative. I Q::-"" Palisades Nuclear Plant Page 2of11 05/30/99
/d-__,,C
5t::e ~~111se.~ fV1, I
f ATTACHMENT 3
'KAI,. 1~3.1-/ 3.3./-f:> DISCUSSION OF CHANGES SPECIFICATION 3.3.1, RPS INSTRUMENTATION I
TECHNICAL CHANGES - MORE RESTRICTIVE (M) M.l CTS 3 .17 .1 allows some Functions to not be OPERABLE, as allowed by the rmissible operational bypasses. CTS 3.17. l does not contain any specific r uirements for inoperable automatic bypass removal instrument channels a ociated
. witti
rtain RPS Instrument Functions. CTS Table 3.17.1 Functions 3, ( 1gh Start Up Rate), unction 4 (Thermal Margin/Low Pressure), Function 6 (Low P Flow),
Function (Loss of Load), Function 10 (Low "A" Steam Generator P, essure), and Function 1 (Low "'B" Steam Generator Pressure) all have at least e operational bypass featur associated with them. ITS 3.3. l places a require nt in the LCO that applicable bypa removal channels must be OPERABLE and e blishes Required ' Actions specifical for when one or two bypass removal cha els are inoperable (ITS Condition D). he Required Actions are essentially t same as when one or two RPS trip units or assoc ted instrument channels are ino able, except that the bypass channel is allowed to be
bled or removed.

ITS Table 3.3.1-1, Functions , 3, 6, 7, 9, and 10 re associated with bypass*
removal channels as indicated i he Table 3.3.1- Footnotes (a), (b) and (c}. The automatic bypass removal fea res are r red for each of these safety related trips to ensure the RPS Fune
ns are ot operationally bypassed when the safety arialysis assumes the Functio e OPERABLE. The specific RPS operating bypasses are discussed in the osed Applicable Safety Analyses

Bases for ITS 3.3.1.
The addition of this new require nt is conside more restrictive. This change will have no adverse im ct on safety beca se the action to remove the bypass channel when the auto atic bypass removal i inoperable ensures that the trip Function is OPE LE. This change is consi ent with NUREG-1432. New Conditions E and attd new Required Actions G.2 an H.2 have been. added for the Loss of ad and High Startup Rate Functions, d a Note has been added to the uired Actions specified in ITS 3.3.1 Con
ions A and B which states that onditions A and Bare not applicable to the Los of Load and High Startup te Functions. These actions require restoration of th inoperable c nel(s) prior to entering operational conditions during w 'ch the Functions ay be needed. These Required Actions and Completion Tim are consiste with current practice though not required by CTS for these Fune( ns for o and two inoperable channels. This Required Actions and Completion Ti s are acceptable since these Functions are not credited in the plant safety ysis. ' *

Palisades Nuclear Plant Page 5 of 9 01/20/98
ATTAC1'ENT 3
TECHNICAL CHANGES - MORE RESTRICTIVE (M)
DISCUSSION OF CHANGES SPECIFICATION 3.3.1, RPS INSTRUMENTATION M.1 The allowances for channels to be bypassed and the requirements for automatic removal of these bypasses is treated differently in CTS and ITS. Both treatments allow some trip Functions to be bypassed during conditions when the LCO requires them to be Operable. Both treatments require the bypasses on the safety related trips (those assumed to function by the safety analyses) to be automatically removed. Operability: In CTS, the automatic bypass removal channels (for safety related channel bypasses) are required to be Operable by footnote (a) of Table 3 .17 .1. (When the Wide Range Nuclear Instrument channels are Operable, if the indicated power level increases ab.ove the setpoint, the permissive signal is removed, automatically removing the bypass, ~ regardless of the position of the. manual bypass switch. This bypass is called the Zero Power Mode or "ZPM" bypass.) In ITS, the bypass removal channels are required to -I be Operable as part of the LCO statement and are specifically listed in Table 3. 3 .1-1 as
required Functions. The ITS LCO and Actions are worded to differentiate between "trip Functions" and "bypass removal Functions." These treatments are equivalent, with the ITS treatment being more explicit.
Applicability: In CTS, the automatic bypass removal channels are required to be Operable "When there is fuel in the reactor, more than one CONTROL ROD is capable of being withdrawn, and PCS boron concentration is less than REFUELING BORON. - CONCENTRATION, if the Zero Power Mode bypass is used"; in ITS, they are required to be Operable whenever the associated trips are required to be Operable, i.~., in Modes 1 and 2, and in M9des 3, 4, and 5 with more than one full-length control rod capable of being withdrawn and PCS boron concentration less than REFUELING BORON CONCENTRATION. The ITS, therefore, require the bypass removal channels to be Operable over a broader range of plant conditions than do th!;! .CTS, - making this change More Restrictive .
Palisades Nuclear Plant Page 5of11 05/30/99 Jd- -A
ATTACHMENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.3.1, RPS INSTRUMENTATION M.1 (continued) Bypass Allowance: In CTS, these bypass allowances are provided by footnote (b) of Table 3 .17 .1 which apply to the TM/LP trip, Low PCS Flow trip, and Low _SG Pressure trip functions.
That footnote allows these trips to be manually bypassed if the associated wide range nuclear instrument channels indicate below the specified power level. In ITS, the allowance to bypass the associated trip channels is provided in Table 3. 3 .1-1, footnote (c). These treatments are equivalent, with the ITS treatment being more explicit.
Automatic Bypass Removal: ...) I While not explicitly stated, CTS requires automatic removal of the ZPM bypasses by requiring the associated instrument channels to be Operable: footnote (a) of Table 3 .17 .1 requires both wide range channels to be Operable if the subject bypass is used. Since the wide range channels, which supply the input for the High Startup Rate trip, are subject to requirements for Channel Calibration (SR 4.17 .1, item 3), which ct
explicitly requires verification of the bypass removal function. In ITS, the requirement for automatic removal of the ZPM bypass is provided as part of Table 3. 3 .1-1 footnote (c). These treatments are equivalent, with the ITS treatment being more explicit.
Actions: C( In CTS, the Required Actions for the bypass removal (i.e., wide range nuclear instrument) channels are those provided for inoperable RPS trip units. If a _channel were bypassed (other than as allowed by the footnotes), it could not perform its rn. specified function and would be inoperable. Therefore, if a trip channel were ZPM ('(l bypassed, when a bypass removal (wide range NI) channel became inoperable, either the bypass would have to be removed (which places the plant in a condition outside the applicability), or the bypassed trip channel(s) would have to be declared inoperable. In CI ITS, those actions have been preserved as ITS actions D. l and D.2. These treatments I C(' are equivalent, with the ITS treatment being more explicit . I
Palisades Nuclear Plant Page 6of11 05/30/99 J,;;;. -.1
ATTACHMENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.3.1, RPS INSTRUMENTATION M.6 M. 7 The CTS Table 3.17.6 operating requirements for the Flux-6T Comparator are modified by a footnote which states that Specification 4.0.4 is not applicable. This allowance is not permitted in ITS 3.3.1. Deletion of the aHowance to enter a MODE or other specified condition of the applicability without having met the associated surveillance requirements is a reduction in flexibility and is considered more restrictive. Deletion of this allowance provides additional assurance that the required Function will be OPERABLE when it is required. This change is consistent with NUREG-1432 .. *
LESS RESTRICTIVE CHANGES - REMOVAL OF DETAILS TO LICENSEE CONTROLLED DOCUMENTS (LA)
LA.I CTS Table 4.17.1, Functional Unit 15, "Thermal Margin Monitor" requires that the constants be verified every 92 days. In the ITS, this surveillance is considered to be a part of SR 3.3.1.5, the CHANNEL FUNCTIONAL TEST for the Thermal Margin/Low Power (TM/LP) Function and other Functions. The constants of the Thermal Margin Monitor are specified in CTS Table 2.3.1 (ITS Table 3.3.1-2). These constants must be verified to ensure the trip setpoint of the TM/LP Function is set correctly. The details of the CHANNEL FUNCTIONAL TEST have been relocated to the Bases. The Bases for ITS SR 3.3.1.5 states that the thermal margin monitor constants must also be checked to be within tolerances. Removing the details of the CHANNEL FUNCTIONAL TEST from the CTS and placing them in the Bases of the ITS is acceptable since these details are not pertinent to the actual requirements. 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 with NUREG-1432 .
Palisades Nuclear Plant Page 7 of 9 01/20/98
ATTAC1'ENT 3
M.2 DISCUSSION OF CHANGES SPECIFICATION 3.3.1, RPS INSTRUMENTATION CTS 3 .17 .1. 6 requires specific actions be taken when the number of OPERABLE channels is less than specified in the "Minimum OPERABLE Channels" column of Table 3 .17 .1. The actions are to place the reactor in HOT SHUTDOWN within 12 hours; and place the reactor in a condition where the affected equipment is not required, within 48 hours. In the ITS, the actions when the "number of OPERABLE Channels" is less than the CTS minimum required is to enter LCO 3. 0. 3. The actions of ITS LCO 3.0.3 are to initiate action within 1 hour to place the plant, as applicable, in MODE 3 within 7 hours; in MODE 4 within 31 hours; and in MODE 5 within 37 hours. Since this change will require the plant to be in a lower MODE in a shorter time frame, this change is considered more restrictive. This change continues to assure that a plant shutdown can be achieved in a controlled manner without challenging plant systems. This change is an additional restriction on plant operations and is consistent with NUREG-1432.
M. 3 CTS 3 .17 .1. 6b requires the reactor to be in a condition where the affected equipment is not required within 48 hours. ITS 3.3.1, Condition G Actions require the same condition be achieved within 6 hours. Since the plant is required to be in a lower MODE in a shorter time frame, this change is considered more restrictive. With the
plant required to be in MODE 3 in 6 hours, de-energizing the clutch power supplies or borating to the REFUELING BORON CONCENTRATION can also be performed within the same time frame. These actions will not challenge plant systems, since the reactor is subcritical in MODE 3; This change is consistent with NUREG-1432.
M.4 Not Used M.5 CTS Table 3.17.1 Footnote (c) provides an allowance to change the setpoint of the low power bypass setpoint from 104 % RTP to 10-1 % RTP during LOW POWER PHYSICS TESTING. Since this allowance is not used, it is not included in the ITS. Since this change deletes an allowance and the plant will no longer be able to change this low power setpoint during the ITS PHYSICS TEST (see Discussion of Changes for ITS Section 1.0), this change is con5idered more restrictive. This change is consistent with NUREG-1432. M.6 New Conditions Band C and new Required Actions B.l and C.l have been added for I the Loss of Load and High Startup Rate Functions. These actions require restoration of 10 the inoperable channel(s) prior to re-entering operational conditions during which the I I Functions are required to be Operable. CTS Actions applicable to single channel I ---:. inoperability for other RPS trip Functions are not required for High Startup Rate and I Nl Loss of Load. In ITS, some Action must' be specified to avoid an LCO 3.0.3 entry I II) upon channel inoperability. These added Actions are unique to Palisades, since, to the I...._ best of our knowledge, Palisades* is the only CE plant with non-safety grade High I ct Startup Rate and Loss of Load trips which are not credited in the safety analyses. I~ Palisades Nuclear Plant Page 7of11 05/30/99
ATTAC1'ENT 3
M.6 (continued)
DISCUSSION OF CHANGES SPECIFICATION 3.3.1, RPS INSTRUMENTATION I I CTS allow continued operation for an unlimited peri1Jd of time with one High Startup I* Rate of Loss of Load Trip inoperable. Therefore, LCO 3.0.4 would not require their . I restoration prior to re-entry of their applicable conditions, even after an extended I shutdown. This open ended allowance is unnecessary and was not intended when the current CTS Actions were written. CTS contain other Actions for inoperable channels of supplementary equipment which allow continued operation for an unlimited period of time, but require restoration prior to the next startup. (See CTS Actions 3. 17. 6. 8, 9, 10, 11, & etc.) The ITS Actions provide this same limitation. The proposed Required Actions and Completion Times are considered to be acceptable because:
1. They are consistent with current practice, and do not impose unacceptable
' operational restriction,
2. They are more restrictive than the Actions required by CTS,
3.
4. With one channel inoperable, the trip Function is still capable of initiating a reactor trip with a two-out-of-three logic, and These trip Functions are not credited in the plant safety analysis. This is a Mote Restrictive change because the ITS require restoration of an inoperable trip channel where the CTS do not. M.7 Not used .
Palisades Nuclear Plant Page 8of11 05/30/99 l;;<-o

RPS 3 .3 INSTRUM£NTATION 3.3.l Re1ctor Protective Syste* {RPS) Instrument1tionb?OpeJ?'tinj'{An/i'1og)lj channels, and associated Zero Power Mode CZPMI
[3.17.lJ LCO 3.3.1 Four RPS trip units,~usochted instru11ent ~pus RAI removal ch1nnels for e1ch Function in Table 3.3.1-1 shill be 3.3.1-6 OPERABLE. APPLICABILITY: lfa60E$/1 1# 2.kAccordmg to Toble 3.3.1-1. >---@ RA! 3.3.1-1 ACTIONS
-------------------------------------NOTE-*-----------------------------------
Sep1r~te Condition entry is allowed for e1ch ~S t;pip c#""b19A"ss p!movj!~ Function. ' ~
-~-------------------------------------~--------------------------------------
REQUIRED ACTION COMPLETION TIME RA! [3.17.1.2JA. I 3.3.1-6
~ Place affected trip unit in trip. ]
( Insert 1) (continued) I RA! 3.3.1-1
(Palisades Nuclear Plant) jjtoG/Sfs I 3.3-1 !Afv l/04/ft7/9~
Jd--p
* ---------NOTE----------
Not applicable to ZPM Bypass Removal RPS Instrumentat ionl7?0peplt i nj' (An/!1 ogp l_Funct1on. j J.3.
------------------~----~
CONDITION REQUIRED ACTION COMPLETION TIME
------------NOTE-------------
LCO 3.0.4 is not ipplicable. [3.17 .1.3.a] 1 hour RAI 3.3.1-1 3.3.1-6
~ ~hours 1
I I I [3.17.1.2] £Ammed1atel!;j ) D. 0.1 1Ynou [3.17.1.3] QR
@?immed1ately )
0.2~ !Yhou . @ (continued) RA!
< Insert 1) @ I 3.3.1-1 (Palisades Nuclear Plant) /~- Q_., @oG;sts I 3.3-2 0 !Afv 1/ 04/ft?/9;!
*4)
(--Vl RPS Instrumentat i orE<>Pera't-:l!lg Mnahl~)} 1 3.3.r
SURVEILLANCE REQUIREMENTS continued SURVEILLANCE FREQUENCY
[!4-.11-1] SR A.t-lD ve:~1F-Y ~c. ~a:..-.t1..1... N\A~\\..\ MON' TOP! C.0>-1 ~n~ rc;. (f) RA:t
~;~.1-12 Once within 7 days prior to @
each reactor startup SR~ P rform a CHAN~Lr:~NCTION~L TEST on eat Once wit.hin au matic bypas ~oval fu~. 92 days P{ior to each l7~tor art up
~ls@ months SR . 3.~ Veri~PS RES~NSE TIME ~within li~s.) [l month~
on a TAGGERED TEST B
--
IJOIE- - - - - - - - -
t-.lE:tJ'T"IZOU ~TE::"'-Toe.~ Ari!:' ~t..ve->e-t) Ftof'ol\ ~1!: C,l-\~tJ~t:-1... CAU6~Ai\0~*
CEOG STS 3.3-5
/~ -)\_
Rev 1, 04/07/95
SECTION 3 . 3
INSERT I Table 3.3.1-1 (page 1 of 2)
Reactor Protective System Instrumentation APPLICABLE 'SURVEILLANCE K'A"L . g.~.1-1 3.~.\-~ FUNCTION MODES REQUIREMENTS ALLOWABLE VALUE
1. Variable High Power Trip SR 3 .3 .1.1 ~ 15% RTP above SR 3. 3 .1. 2 current THERMAL SR 3. 3 .1. 3 POWER with a SR 3. 3 .1.4 minimum of ~ 30%
SR 3 .3 .1. 5 . RTP and a maximum SR 3. 3 .1. 6 of ~ 106.5% RTP SR 3 .3 .1.8
~. Hi~h Startup Rate Tri p<bl 1,2 SR 3. 3 .1.1 NA SR 3. 3 .1. 7 SR 3 .3 .1.8
3. Low Primary Coolant System Fl ow Tri p<c> SR 3. 3 .1.1 ~ 95%
SR 3.3.1.5 SR 3.3.1.8
4. Low Steam Generator A Level Trip
5. Low Steam Generator B Level Trip 1,2,3<1 l,4(al,5Cal l,2,3< >,4(al,5(a) 1 SR 3 .3 .1.1 SR 3 .3 .1.5 SR 3.3.1.8 SR 3.3.1.1
~ 25.9% narrow range ~ 25.9% narrow SR 3. 3 .1.5 range SR 3. 3 .1.8
6. Low Steam Generator A Pressure Trip(cl l,2,3< 1 >,4(al,5(al SR 3 .3 .1.1 ~ 500 psia SR 3 .3 .1.5 SR 3.3.1.8

7. Low Steam Generator B Pressure Trip(cl l,2,3C 1 >,4(*l,5(*l SR 3 .3 .1.1 ~ 500 psia SR 3. 3 .1.5 SR 3.3.1.8 8, High Pressurizer Pressure Trip 1,2,3 11 >,4Cal ,5Cal SR 3. 3 .1.1 ~ 2255 psia SR 3.3.1.5 SR 3 .3 .1.8 (a) With more than one full-length control rod capable of being withdrawn and PCS boron concentration less than REFUELING BORON CONCENTRATION.
(b) Trip may be bypassed when Wide Range Power is< lE-4% RTP or when THERMAL POWER is
> 13% RTP.
(c) Trips may be bypassed when Wide Range Power is < lE-4% RTP. Bypass shall be automatically removed when Wide Range Power is ~ lE-4% RTP.
/~-~
3.3-6
(a) With more than one full-length control rod capable of being withdrawn and PCS boron concentration less than REFUELING BORON CONCENTRATION. (c) Trips may be bypassed when Wide Range Power is< lE-4% RTP. Bypass shall be. automatically removed when Wide Range Power is ~ lE-4% RTP. (d) When THERMAL POWER is ~ 17% RTP. ,
/;J-.f 3.3-7
l SECTION 3.3
INSERT 1 Table 3.3.1-2 (page 1 of 1)
Thermal Margin/Low Pressure Trip Function Allowable Value The Allowable Value for the Thermal Margin/Low Pressure Trip, Ptrip* is the higher of two values, Pmin and Pvar* both in psia: Pmin = 1750 pvar = 2012(QA) (QR1) + 17 .O(T1n) - 9493 Where: QA = - 0.720(ASI) + 1.028; when - 0.628 ~ ASI < - 0.100 QA = - 0.333(ASI) + 1.067; when - 0.100 ~ ASI < + 0.200 QA = + 0.375(ASI) + 0.925; when + 0.200 ~ ASI ~ + 0.565 ASI = Measured ASI when Q ~ 0.0625 . ASI = 0.0 when Q < 0.0625
QR 1 = 0.412(Q) + 0.588; QR1 = Q; Q = THERMAL POWER/RATED THERMAL POWER when Q ~ 1.0 when Q > 1.0 Tin = Maximum primary coolant inlet temperature, in °F ASI, Tin, and Qare the existing values as measured by the associated instrument channel .
3.3-8
Change ATTACHMENT 6 JUSTIFICATION FOR DEVIATIONS SPECIFICATION 3.3.1, RPS INSTRUMENTATION Discussion
26. TSTF-179 is incorporated to revise NUREG Required Action C.2 text from "70% of the maximum allowed THERMAL POWER level" to "70% RTP." Replacing the undefined phrase "maximum allowed THERMAL POWER level" with the defined phrase "RTP" eliminates possible misinterpretation of the Action and is consistent with the conventions in the NUREGs. This change is also consistent with the CEOG Digital Actions.

27. TSTF-189 is incorporated to remove the.discussion of the allowable value and it's uncertainty from the Containment Pressure - High Bases. References to instrument uncertainty in the Bases are inconsistent with the ITS conventions and not given in other Specifications. The change is consistent with the CEOG Digital Bases.

28. TSTF-80, Rev. 1, is not incorporated, rather the current licensing basis is retained as an added Action for the Loss of Load trip function.
~:
Palisades Nuclear Plant Page 4 of 4
/~-¥'
01/20/98
ATTACH1\1ENT 6 JUSTIFICATION FOR DEVIATIONS SPECIFICATION 3.3.1, RPS INSTRUMENTATION Change Discussion
25. TSTF-178 is incorporated to omit "trip or bypass removal" from the ACTIONS Note.
The RPS Functions listed in Table 3. 3 .1-1 include trip and bypass removal features where appropriate. Referring to the trip or bypass removal features as separate Functions is incorrect and confusing. Removing the words "trip or bypass removal" satisfies the intent of the Note and eliminates the error. This change is also consistent with the CEOG Digital LCO.
26. TSTF-179 is incorporated to revise NUREG Required Action C.2 text from "70% of the maximum allowed THERMAL POWER level" to "70% RTP." Replacing the undefined phrase "maximum allowed THERMAL POWER level" with the defined phrase "RTP" eliminates possible misinterpretation of the Action and is consistent with the conventions in the NUREGs. This change is also consistent with the CEOG Digital Actions.

27. TSTF-189 is incorporated to remove the discussion of the allowable value and it's uncertainty from the Containment Pressure - High Bases. References to instrument uncertainty in the Bases are inconsistent with the ITS conventions and not given in other Specifications. The change is consistent with the CEOG Digital Bases.

28. TSTF-80, Rev. 1, is not incorporated, rather the current licensing basis is retained as an added Action for the Loss of Load trip function.

29. The following alterations were made to STS LCO 3.3.1 to implement the CTS Operability, Action, and Surveillance requirements for the instrument channels which provide automatic removal of the Zero Power Mode (ZPM) bypass of certain RPS trip functions.
The LCO wording and Table 3.3.1-1 have been changed to explicitly call out the ZPM bypass removal channels as a required Function. Condition "D" and the associated Actions have been revised to reflect the CTS requirements for an inoperable ZPM bypass removal channel. SR 3. 3 .1. 8 has been revised to omit the reference to bypass removal channels ('() because Table 3.3.1-1 explicitly requires that SR for the ZPM bypass removal Function. The ZPM bypass removal channels are ilie only bypass removal channels required to be Q: Operable in the CTS and the only ones assumed to function in the accident analyses. C<'
This is a plant specific change.
Palisades Nuclear Plant Page 4 of 4 05/30/99
/d.-w
CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO JANUARY 6, 1999 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.3, INSTRUMENTATION NRC REQUEST: 3.3.1-7 CTS 3.17.2.5 DOC M.1 Comment: The last paragraph of this DOC discusses an M-DOC change that is unrelated to the first three paragraphs of the discussion. Revise the submittai. Revise the justification frorn "consistent with current practice" to a more specific statement to explain the more restrictive requirements that are being adopted in ITS. Consumers Energy Response: DOC M.1 was revised in response to RAls 3.3.1-1 and 3.3.1-6. The last paragraph was deleted, and DOC M.1 no longer addresses the subject added ITS Actions. Doc M.6 has also been revised to address the added ITS Actions for inoperable High Startup Rate and Loss of Load trip Function channels. See the response to RAls 3.3.1-1 and 3.3.1-6. Affected Submittal Pages: See the response to RAls 3.3.1-1 and 3.3.1-6 .
13
CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO JANUARY 6, 1999 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.3, INSTRUMENTATION NRC REQUEST: 3.3.1-8 CTS 4.18.2.1.b CTS 4.18.2.1 b requires a comparison of excore channel measured Axial Offset (AO) to the total core AO measured by the incores. If the excore and incore measurements are greater than .02 apart, the excores are to be adjusted. STS SR 3.3.1.3 requires that the excore detectors be "calibrated" using the incore detectors. Comment: Revise the ITS to replace ITS SR 3.3.1.4 requirement to "Compare" with "Calibrate." Consumers Energy Response: ITS SR 3.3.1.4 has been revised to utilize the STS wording. Conforming changes have been made to the ITS bases. In addition, the wording of ITS SR 3.3.1.5 has been revised to explicitly call out the verification of the Thermal Margin Monitor constants, as is required by CTS Table 4.17 .1 Item 15. DOC LA.1 has been revised to address these changes. Affected Submittal Pages:
Att 1, Att 2, Att 3, Att 3, Att 3, ITS 3;3.1, page 3.3.1-4 ITS B 3.3.1, page B 3.3.1-28 CTS, page 4-76 CTS, page 4-83 DOC 3.3.1, page 7 of 9 (former submittal)
Att 3, DOC 3.3.1, page 8 of 10 Att 5, NUREG page 3.3-4 Att 5, NUREG page 3.3-4 insert Att 5, NUREG page 3.3-5 Att 6, JFD 3.3.1, page 3 of 4 (former submittal) Att 6, JFD 3.3.1, page 3 of 4
14
RPS Instrumentation 3.3.1
SURVEILLANCE REQUIREMENTS
-------------------------------------NOTE-------------------------------------
Refer to Table 3.3.1-1 to .determine which SR shall be performed for each Function.
SURVEILLANCE FREQUENCY SR 3.3.1.l Perform a CHANNEL CHECK. 12 hours SR 3.3.1.2 Verify control room temperature is ~ 90°F. 12 hours SR 3.3.1.3 -------------------NOTE--------------------
Not required to be performed unt i1 12 hours after THERMAL POWER is ~ 15% RTP.
'E.'t.C.O~
Perform calibration (hea. balance only) a*nd 24 hours
adjust the power range and ~T power channels to agree with calorimetric calculation if the absolute difference is
~
fa'AI:. 3."3.\-4 I SR J.3.1.4 31 days SR 3.3.1.5 92 days SR 3.3.1.6 r e h power range excore channels 92 days a test signal . I
, J I // Palisades Nuclear Plant 3.3.1-4 Amendment No. 01/20/98
RPS Instrumentation B 3.3.1
BASES SURVEILLANCE SR 3.3.l.3 (continued)
REQUIREMENTS The Frequency of 24 hours is based on plant operating experience and takes into account indications and alanns located in the control room to detect deviations in channel outputs. The Frequency is modified by a Note indicating this Surveillance must be perfonned within 12 hours after THERMAL POWER is ~ 15% RTP. The secondary calorimetric is inaccurate at lower power levels. The 12 hours allows time requirements for plant stabilization, data taking, and instrument calibration. SR 3.3.1.4
Fr u c indicates the Surveillance is~
TO 1-\A\JE: n 12 ours after THERMAL POWER is ~ 25% RTP.
"W~RME.D Uncertainties in the excore and incore measurement process '-,;.__ _ _..,make it impractical to calibrate when THERMAL POWER is KA"!... < 25% RTP. The 12 hours allows time for plant stabilization, data taking, and instrument calibration. 3.3.\-8 The 31 day Frequency is adequate, based on operating experience of the excore linear amplifiers and the slow burnup of the detectors. The excore readings are a strong function of the power produced in the peripheral fuel bundles and do not represent an integrated reading across the core. Slow changes in neutron flux during the fuel cycle can also be detected at this Frequency.
ASI. \S lJliLl~b AS At-t \NPt.rr TO 1~=- Tf'/\/LP ,-e,p F\Jt-lC:\\0"1 v..t~C:i!c rr IS USE:-"D To Ef..l~U~ ,...~U:\ i°PI~ t-1\"E:A.S\.l~'t> )i..~\~'- ~e. '?E:o1=11..g At!:.E: &v~b&'b B ~ "'f"\.\~ -'~'~'- ~=tc:. ~oi;-1\-&.S l/~E"t:> 1tJ iwc::.
'D~E:\.OPM'E:"t-ll O'F- ,...+il'E:- Tit-ll.'ET" L\M\i'A.TIOJ.i OF-LCO "3,'-f, \.
Palisades Nuclear Plant I -b B 3.3.1-28 01/20/98
3.3.1
4.17 INSTRUMENTATION SYSTEMS TESTS ~
Table[¥1~(~.1-1) Instrymentatjon Survei11ance Regyjrements'for Reactor protective System CSR 3.3.1.ll CS~H~N~~i..5J CSR 3.3.1.BJ CHANNEL FUNCTIONAL CHANNEL Fynctional Unit CHECK TEST CALIBRATION
1. ~nuef Trrt / / / /f!A / / /{a)/ / / NAJ-\sEE 3.3.2)
[l] 2. Variable High Power 12 hours days (b, c, & d) [2] 3. High Start Up Rate 12 hours (a) 18 months<el [9] 4. Thermal Margin/ 12 hours 18 months Low Pressure [8] 5. High Pressurizer 12 hours days 18 months Pressure RAI [3] 6. Low PCS Flow 12 hours days 18 months 3.3.1-6 [10]7 Loss of Load NA (a) 18 months
[4] 8. Low 11 A11 SG Level 12 hours days 18 months
[5] 9. Low "B" SG Level 12 hours days 18 months [6] 10. Low "A" SG Pressure 12 hours days 18 months [7] 11. Low "B" SG Pressure 12 hours days 18 months [11] 12. High Containment Pressure NA days 18 months 13. EE 11~ 14. CSR 3.3.1.5] 15. Thermal Margin Mani tori Verify constants each 92 days. . A.12 13.~~~ 8 CSR 3.3.1.2] 16. erm Ma in oni r: Verify Control Room Temperature ~ 90°F each 12 hours. CSR 3.3.1.71 (a) Once within 7 days prior to each reactor startup. <Add SR 3.3.1.3 Note& CSR 3.3.1.3J (b) Calibrate with Heat Balance each 24 hours, when> 15% RATED POWER.~ CSR 3.3.1.6] (c) Calibrate Excores channels with test signal each~ CSR 3.3.1.81 (d) CHANNEL CALIBRAT~ON each 18 months. CSR 3.3.1.8] (e) Include verification of au.tomatic Zero Power Mode Bypass removal. Apply to ITS Function gr\~ '-"
I I Amendment No. ta, 1:3e, -%-36, TSe,
, I -/"'---- 4-76 ~. 1-64,-rrl, Poge 4 of 7 186
g_~. I r~_
*~~
4.18 4.18. l STRUM NTATION T STS 4.18.1.l The detection system shall be demonstrated operable:
a. -By erfonnance of a Channel C eek prior to its use fol a eration and at least once per 7 days during power r, quired for the functions isted in Section 3.11.1.
At least once per refueli g by performance of a Ch which exempts the neutro detectors but includes components. ( he incore alann system is ~emonstrated operable thr ugh use of the datalogger ~or alann. The~ alann is
demonstratea opera61 e- ~ - per refueling by per'(" ance of a Channel Check.
4.18.2 Excore Monitoring SyStem
a. A target and excore monitoring. llowable power levtl' shall be determine using excore and incor detector readings steady-state n r equilibrium ~ondition * ** * -*--*-----
[-sR 3.3. I. 4-J b. red to thel~A II .*~ eren is ~ [S'R 3.3,1,t./ ~o+e '0 p.:!~..:r:.!:::~:1..llCIL.l...J&...1-.:.~~~..-.......- " ".'...__:;::.fAdi~~:i i.._ JJofpf 1_
* (...See o..lso 3 ,'Z.) ....~;.;..;..;;L;..;;;.;;.;;,~
c. The excore asured Quadrant Pow Tilt shall be compa~*d to the incore meas red Quadrant Power T lt. If the differenc is greate than 21, ta excore monitoring* ystem shall be recali rat.ed.
RA1. 3$.f-B Amendment No. 371 &8, 118, 162 October 26, 1994 4-83 J+-d
M.6 SPECIFICATION 3.3.1, RPS INSTRUMENTATION ATTACHMENT 3 DISCUSSION OF CHANGES CTS Table 3 .17 .1 permits operational bypasses to be used for certain Functions. ITS Table 3. 3 .1-1 requires that the operational bypasses for these Functions be automatically removed when the Function variable reaches the value at which use of the bypass was permitted to be placed in use. CTS Table 4.17 .1 is also revised to include

bypass removal channels in the CHANNEL CALIBRATION. This change imposes additional requirements which did not previously exist, and is considered more restrictive. This change *will provide additional assurance that RPS Trip Functions are not bypassed when the Function is required to be OPERABLE. This change is consistent with NUREG-1432.
M.7 The CTS Table 3.17.6 operating requirements for.the Flux-AT Comparator are modified by a footnote which states that Specification 4.0.4 is not applicable. This allowance is not permitted in ITS 3.3.1. Deletion of the allowance to enter a MODE or other specified condition of the applicability without having met the associated surveillance requirements is a reduction in flexibility and is considered more restrictive. Deletion of this allowance provides additional assurance that the required Function will be OPERABLE when it is required. This change is consistent with NUREG-1432 .
LESS RESTRICTIVE CHANGES - REMOVAL OF DETAILS TO LICENSEE CONTROLLED DOCUMENTS (LA)
~J!I'-. rtavt~I!.">' LA-I f?A-1 '!.l.1-8 LA.1 S Table 4.17.1, Functional Unit 15, "T erm argm omtor require.
that constants be verified every 92 days. In the ITS, this surveill 1s consider be a part of SR 3.3.1.5, the CHANNEL FUNCTIO L TEST for the Therm gin/Low Power (TM/LP) Function and er Functions. The constants of the rmal Margin Monitor are spec* a in CTS Table 2.3. l (ITS Table 3;3.1-2). Thes onstants must be ver* Cl to ensure the trip setpoint of the TM/LP Function
et correc . The details of the CHANNEL FUNCTIONAL TEST have been relo to the Bases. The Bases for ITS SR 3.3.1.5 states that the thermal gin itor constants must also be checked to be within toleranc . Removing the FUNCTIONAL TEST f the CTS and placing the . the Bases of the ITS .
acceptable since the aetails are not pertinent to the actual uirements. Placing these ls in the Bases provides adequate assurance tha ey will be maintain mce the Bases are controlled by the Bases Control Progra pro Ed in ITS Chapter 5.0. This change is consistent with NUREG-1432 .
Palisades Nuclear Plant Pag~ 7of9 01/20/98
ATTACHMENT 3
DISCUSSION OF CHANGES SPECIFICATION 3.3.1, RPS INSTRUMENTATION LESS RESTRICTIVE CHANGES - REMOVAL OF DETAILS TO LICENSEE CONTROLLED DOCUMENTS (LA)
LA.1
CTS SR 4.18.2.lb requires that "Individual excore channel measured AO shall I be compared to the total core AO measured by the incores. If the difference is I greater than 0.02, the excore monitoring system shall be recalibrated.. This SR I wording has been replaced with the STS wording "Calibrate the power range I excore channels using the incore detectors." The adjustment details contained in I *[J.)
CTS 4.18.2. lb have been moved to the procedure. These details merely I *1 provide the "the necessary range and accuracy to known values of the parameter
1-that the cha~el monitors" referred to in the Channel Functional Test definition, 1 {'{)
and are not typically specified in ITS or STS SRs. These details are not I NI necessary in the SR and have been relocated to the Bases. In the ITS, the I pertinent requirements are included in SR 3.3.1.4. The details related to how the power* range detectors are calibrated are more appropriately included in the : ct Bases. Changes to the Bases will' be made in accordance with the Bases Control I ct Program as discussed in ITS Chapter 5.0, Administrative Controls. *This change I maintains consistency with NUREG-1432. I -
. LESS RESTRICTIVE CHANGES (L)
L.1 CTS 3.17.1.3 requires entry when two RPS trip units or associated instrument
.. channels are inoperable in one or more Functions. In the ITS, these conditions are addressed in ITS 3.3.1, Condition B. The difference is that the Condition B Required Actions include a Note which excludes the applicability of LCO 3.0.4.
This provision was added to allow MODE changes even though two channels are inoperable, with one channel in trip. In this condition, with one channel in trip the protection system is in a one-out-of-two logic. Since the probability of a
common cause failure affecting both of the OPERABLE channels during the 7 day Completion Time to restore one trip unit to OPERABLE status is remote, and the low probability of occurrence of an event during this interval, this change is considered acceptable. This change is consistent with NUREG-1432.
\ .
Palisades Nuclear Plant Page 9of11 05/30/99
;~-f-
~
RPS Instrumentat10~.,..~~-e~r"""a"bajl~g~t41~n=a-Pe,g---p
3:T.
SURVEILLANCE REQUIREMENTS
-------------------------------------NOTE-------------------------------------
(4..i 11
~~!~r~~~ Table 3.3.1-1 to determine which SR shall be performed for each I Ull\o" I \Ill* ---------------~--------------------------------------------------------------
SURVEILLANCE FREQUENCY [Te..11.1] SR 3.3.1.1 12 hours SR Perform calibration heat balance only) and 24 hours adjust the excore power range and AT power ~AI. channels to agree with calorimetric '3."3. I- L.I calculation if the absolute difference is
~ [W.s~. © SR 3.3.1.~. -------------------NOTE--------------------
Not required to be performed until 12 hours after THERMAL POWER is ~ ~RTP.
~
Calibrate the power range excQre channels 31 days [4-. IB .'"Z.. I. b] using the i ncore detectors. (continued)
CEOG STS Rev 1, 04/07/95
,,..-~-
SECTION 3.3 INSERT 1 SURVEILLANCE FREQUENCY SR 3.3.1.2 Verify control room temperature is ~ 90'F. 12 hours

3.3-4 I ~-L
(3+, RPS InstrumentationE<>P'era'N.rig MnaN:i[D "' 3.3.r
SURVEILLANCE REQUIREMENTS continued SURVEILLANCE FREQUENCY
[T 4.11 -1] SR A."1D ve:~1t=-Y n.u:. '"'ci:l-\t..1... N'\A.~\~ Mo1-.11 TOI'! C.ON ~"IA.,,. TS ( -';/ 3/
*~
92 days I(j) . r-rii-.11-1 1 SR 3.3.1.- Perform a CHANNEL FUNCTIONAL TEST of~ Once within Lfoo+Mte (<>.)J @*qh S-tw-tvP)~ Rate cqj§t@el §{laMelJ and' Loss 7 days prior to each reactor (2) of Load ifTkls a ~*
CJ v. ~d ; 0.,,. ~ startup SR~ P rform a CHAN~Lr:~NCTION L TEST on eac Once wit.hin au matic bypas ~oval fu tion. 92 days P(ior to each l'.'~tor art up SR 3 .3 .1.8 ~ls@ months SR 3.~ Veri~S RES~ TIME ~thin li~

CEOG STS 3.3-S J4-J Rev 1, 04/07/95
\
Change Discussion ATTACHMENT 6 JUSTIFICATION FOR DEVIATIONS SPECIFICATION 3.3.1, RPS INSTRUMENTATION
17. Figures 3.3.1-1, 3.3.1-2, and 3.3.1-3 do not apply to Palisades. Table 3.3.1-2 has been included which provides the required relationship and allowable values at the Thermal Margin/Low Pressure (TM/LP) Function.

18. Table .3.3.1-1 Footnote (c) does not apply, subsequent Footnotes have been renumbered, where applicable.

19. The specific wording which discusses other plants is being deleted the bases are specific to Palisades and will not contain, where possible, only Palisades specific information.

20. The bracketed Reviewer's' Note has been deleted since it is not meant to be maintain:ed in the plant specific ITS ..

21. ISTS Figure B 33.1-1 has been deleted since similar diagrams are already included* in*
FSAR Figure 7-1 and 7-2 .
22.
23. same basis.
24. t to verify constants associate with the thermal ma in monitors is added, t with details relocated from S Table 4.17 .1, Fu tional Unit 15 OC LA.I).

25. TSTF-178 is incorporated to omit "trip or bypass removal" from the ACTIONS Note.
The RPS Functions listed in Table 3.3.1-1 include trip and bypass removal features where appropriate. Referring to the trip or bypass removal features as separate Functions is incorrect and confusing. Removing the words "trip or bypass removal" satisfies the intent of the Note and eliminates the error. This change is also consistent with the CEOG Digital LCO .
Palisades Nuclear Plant Page 3 of 4 1'-f-t 01/20/98
ATTAC1'ENT 6 JUSTIFICATION FOR DEVIATIONS SPECIFICATION 3.3.1, RPS INSTRUMENTATION Change Discussion
14. ISTS SR 3.3.1.7 has been deleted since the 18 month CHANNEL CALIBRATION surveillance (ITS SR 3. 3 .1. 8) is considered adequate to ensure the bypass removal channels are functioning properly. This change is consistent with the plant current licensing basis. Subsequent surveillances have been renumbered, where applicable.

15. ISTS SR 3.3.1.9, the RPS RESPONSE TIME test is not included in the ITS. This test is not required by the current licensing basis since the conclusions of NUREG-082, "Integrated Plant Safety Assessment Systematic Program Palisades Plant" - Final Report USNRC 10/82, concluded that backfitting this testing into the current program was not required.

16.
ISTS 3.4.17, "RCS Loops - Test'Exceptions" is not being proposed therefore reference
. to this specification is deleted.
17. Figures 3.3.1-1, 3.3.1~2, and 3.3.1-3 do not apply to Palisades. Table 3.3.1-2 has been included which provides the required relationship and allowable values at the
18.
Thermal Margin/Low Pressure (TM/LP) Function.
Table 3.3.1-1 Footnote (c) does not apply, subsequent Footnotes have been renumbered, where applicable.
19. The specific wording which discusses other plants is deleted; the bases are specific.to Palisades and will contain, where possible, only Palisades specific information.

20. The bracketed Reviewer's Note has been deleted since it is not meant to be maintained in the plant specific ITS. *

21. .ISTS *Figure B 3. 3 .1-1 has been deleted. since similar diagrams are already included in FSAR Figure 7-1 and 7-2 ..

22. Note 2 of ISTS SR 3.3.1.2 has been deleted since the PHYSICS TESTS are perfo1,"llled ct) below 2 % RTP and therefore, the allowance in Note 1 suffices. 1

23. Not Used I~
I~
24. Requirement to verify constants associate~ with the thermal margin monitors is added, 1-consistent with CTS Table 4.17.1, Item 15. 1 tt
Palisades Nuclear Plant .Page 3 of 4 05/30/99 I~
14~
CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO JANUARY 6, 1999 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.3, INSTRUMENTATION
.. NRC REQUEST:
3.3.1-9 ITS Required Actions E, F G & H CTS 3.17.1, DOC M.1 In the ITS Note* applies to 3.17.1.2.a and to 3.17.1.3.c. This note states that Actions are not required for inoperable High Startup Rate or Loss of Load instrument channels .. No othei CTS actions apply to these functions. DOC M.1 justifies adding new conditions E, F, G, and H for Loss of Load and High Startup Rate instrument channels based on these requirements being consistent with current practice though not required for CTS for these functions. Comment: Thus, the proposed ITS requirements represent a generic change to the ISTS and a change to the CTS. Revise the ITS and associated DOCs to adopt the ISTS Conditions or provide additional justification giving a design basis justification or operational hardship position for not adopting the ISTS. Consumers Energy Response: These added Actions are unique to Palisades, since, to the best of our knowledge, Palisades is the only CE plant with non-safety grade High Startup Rate an.d Loss of Load trips which are not credited in the safety analyses. They perpetuate Actions in CTS which were approved by Amendment 162.
DOC M.1 was revised in response to HAis 3.3.1-1 and 3.3.1-6. The last paragraph was deleted, and DOC M.1 no longe.r addresses the subject added ITS Actions. Doc M.6 has also been revised to address the added ITS Actions for inoperable High Startup Rate and Loss of Load trip Function channels.
See also the response to RAls 3.3.1-1 and 3.3.1-6. Affected Submittal Pages: See the response to RAls 3.3.1-1 and 3.3.1-6 .
15
CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO JANUARY 6, 1999 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.3, INSTRUMENTATION NRC REQUEST: . 3.3.1-10 Resolved during 10/27/98 meeting; Consumers Energy not required to respond. NRC REQUEST: 3.3.1-11 ITS SR 3.3.1.4, Note CTS 4.18.2.1.b DOC L.5 L.5 justifies the addition of the note to allow 12 hours for performing the excore calibration using the incore instruments after thermal power exceeds 25%. Comment: Provide additional discussion in DOC L.4 to justify using 25% RTP for starting
the 12 hour window for completing the required surveillance.
Consumers Energy Response: DOC L.5 has been revised as requested. Affected Submittal Pages: Att 3, DOC 3.3.1, page 9 of 9 (former submittal) Att 3, DOC 3.3.1, page 10 of 10
16

L.3 ATTACHMENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.3.1, RPS INSTRUMENTATION CTS Table 4.17.1, Footnote (c), requires that the excore channels be calibrated with a test signal every 31 days. ITS SR 3.3.1.6 requires the surveillance to be performed every 92 days. The proposed change revises the calibration Frequency for the excore power range channels from 31 days to 92 days. This test leaves the channel adjusted to account for instrument drift between successive calibrations to ensure that the channel remains OPERABLE between

calibrations. Other surveillances are performed on the excore power range channels more frequently to account for overall gain and to ensure the upper and lower subchannel amplifiers are calibrated correctly to correspond to the in core detectors. These surveillances are considered sufficient to ensure the excore power range channels are functioning properly. This change is consistent with NUREG-1432. . .
L.4 CTS Table 4.17.1 Footnote (b) requires calibration of the Variable High Power Function with heat balance when power is> 15% RTP. ITS 3.3.1 will add a Note to SR 3.3.1.3 (heat balance) which states that the SR is not required to be performed until 12 hours after THERMAL POWER is > 15% RTP. The
allowance to delay performance of the SR for 12 hours after power is
> 15 % RTP provides time for the plant to achieve stable operating conditions to calibrate the instruments at a power at which the heat balance is accurate. This will provide more accurate results, and thereby providing assurance that the RPS Functions will actuate at the required setpoints. The 12 hours interval is within the 24 hour delay allowed for a missed SR. This change is consistent with NUREG-1432.
See. E.\JlSit.'>
Palisades Nuclear Plant Page 9 of 9 I & -a._..,.
01/20/98
I* ATTACHMENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.3.1, RPS INSTRUMENTATION L.5 CTS 4.18.2. lb requires calibration of the excore monitoring system with the incore monitoring system at least every 31 days of power operation. ITS 3.3.1.4 requires the sa.me surveillance, but is modified by a Note which states that the SR is not required to be performed until 12 hours after THERMAL POWER is~ 25% RTP. Since the CTS "power operations" is defined as "greater than 2 %, " this change is considered less restrictive. I, This 1 SR need only be performed above 25% RTP since the calibration performed adjusts the Axial Shape Index (indicated by the excore power range) to agree with Axial Offset (indicated by the incore monitoring system). The ITS Axial Shape Index LCO, 3.2.4, has an Applicability of "MODE 1 with THERMAL POWER > 25 % ". Below 25 % RTP, Axial Shape Index (ASI) is not limited by Technical Specifications because there is sufficient thermal margin to allow operation with potential axial power imbalances. . Since the ASI LCO is not applicable below 25 % RTP, the excore power range channels need not be re-adjusted to prec_isely indicate ASL Selecting 25 % RTP as the starting point for the 12 hour window provided in SR 3.3.1.4 is consistent with other ITS requirements which utilize the incore monitoring system. -- The allowance to delay the surveillance for 12 h.ours after power is ~ 25 % RTP provides time for the plant to stabilize at a power level- sufficiently high enough to accurately calibrate the instruments. The 12 hour interval is acceptable because the need to calibrate the excores with the incores is dependant on potential changes in the core due to power generation. After outages, other than refueling outages, while the 31 day SR interval may have expired, the former calibration would still be valid; following a refueling outage, post-refueling power assention physics testing, completed
prior to exceeding 25 % RTP, would uncover any unacceptable disparity between AO (measured by the incores) and ASI (measured by the excores). The 12 hours in~erval is within the 24 hour delay allowed for a missed SR. This change is consistent with NUREG-1432 .

Palisades Nuclear Plant J
Page 11of11 05/30/99
CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO JANUARY 6, 1999 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.3, INSTRUMENTATION NRC REQUEST: 3.3.1-12 ITS SR 3.3.1.6 CTS T 4.17-1, footnote (c) JFD 5 The ITS proposes to modify the STS by eliminating the use of the defined term "CHANNEL CALIBRATION" and addition of the surveillance detail "with a test signal" for the power range excore channel test requirements. Comment: Revise ITS SR 3.3.1.6 to use the defined term CHANNEL CALIBRATION. Surveillance requirement details are moved to the Bases in the ITS program. Provide an LA DOC for eliminating CTS details related to using "test signals" to perform channel calibrations. Consumers Energy Response: ITS SR 3.3.1.6 is not intended to emulate any STS requirement. As shown in the markup of CTS page 4-: 76 (Table 4.17 .1 Footnote c), ITS SR 3.3.1.6 is intended to retain that CTS requirement. The CTS wording is: "Calibrate Excores channels with test signal each 31 days." The use of this test signal only verifies the response of the power range instrument drawer to specific values of input. *it is not intended to, and does not, constitute a Channel Calibration as defined in either CTS or ITS .
To avoid potential confusion between this requirement and the defined Channe.I Calibration, which is required by CTS Table 4.17.1 Footnoted and ITS SR 3.3.1.8, ITS SR 3.3.1.6 has been reworded to require: "Perform a calibration check of the power range excore channels with *a test signal." Conforming changes have been made to the Bases.
Affected Submittal Pages: Att 1, ITS 3.3.1, page 3.3.1-4 Att 2, ITS B 3.3.1, page B 3.3.1-29 Att 2, ITS B 3.3.3, page B 3.3.1-30 Att 5, NUREG, page 3.3-5
17
-* RPS Instrumentation 3.3.1
SURVEILLANCE REQUIREMENTS
-------------------------------------NOTE-------------------------------------
Refer to Table 3.3.1-1 to determine which SR shall be performed for each Function. SURVEILLANCE FREQUENCY SR 3.3.1.1 Perform a CHANNEL CHECK. 12 hours SR 3.3.1.2 Verify control room temperature is ~ 90°F. 12 hours SR 3.3.1.3 -------------------NOTE-------------------- Not required to be performed until 12 hours after THERMAL POWER is ~ 15% RTP. Perform calibration (heat balance only) and 24 hours adjust the power range and ~T power channels to agree with calorimetric calculation if the absolute difference is
~ 2%.
SR 3.3.1.4 -------------------NOTE-------------------- Not required to be performed until 12 hours after THERMAL POWER is ~ 25% RTP. Compare each power range excore channel ASI 31 days to total core AXIAL OFFSET measured by the incore detectors, and adjust the excore channel if the difference is > 0.02. SR 3.3.1.5 Perform a CHANNEL FUNCTIONAL TEST. 92 days
'\ 'irA-\ 1.1. I* l Z } ~~,_&"" It cM.11Sc ..'"*lllA c1i1ac..c cl=
SR 3.3.~ ~ eali~Pa!e the power range excore channels 92 days with a test signal .
Palisades Nuclear Plant 3.3.1-4 17-o..-
Amendment No. 01/20/98
RPS Instrumentation B 3.3.1
BASES SURVEILLANCE SR 3.3.1.5 REQUIREMENTS (continued), A CHANNEL FUNCTIONAL TEST is performed on each RPS instrument channel, except Loss of Load and High Startup Rate, every 92 days to ensure the entire channel will perform its intended function when needed. For the TM/LP Function, the constants associated with the thermal margin

monitors must be verified to be within tolerances.
Bistable Tests The bistable setpoint must be found to trip within th~ AlJowable Values specified in the LCO and left set consistent with the assumptions of the setpoint analysis. A test signal is superimposed on* the input in one channel at a time to verify that the bistable trips within the specified tolerance around the setpoint. This is done with the affected RPS channel (trip channel) bypassed. Any
setpoint adjustment must be consistent with the assumptions of the current setpoint analysis.
The as found and as left values must also be recorded and reviewed for consistency with the assumptions of the Frequency extension analysis. The requirements for this review are outlined in Reference 5. The Frequency of 92 days is based on the reliability analysis presented in topical report CEN-327, "RPS/ESFAS Extended Test Interval Evaluation" (Ref. 5). SR 3.3.1.6 range excore channels using the required every 92 days. rec~...,,... of the uirements for this
Palisades Nuclear Plant B 3.3.1-29 01/20/98 17-h

NRC REQUEST:
3.3.1-12 ITS SR 3.3.1.6 CTS T 4.17-1, footnote (c) JFD 5 Bases insert for SR 3.3.1.6: This SR uses an internally generated test signal to check the 0% and 50% levels read within limits for both the upper and lower detector on the analog meter and on the TMM screen. This check verifies that neither the zero point nor the amplifier gain adjustment have undergone excessive drift since the previous complete CHANNEL CALIBRATION .
17-~
RPS Instrumentation B 3.3.1 BASES SURVEILLANCE SR 3.3.1.6 (continued) REQUIREMENTS The Frequency of 92 days is acceptable, based on plant operating experience, and takes into account indications and alanns available to the operator in the control room. SR 3.3.1.7 A CHANNEL FUNCTIONAL TEST on the Loss of Load and High Startup Rate channels is perfonned prior to a reactor startup to ensure the entire channel will perfonn its intended function if required. The High Startup Rate trip is actuated by either of the Wide Range Nuclear Instrument Startup Rate channels. NI-01/03 sends a trip signal to RPS channels A and C; NI-02/04 to channels B and D. Since each High Startup Rate channel would cause a trip on two RPS channels, the High Startup Rate trip is not tested when the reactor is critical. The High Startup Rate trip Function is required during startup operation and may be operationally bypassed when below lE-4% RTP or.above 13% RTP. The four Loss of Load Trip channels are all actuated by a single pressure switch monitoring turbine auto *stop oil pressure which is not tested when the reactor is critical. Operating experience has shown that these components usually pass the Surveillance when perfonned at a Frequency of once per 7 days'prior to each reactor startup . Palisades Nuclear Plant a 3.3.1-30 01/20/98
RPS Instrumentatio~ra'N,ng Mnahlg)) 1 3.3.r
SURVEILLANCE REQUIREMENTS continued SURVEILLANCE FREQUENCY
[T 4,11 ~1] SR
.fl Once w.ithin 7 days prior to each reactor startup SR~ rform a CHAN~~r:~NCTION~Iz~~ on eac Once wi~hin au matic bypa ~oval fu ~* 92 days 'P{ior to each ~~tor art up SR 3 .3 .1.8 ~ls@ months SR 3.~ Veri~PS RES~NSE TIME ~ithin li~ts. [l month~
on a TAGGERED TEST B S
- - - - fJOIE- - - - - - - - -
t-.if:Ui"~O...i t*TE::"'-TO'Z.S. ,Ai:e'" ~t..t.!t>i:"1' F~to'\ n\l!: C.~~..it*ff:*I.. CAU6~ATIO~*
CEOG STS 3.3-5 I 7 -e__
Rev 1, 04/07/95
CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO JANUARY 6, 1999 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.3, INSTRUMENTATION NRC REQUEST: 3.3.1-13 ITS SR 3.3.1.6, SR 3.3.1.8 CTS T 4.17-1, footnote (c) CTS T 4.17-1, footnote (d), (e); T 4-17-6 JFD 23 JFD 23 states that neutron detectors are specifically excluded from CHANNEL CALIBRATION. Additionally, this JFD states that TSTF-81 is not incorporated. TSTF-81 is approved and provides the allowance in a note to the appropriate SRs that neutron detector* are excluded from CHANNEL CALIBRATION. The ITS proposes to delete all occurrences of this note. This results in requirements to calibrate all neutron detectors because a CHANNEL CALIBRATION includes testing of "all required sensors." Comment: The staff has not accepted changing the STS definition of CHANNEL CALIBRATION. Revise the ITS to incorporate the STS SR Note for instrument channels with neutron detectors. Consumers Energy Response: See response to "Definitions RAI" (first RAI response in this enclosure). Affected Submittal Pages:
See response to "Definitions RAI" .

18
CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO JANUARY 6, 1999 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.3, INSTRUMENTATION NRC REQUEST: 3.3.1-14 STS SR 3.3.1.7 JFD 14 STS SR 3.3.1.7 requires a Functional Test be performed on each automatic bypass removal function once within 92 days prior to each reactor startup. This requirement is deleted in the ITS. JFD 14 states that the 18 month CHANNEL CALIBRATION adequately tests these functions. However, the 18 month CHANNEL CALIBRATION (STS SR 3.3.1.8) Bases does not discuss testing of the automatic bypass removal functions. Comment: Revise the ITS to include this requirement. Otherwise all changes to the STS format and content require an approved NEI TSTF unless a plant specific justification based on operational hardship or design is accepted by the staff for the STS deviation. Consumers Energy Response: Table 3.3.1-1 was revised to add the ZPM Bypass Removal Function as a required Function with SR 3.3.1.8, an 18 month Channel Calibration (which must include a Channel Functional Test), as the required surveillance and an applicability the same as the associated trip functions. In addition, the parts of Section 3.3 Bases which discuss the RPS bypasses were revised and rearranged to add more detail.
For additional discussion on changes related to RPS bypass removal requirements, see response to RAI 3.3.1-6.
Affected Submittal Pages: See response to RAI 3.3.1-6 .
19
CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO JANUARY 6, 1999 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.3, INSTRUMENTATION NRC REQUEST: 3.3.1-17 Resolved during 10/27/98 meeting; Consumers Energy not required to respond. NRC REQUEST: 3.3.1-18 ITS 3.3.1, Applicability CTS 3.17.1 DOCA.2 The CTS Applicability requirement "when there is fuel in the reactor" is deleted without discussion. Comment: Provide a L-DOC for this CTS change. Consumers Energy Response: This change is not a less restrictive change, but rather an administrative change. The applicability limitation of "when there is fuel in the reactor" is retained in the ITS by the use of MODES in describing the applicable conditions. The ITS (and STS} definition of MODE includes "with fuel in the reactor vessel." Therefore, the ITS and CTS applicabilities are equivalent. An additional administrative change has been made to the applicability of LCOs 3.3.1, 3.3.2,
3.3.9, and other places where the term "control rod" is used. The words "control rod" have been replaced, where appropriate, by "full length control rod". This change was necessitated by the ITS omission of ttte 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. The part length control rods have no clutches, remain fully withdrawn during operation, and are unaffected by RPS functions. This is also discussed in the new JFD 8 provided in response to RAI 3.3.1-1.
DOC A.2 has been revised to clarify this part of the administrative change. Affected Submittal Pages: Att 3, DOC 3.3.1, page 1 of 9 (former submittal) Att 3, DOC 3.3.1, page 1 of 10
21

ADMINISTRATIVE CHANGES (A)
ATTACHMENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.3.1, RPS INSTRUMENTATION 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 chapges (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. ed A.2 A. 3 A Note was added to the Actions of CTS 3 .17 .1 which allows separate Condition entry for each RPS Function. The Note in ITS 3.3.1 provides explicit instructions for proper application of the Actions for Technical Specification compliance. In conjunction with the proposed Specification 1.3 - "Completion Times, " this Note provides direction consistent with the intent of the existing Actions for the RPS instrumentation. This change is consistent with NUREG-1432 (TSTF-178). \
Palisades Nuclear Plant Page 1of9 c2_/-o-01/20/98
ATTAC1'ENT 3
ADMINISTRATIVE CHANGES (A)
A.1 DISCUSSION OF CHANGES SPECIFICATION 3.3.1, RPS INSTRUMENTATION All reformatting and renumbering are in accordance with NUREG-1432. As a result,* ihe Technicai 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 subsest_ion. 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 CTS 3.17.1 Applicability is when there is fuel in the reactor, more than one CONTROL ROD is capable of being withdrawn, and the PCS is less than REFUELING BORON CONCEN,TRATION. In the ITS, the proposed Applicability is MODES 1and2, and MODES 3, 4 and 5 with more than one control rod capable of being withdrawn, and the PCS boron concentration is less than REFUELING BORON CONCENTRATION. This Applicability is equivalent to the current Applicability, and the CTS has been revised to reflect the ITS wording. In MODES 1and2, more than one rod must be capable of being withdrawn, and therefore the phrase "more than one full length control rod is capable of being withdrawn" is only associated with MODES 3, 4 and 5. The term "Control Rod" has been revised to read "full length control rod," in footnote (a) of ITS Table 3.3.1-1 (and in the Bases, where appropriate). This change in wording, from "control rod" to (revised) "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. The part length control rods have no clutches, remain fully withdrawn during operation, and are unaffected by RPS functions. The applicability limitation of "when there is fuel in the reactor" is retained in the ITS by the use of MODES in describing the applicable conditions. The ITS (and STS) definition of MODE includes the limitation "with fuel in the reactor vessel." This change only involves a difference in'presentation, and is considered administrative .
Palisades Nuclear Plant Page 1of11 05/30/99
&1-b
CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO JANUARY 6, 1999 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.3, INSTRUMENTATION NRC REQUEST: 3.3.1-15 Resolved during 1qt27 /98 meeting; Consumers Energy not required to respond. Refer to comment 3.3-14 NRC REQUEST: 3.3.1-16 ISTS T 3.3.1-1(page2of2) JFD 28 Comment: JFD 28 is included for justification to replace STS table functions with CTS
.requirements.' JFD 28 does not justify the proposed changes. Provide a correction of the STS markup.
Consumers Energy Response: JFD 28 is not included to justify replacement of STS table functions with CTS requirements, JFD 5 is intended to justify that replacement. JFD 28 is included to explain that TSTF 80.(which revises the STS to limit the applicability of the Axial Power Distribution and Loss of Load trip functions operation above 15% RTP) was not used as the basis for differences between the ITS and STS .
Palisades design does not include an Axial Power Distribution trip, and the CTS applicability for the Loss of Load trip is limited to operation above 17%. Therefore, the ITS requirements to limit the applicability of the Loss of Load trip to operations above 17% is a current licensing basis (JFD 5) not utilization of the changes made by TSTF 80.
Affected Submittal Pages: No page changes .
20
CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO JANUARY 6, 1999 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.3, INSTRUMENTATION NRC REQUEST: 3.3.1-19 Resolved during 10/27/98 meeting; Consumers Energy not required to respond. NRC REQUEST: 3.3.1-20 ITS 3.3.1 ITS 3.3.2 CTS T 3.17.1 CTS 3.17.1 DOC A.5 DOC A.4 The CTS Table column, "Minimum Operable Channels" is deleted. Comment: Revise A.5 to include discussion that total channels is unchanged and ITS conditions do not depend on minimum operable channels for entry into actions. Consumers Energy Response: DOC A.5 has been revised. Affected Submittal Pages:
Att 3, DOC 3.3.1, page 2 of 9 (former submittal)
Att 3, DOC 3.3.1, page 2 of 10
22

ATTACHMENT3 DISCUSSION OF CHANGES SPECIFICATION 3.3.1, RPS INSTRUMENTATION A.4 CTS 3.17.1.6b requires the reactor to be placed in a condition where the affected equipment is not required. In the ITS, this requirement is presented by specifying the actual actions necessary to satisfy this requirement. ITS 3.3.1 Required Actions 1.2.1 and 1.2.2 require that no more than one control rod be capable of being withdrawn or that the PCS boron concentration be of the REFUELING BORON CONCENTRATION. These actions place the plant in a condition where the affected equipment is not required. Since this change only provides more specific actions to be taken, it is considered administrative.
)ee. .f2'e.,1cae.'> "Doc..
A.S- t~AI ;,:s.1- Zo A.5
A.6 CTS Table 3.17.1 Footnote (a) requires two OPERABLE wide range nuclear instrument channels if the Zero Power Mode (ZPM) bypass is used. This Footnote is associated with the High Startup Rate Function. ITS 3.3.1 does not include this Footnote. There are two wide range nuclear instrumentation channels which provide input into the ZPM bypass. Each instrument channel provides input into two High Startup Rate channels and also provides the permissive signal for the ZPM bypass. The ZPM Bypass associated with these Functions is automatically removed when the wide rqnge instruments indicate~ 104 % RTP. Since the ITS Condition D provides adequate guidance when bypass channel(s) are inoperable, this Footnote is not necessary as the bypass channels will either be disabled or the associated High Startup Rate channels will be placed in trip. In addition, the footnote is associated with the High Startup Rate Function, and this bypass is automatically enabled. Therefore, retaining the Note does not provide any restriction to the operator.

Palisades Nuclear Plant Page 2 of 9 01/20/98
ATTACHMENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.3.1, RPS INSTRUMENTATION A.3 A Note was added to the Actions of CTS 3.17.1 which allows separate Condition entry for each RPS Function. The Note in ITS 3.3.1 provides explicit instructions for proper application of the Actions for Technical Specification complia.nce. In conjunction with the proposed Specification 1.3 - "Completion Times, " this Note provides direction consistent with the intent of the existing Actions for the RPS instrumentation. This change is consistent with NUREG-1432 (TSTF-178). A.4 CTS 3 .17 .1. 6b requires the reactor to be placed in a condition where the affected equipment is not required. In the ITS, this requirement is presented by specifying the actual actions necessary to satisfy this requirement. ITS 3.3.1 Required Actions I.2.1 and I.2.2 require that no more than one control rod be capable of being withdrawn or that the PCS boron concentration be of the REFUELING BORON CONCENTRATION. These actions place the plant in a condition where the affected equipment is not required. Since this change only provides more specific actions to be taken, it is considered administrative. A.5 CTS Table 3.17.1 contains. a "Minimum Operable Channels" column. This column is
\ I deleted in the ITS because the Actions in the ITS are based on the number of channels I
inoperable, from the total number of channels, which is specified in LCO 3.3.1. The total number of channels required to be Operable is unchanged from CTS. ITS conditions do not depend on a specified minimum operable channels for entry into actions. The CTS use the specified minimum number of channels to initiate entry into a shutdown action statement. In ITS, when the number of inoperable channels in a trip Function exceeds that specified. in any related Condition, the plant does not meet the I
lo
~
I~ I fYl LCO and no associated Action is provided, therefore, a shutdown is required in I r-n accordance with LCO 3.0.3. is required. The discussion of the change in Required I_ Actions is discussed in DOC M.2. This change only involves a change in presentation. I c--- L format, and is considered administrative. This change is consistent with the I Q:::' presentation format in NUREG-1432. I A.6 CTS Table 3.17.1 Footnote (a) requires two OPERABLE wide range nuclear I instrument channels if the Zero Power Mode (ZPM) bypass is used. This Footnote is I associated with the automatic ZPM bypass removal Function. ITS 3.3.1 does not I include this Footnote, instead, the automatic ZPM bypass removal function has been specified as Function 12 in Table 3. 3. 1-1. ITS Condition D provides guidance when I _J I bypass channel(s) are inoperable, allowing indefinite continued operation with an I-inoperable ZPM bypass removal channel only if the ZPM bypass is not used. These I rri two ITS features are the equivalent of the CTS footnote (a) requirement that these I ('{\ bypass removal channels be operable when the bypass is used. Since the ITS I_ requirements are equivalent to ,the CTS requirements, this change is considered to be I er administrative. I cc Palisades Nuclear Plant Page 2of11 05/30/99 02d.-b
CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO JANUARY 6, 1999 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.3, INSTRUMENTATION NRC REQUEST: 3.3.1-21 ITS 3.3.1 CTS 3.17.1 footnote (a) DOCA.6 Comment: Confirm that ITS requires same two channels during the same applicable conditions to confirm the deletion is an administrative change. Consumers Energy Response: See Discussion and revised DOC A.6 provided in response to RAI 3.3.1-6; Affected Submittal Pages: See response to RAI 3.3.1-6 .
23
CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO JANUARY 6, 1999 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.3, INSTRUMENTATION NRC REQUEST: 3.3.1~22 ITS 3.3.1 CTS 3.17.1 Footnote (b) DOC A.7 Comment: Revise the DOC to include the statement from A.9, "This change essentially moves requirements from one TS to another, and therefore, is considered administrative." Consumers Energy Response: Doc A. 7 has been revised to provide the requested clarification and to address changes made in response to RAI 3.1-01 in our March 1, 1999 response to RAls on ITS Sections 2.0, 3.1, and 3.2. That change moved the Shutdown Margin Requirements to the COLR, and combined SRs 3.1.1.1 and 3.1.1.2, in accordance with TSTF-9. Affected Submittal Pages: Att 3, DOC 3.3.1, page 3 of 9 (former submittal) ( Att 3,, DOC 3.3.1, pa~e 3 of 10 ' .
24

A.7 ATTACHMENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.3.1, RPS INSTRUMENTATION A.8 CTS Table 4.17.1 Footnote (b) requires a calibration of the Variable High Power Function with a heat balance to be performed when power is > 15 % RTP. In addition Footnote (c) requires the excore channels to be calibrated with a test signal. In ITS

Table 3. 3 .1-1, these surveillances (SR 3. 3 .1. 3 and SR 3. 3 .1. 6) are also associated with the Thermal Margin/Low Pressure (TM/LP) Trip Function. This association is necessary since there is no additional testing required. This change is considered administrative, and is consistent with NUREG-1432.
A.9 CTS 2.3 specifies the requirements for the RPS Limiting Safety System Settings. The Applicability of this Specification is when the associated RPS channels are required to
. be OPERABLE by Specification 3.17.1. In addition, CTS 2.3.l specifies that if the RPS instrument setting is not within the allowable settings of Table 2. 3 .1, the instrument must be declared inoperable and complete corrective action as directed by Specification 3 .17 .1. In the ITS the RPS Allowable Values are listed in ITS Table 3.3.1-1 and the ITS 3.3.1 Actions provide identical protection. This change essentially moves requirements from one Technical Specification to another, and therefore, is considered administrative. This change is consistent with NUREG-1432 .
Palisades Nuclear Plant Page 3 of 9 c!lz/--a__
01/20/98
ATTACHMENT3
A.7 DISCUSSION OF CHANGES SPECIFICATION 3.3.1, RPS INSTRUMENTATION CTS Table 3 .17 .1 Footnote (b) provides a restriction that the low power bypass (i.e., Zero Power Mode Bypass) cannot be enabled unless the SHUTDOWN BORON.
CONCENTRATION for the COLD SHUTDOWN condition is achieved. This . I I I provision is not inciuded in the associated ITS Table 3.3.1-1 Footnote (o) since the I SHUTDOWN MARGIN (SDM) requirements are met through ITS Section 3 .1. The I CTS definition of SHUTDOWN BORON CONCENTRATION required is > 2 % D-p I. with all control rods inserted in the core and the highest worth control rod fully I rJ withdrawn. In ITS Specification 3.1.1, the minimum SDM required during any I N cooldown is specified in the COLR, and the COLR requirements must be calculated in I accordance with approved methodology. Changes to the COLR must be performed in I ~ accordance with the requirements of ITS 5.6.5. The COLR SDM requirements provide I
the same assurance that the reactor will remain shutdown as do those in CTS. I ""
Therefore, since appropriate SDM must still be maintained throughout plant operation and cooldown, the removal of this provision can be considered administrative. This I I ~ change essentially moves requirements from one Technical Specification to another, I and therefore, is considered administrative. This change is consistent with NUREG- I
-1432. I A.8 CTS Table 4.17.1 Footnotes (b) and (c) require, a calibration of the Variable High I
Power trip Function channels (excore nuclear power and AT power) with a heat
. balance, and calibration of the excore power channels with a test signal. In ITS, a requirement to adjust the instrumentation if the absolute difference between the instrument readings and the results of the heat balance exceeds l .5 %. That requirement corresponds to "th~ necessary range and accuracy" requirement of the I
I I I t" I~ Channel Calibration definition, and does not constitute a change in requirements. Also I. in ITS, Table 3. 3 .1-1, these surveillances. (SR 3. 3 .1. 3 and SR 3. 3 .1. 6) are also I~ associated with the Thermal Margin/Low Pressure (TM/LP) Trip Function. In CTS, I ~' this association is provided by a table in the bases. This association is necessary since I r" the power signal from the excore power range is an input to the TM/LP trip function I setpoint calculator. There is no additional testing required. This change is considered I a:. administrative, and is consistent with NUREG-1432. I~ A.9 CTS 2.3 specifies the requirements for the RPS Limiting Safety System Settings. The
. Applicability of this Specification is when the associated RPS channels are required to be OPERABLE by Specification 3.17.1. In addition, CTS 2.3.1 specifies that if the RPS instrument setting is not within the allowable settings of Table 2. 3 .1, the
instrument must be declared inoperable and complete corrective action as directed by Specification 3.17.1. In the ITS the RPS Allowable Values are listed in ITS Table 3.3.1-1 and the ITS 3.3.1 Actions provide identical protection. This change essentially moves requirements from one Technical Specification to another, and therefore, is considered administrative. This change is consistent with NUREG-1432.
Palisades Nuclear Plant Page 3of11 05/30/99
~</-b
CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO JANUARY 6, 1999 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.3, INSTRUMENTATION NRC REQUEST: 3.3.1-23 ITS 3.3.1 CTS T 4.17.1 Footnote (b) DOC A.8 This DOC conciudes that the addition of iTS SR to TMiLP is necessary since there is no additional testing required. Comment: Explain why the addition is purely administrative. Consumers Energy Response: DOC A.8 has been revised. Affected Submittal Pages: Att 3, DOC 3.3.1, page 3 of 9 (former submittal) Att 3, DOC 3.3.1, page 3 of 10
25

A.7 ATTACHMENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.3.1, RPS INSTRUMENTATION CTS Table 3 .17 .1 Footnote (b) provides a restriction that the low power bypass (i.e., Zero Power Mode Bypass) cannot be enabled unless the SHUTDOWN BORON CONCENTRATION for the COLD SHUTDOWN condition is achieved. This provision is not.included in the associated ITS Table 3.3.1-1 Footnote (b) since the SHUTDOWN MARGIN (SDM) requirements are met through ITS Section 3.1. The CTS definition of SHUTDOWN BORON CONCENTRATION required is > 2 % AP
.with all control rods inserted in the core and the highest worth control rod fully withdrawn. In ITS Specification 3.1.1, the minimum SDM required during any cooldown must be > 2 % llp and is increased to > 3.5 % AP when the PCS average temperature is reduced below 525°F. Therefore, since the CTS SHUTDOWN BORON CONCENTRATION requirements are maintained throughout the plant cooldown, the removal of this provision can be considered administrative. This change is consistent with NUREG-1432.
Set:- 1?£\llS~"> A.B g41 "J.1.1-z:s A.8 A.9 CTS 2.3 specifies the requirements for the RPS Limiting Safety System Settings. The Applicability of this Specification is when the associated RPS channels are required to be OPERABLE by Specification 3 .17 .1. In addition, CTS 2. 3. 1 specifies that if the RPS instrument setting is not within the allowable settings of Table 2.3.1, the instrument must be declared inoperable and complete corrective action as directed by Specification 3 .17 .1. In the ITS the RPS Allowable Values are listed in ITS Table 3. 3 .1-1 and* the ITS 3. 3 .1 Actions provide identical protection. This change essentially moves requirements from one Technical Specification to another, and therefore, is considered administrative. This change is consistent with NUREG-1432 .
Palisades Nuclear Plant Page 3 of 9 c?..5 - 6.--
01120/98
ATTAC1'ENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.3.1, RPS INSTRUMENTATION A.7 CTS Table 3 .17 .1 Footnote (b) provides a restriction that the low power bypass (i.e., Zero Power Mode Bypass) cannot be enabled unless the SHUTDOWN BORON CONCENTRATION for the COLD SHUTDOWN condition is achieved. This provision is not included in the associated ITS Table 3. 3 .1-1 Footnote (b) since the I SHUTDOWN MARGIN (SDM) requirements are met through ITS Section 3 .1. The CTS definition of SHUTDOWN BORON CONCENTRATION required is > 2% e.p la with all control rods inserted in the core and the highest worth control rod fully :~ withdrawn. In ITS Specification 3 .1.1, the minimum SDM required during any I I cooldown is specified in the COLR, and the COLR requirements must be calculated in I-: accordance with approved methodology. Changes to the COLR must be performed in I rf\ accordance with the requirements of ITS 5.6.5. The COLR SDM requirements provide I l'n the same assurance that the reactor will remain shutdown as do those in CTS. I Therefore, since appropriate SDM must still be maintained throughout plant operation I ct and cooldown, the removal of this provision can be considered administrative. This I Ct' change essentially moves requirements from one Technical Specification to another, and I therefore, is considered administrative. This change is consistent with NUREG-1432. I A.8 CTS Table 4.17.1 Footnotes (b) and (c) require, a calibration of the Variable High I
Power trip Function channels (excore nuclear power and .6.T power) with a heat balance, and calibration of the excore power channels with a test signal. In ITS, a requirement to adjust the instrumentation if the absolute difference between the instrument readings and the results of the heat balance exceeds 1.5 % . That requirement corresponds to "the necessary range and accuracy" requirement of the Channel I
I I ('{') IN I Calibration definition, and does not constitute a change in requirements. Also in ITS, I . Table 3. 3 .1-1, these surveillances (SR 3. 3 .1. 3 and SR 3. 3 .1. 6) are also associated with the Thermal Margin/Low Pressure (TM/LP) Trip Function. In CTS, this association is I rt: I ('{) provided by a table in the bases. This association is necessary since the power signal I from the excore power range is an input to the TM/LP trip function setpoint calculator. I~ There is no additional testing required. This change is considered administrative, and I~ is consistent with NUREG-1432. I A.9 CTS 2.3 specifies the requirements for the RPS Limiting Safety System Settings. The Applicability of this Specification is when the associated RPS channels are required to be OPERABLE by Specification 3.17.1. In addition, CTS 2.3.1 specifies that if the RPS instrument setting is not within the allowable settings of Table 2. 3 .1, the instrument must be declared inoperable and complete corrective action as directed by Specification 3 .17 .1. In the ITS the RPS Allowable Values are listed in ITS Table 3. 3 .1-1 and the ITS 3. 3 .1 Actions _{Jrovide identical protection. This change essentially moves requirements from one Technical Specification to another, and
therefore, is considered administrative. This change is consistent with NUREG-1432 .
Palisades Nuclear Plant Page 3of11 05/30/99
CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO JANUARY 6, 1999 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.3, INSTRUMENTATION NRC REQUEST: 3.3.1-24 ITS 3.3.1 CTS T2.3.1 DOC A.10 This DOC discusses changes to the CTS that result in deleting limitations associated with RPS trip settings with three primary coolant pumps operating. Comment: Revise the DOC to be an L-DOC. Consumers Energy Response: This change is not Less Restrictive. DOC A.1 O has been revised to clarify the basis for this change classification. Affected Submittal Pages: Att 3, DOC 3.3.1, page 4 of 9 (former submittal) Att 3, DOC 3.3.1, page 4 of 10
26

A.10 SEE. T{evt~e°r.) A.10 ATTACHMENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.3.1, RPS INSTRUMENTATION

A.11 CTS 4.18.2. lb requires a surveillance to compare the individual excore channel measured AO (ASI) to the total core AO measured by the incores, and associated calibration of the excore channel if the difference exceeds 0.02. ITS includes this
. requirement as SR 3.3.1.4 and applies it to both the Variable High Power Trip (VHPT) and the Thermal Margin/Low Power (TM/LP) Functions since these excore channels provide input to each of these RPS trips. This change is considered administrative since it provides only a different format for identifying the requirements.
A.12 CTS Table4.17.1, item 16, requires a surveillance to verify control room temperature is acceptable for operability of the Thermal Margin Monitor (TMM). The TMM is not* specifically identified in the CTS LCO, nor in the ITS. However, the TMM provides an input to both the VHPT and the TM/LP Functions. Therefore, this CTS requirement is reflected in ITS SR 3.3.1.2 which is identified in ITS Table 3.3.1-1 as a required SR for both the VHPT and the TM/LP Functions. This change is considered administrative since it provides only a different format for identifying the requirements .
Palisades Nuclear Plant Page 4 of 9 c9lo -
01/20/98
ATTAC1'ENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.3.1, RPS INSTRUMENTATION A.10 CTS Table 2.3.1 provides.the RPS Trip Setting Limits with four Primary Coolant Pumps (PCPs) operating and with three PCPs operating. ITS Table 3.3.1-1 provides the required Allowable Values for four PCP condiiim1s, the t.lrree PCP settings have been omitted. With the RPS Operable (including the trip setpoints being adjusted as required by ITS), but with less than four PCPs operating, the Low Primary Coolant System Flow Trip will be actuated, ensuring all full length control rods are inserted. Therefore removal of the "Three Primary Coolant Pumps Operating" trip settings will not result in an unacceptable operating situation. In addition, ITS LCO 3.4.4, "PCS Loops - MODES 1 and 2," requires the plant to be placed in MODE 3 if less than four PCPs are operating (an action the RPS performs automatically). The CTS Primary Coolant System Section requires four primary coolant pumps to be in service, but contains an allowances to operate for up to 12 hours with only three (of the four installed) primary coolant pumps in service. The CTS Safety Limits Section contains RPS setting requirements for both the three and four pump conditions. This allowance for three pump operation is a holdover from the original Palisades Technical
Specifications which allowed continuous operation, at specjfied RPS settings, with two *.
three, or four pump operation. The deletion of the remaining 12 hour allowance for three pump operation is addressed in Section 3.4 of the ITS submittal as change 3.4.4 M.l. Both ITS LCOs 3.4.1and3.4.4 require four primary coolant pumps to be in operation, rendering the three pump RPS settings unnecessary and unusable. Since three pump operation will no longer be allowed (due to the more restrictive change in Section 3.4) deletion of these alternative settings, which can no longer be used, is considered to be an administrative change. A.11 CTS 4.18.2.lb requires a surveillance to compare the individual excore channel measured AO (ASI) to the total core AO measured by the incores, and associated calibration of the excore channel if the difference exceeds 0.02. ITS includes this requirement as SR 3.3.1.4 and applies it to the Thermal Margin/Low Power (TM/LP) Function since these excore channels provide input to this RPS trip. This change is corisidered administrative since it provides only a different format for identifying the requirements. A.12 CTS Table 4.17 .1, item 16, requires a surveillance to verify control room temperature is acceptable for operability of the Thermal Margin Monitor (TMM). The TMM is not
specifically identified in the CTS LCO, nor in the ITS. However, the TMM provides an input to both the VHPT and the TM/LP Functions. Therefore, this CTS requirement is reflected in ITS SR 3.3.1.2 which is identified in ITS Table 3.3.1-1 as a required SR for both the VHPT and the TM/LP Functions. This change is considered administrative since it provides only a different format for identifying the requirements.
Palisades Nuclear Plant Page 4of11 05/30/99 cilo-b.
CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO JANUARY 6, 1999 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.3, INSTRUMENTATION NRC REQUEST: 3.3.1-25 CTS 3.17.1 DOC M.1 Comment: Third paragraph on the third line, revise to read "channel bypass". Comment: The fourth paragraph justifies the addition of ITS conditions E, F, G, and Has consistent with current practice. Add details to this discussion to explain the safety basis for this position.
Consumers Energy Response:
Doc 3.3.1 M.1 has been revised, addressing this issue, in response to RAI 3.3.1-6. Affected Submittal Pages: See response to RAI 3.3.1-6 .
27
CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO JANUARY 6, 1999 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.3, INSTRUMENTATION NRC REQUEST: 3.3.1-26 ITS 3.3.6 CTS Tables 3.17.6; 4.17.6 CTS Tables 3.17.6 and 4.17.6, Function 19, Fuel Pool Rad Monitor, and Function 20, Containment Refueling Radiation Monitor, show that these functions are moved to ITS 3.3.6, Refueling Containment Radiation High Instrumentation. ITS 3.3.6 does not include instrument functions by these names. Comment: Describe the safety function(s) performed these instrument channels and the safety systems they support. Consumers Energy Response: The markup of CTS Tables 3.17.6 and 4.17.6 in the pages associated with ITS LCO 3.3.1 was intended to show that items 19 and 20 are addressed under ITS LCO 3.3.6. In these tables of the CTS markup pages associated with ITS LCO 3.3.6, item 19 is shown as having been relocated. DOC R.4 discusses this item requirement being relocated to the ORM. In the CTS markups associated with ITS LCO 3.3.6, Requirement 20 of these tables (Containment Refueling Radiation Monitor) is shown as becoming ITS LCO 3.3.6, which calls these monitors "Refueling Containment High Radiation Instrumentation." The markup of CTS pages 3-75, 3-78, and 4-82, which show Table 3.17.6 and the associated actions, and Surveillance 4.17 .6 were included with ITS LCO 3.3.1 to show the location, in ITS, of requirement item 12, the LiT-Flux comparitor alarm. That treatment was in error. That requirement should have been relocated because it has no protective function. The affected pages have been corrected; DOCs M.4 and M. 7 have been deleted; and a new relocation DOC, R.5, has been included. There are no longer any requirements on these three CTS pages associated with ITS LCO 3.3.1. The safety function of CTS Table 3.17.6 item 19 is discussed in the CTS 3.17.6 bases; the safety functions for CTS item 20 is discussed in the bases for CTS 3.17.6 and for ITS LCO 3.3.6. Affected Submittal Pages: Att 3, CTS 3.3.1, page 3-75 Att 3, CTS 3.3.1, page 3-78 Att 3, CTS 3.3.1, page 4-82 Att 3, CTS 3.3.6, page 3-78 Att 3, CTS 3.3.6, page 4-82 Att 3, CTS 3.3, page 3-75 Att 3, CTS 3.3, page 3-78 Att 3, CTS 3.3, page 4-82 Att 3, DOC 3.3.1, page 6 of 9 Att 3, DOC 3.3.1, page 7 of 9 (former submittal) Att 3, DOC 3.3.1, page 7 of 10 Att 3, DOC 3.3.1, page 8 of 1O Att 3, DOC 3.3.1, page 8 of 9 Att 3, DOC 3.3.1, page 2 of 2 (former submittal) , Att 3, DOC 3.3.1, page 2 of 2 28
. 3. I 1

3. 17 INSTRUMENTATION SYSTEMS Action (continued)
_ With one requi.red ection channel (7a, b, c / d) inoperable: Restore the channel o OPERABLE status prior to£. next sta frora COLD SHUTDOWN I With two or thr~required Leak Detection inoperable: / _
/
3.17.6.8 a) 3.17.6.9 for one or a) 17.6.10 With one PORV .Blo one or more valves *
a) Restore the chan'Rel from COLO SHUTOO a)
[G.ic A] I 3.17.6.12.1 Q?A M.1] I a) [C....:.ocJI 3.17.6.12.Z r~c.1] up ~equence 1 I Ver1f,¥/'that all regu~tt1ng groups~r. w1th1n the lj.ftl1ts o Spestf1cation,3.10 _within 15 m1nut after movemeit of any reoulat i ng rod. .-' I
\
Amendment No. 31 i71 9&1 981 11&1 1181 1211 1241 1291 Bi, '\tiZ October 26, 19 1.-s.I --zc;,
- 3-75 ' Hr~ p~~ie... R'eMove.l"';> FrcoM LC6 3.~./ S"ecn~1~ ?iur.. t
10
~8-o__
G-], 3. 1-1J 3.17 INSTRUMENTATION SYSTEMS

Table~
/-~ 3. 1- !)
(continued) Instrymentation Ooerating Regyirernents for Other Safety Fynctions Hinimu11 Required OPERABLE Applicable ti2 Channels Channels Condjtjons
10. l/V VI .

11. ~ ii*~
c1] [Cl] 12. 4@ 0 """ 2
13. Bod G oup Sequence Centro Alam

14. (Cone Boric A~ Tank Lo Lev1l Alam 15
Abov1 251 RATED POWER.

....-16. AXIA~
Alarm SHAPE INDEX
17. ~c* ction Valve terl cks bov1 251 RATED POWER.
18.
19. 0

20. 0
[C.01JD ~o1eJ (a) Sp iftcations 3.0.4 ~ot 1pplic1bl1. (b) 3. O. , and 4. O. 4 are I~~~~~~~~~~-'""--~~~~~_.._~~~~ Amendment No. 31 &7 1 9&1 981 11&1 1181 1211
f( ff / 3. ). / - z b
-rnr~ 7M'L .~~MOU~";::::. F~""
L 66 "S.l. I ~"MD"/ 3-78
4.17 INSTBUHENTATIOH SYSTEMS TEST~~.3.
Tible ~ -* (continued) 1-1) Instrymentat1on Surye1llance Reqyirements for Other Sifetx Functions
r. 3 1 J D 7 L.5"-"'* * *'
~
Ni '3'.).1,51 ,......r. ~ 3 3 CHANNEL ~ "L.~"" * *'
~- -
I'*:. I' ~1' CHANNEL CHECK FUNCTIONAL TEST CHANNEL
11. NA (1] [91 12. 12 hours 13.
14. 15.
16. ASI Ahna 17.
18.
19. 31 diyS 20
31 days

t verif1cat1on only.
R,41 -s.~. t -z~ fC.£.MO\J"-'T';> F~0"1 Amendment Ho. ~. ~. 171
) e. c..:r1 (j '"( ' April 5, 1996
4-82
-llS 3.17 INSTRUMENTATION SYSTEMS Table 3.17.6 (continued)
Instrymentat1on Ooerattng Begy1rements for Other Safety fynct1ons Minimum Required OPERABLE App 11 cable HQ. Instrument Channels Channels Conditions
10. PORV B ck Valve 2/valve 1..
Position ndtcation
11. Detector 0

12. .Power Comparator 2

13. 1

14. Cone oric Acid Tank ~
tank 0 Lo Le 1 Al arm
15. Ex co 0
17.
Dev1a
, . 16. AXIAL S PE INDEX Alarm SOC S~ion Valve Interl ks 2
2 0 TED POWER
18. Power Inserti \2 1 HOT STA~BY and ab~e
~ 3, I
19. Fuel Pool Area 2'bl 0 Radiation Monitor
[3.3.(,.] 20. Containment Refueling ~ 0 FUEl:-ING "OPEMTIO Radiation Monitor w irradiated fuel dvr;11, 01!£ /lL.'flmATJ8MJ. *@ 1n the Containment. t1,,J-1d >>i~ "eu,~1.+ t>t
/[JJ (a) Spec1<jcations 3-..,0.4 and ~0.4 are~ *appli~le~
(b) Specifica~ns 3.0. 3.0.4, a Amendment No. 3, &7, 9&, 98, 11§, 118, 121, 124, 129, 13&, 162 October 26, 1994 3-78
3.3.""
r
~
!1'.S ', l 7 lHSTBtmEMTATJON SX$TDSS TUI$
Tab1t 4.17.S (continued) Instryrntotat loo su.r;ttlane1 R1Qti1 c1m1nt1 fgc ot),tc s t.Jy Func:t lOJ\S CHANNEL CHANNEL FUNCTIONAL CHANNEL
!nstrymtnt CHECK TEST CALI BBAT! ON l l. SWS Bruk 01ttctor HA 18 months
12. F1ux * &T Co1110arat r 12 hours

13. Rod Group S1qu1n 1 MA Control/Ala,..
l'* BAT Low L1v11 A1&"9 15.
u. 1--~~-1-~-~~~~-r-~~~~-+-~~~~--,~~~
17. L----+-~~~~~__,..~--~~-r~~--~-r--~--+ 1
18. MA 31 day1 *

19. 24 houri 31 days*
[5R3.3.~.o 20. Conta1n.. nt R1fu1ltn~ ~ours 31 days 18 months 0~ .?.3.1..2.J Radiation Monitor f'Z.
@)
M. I [sit J.1 ~*D [ :;; (d)~polnt vtrH1cat1on on 1Y*~1-----
\ --------~ *~
I
*- A191ndlltnt Ho. ~. 44, 171 Aprn s, 1996
l.17 . !NSTBUMENTAT!ON SYSTEMS
~ (cont1nu*d) l.17.5.7. Luk O*Uc:tton c:han *1 (71, b, c:, or d) &)
l.17.5.7.Z Wit two or thrtt rtquirtd 1ak Otttction channels c:, or d) I p*r&blt:
&) Bestor* thr*t c:hanntls to OPERAILE status *tth n 30 days.
l.17.5.8 Wtth one Primary Saf*ty Valvt Positton Indicator c:hanntl lnop*rablt, fo ont or more valves:
1) estort the c:hanntls to f ~ COLD SHUTDOWN.
3.17.5.9 With one o two PORV Positton lnd1 tor c:hanntls tnaptrablt or on* or rnort Y& 1YU:
a) Restore th c:hanntls to OPERAILE s tus prtor to th* ntxt s fro* COLD S TOOWN.
3.17.5.10 Wtth one PORY Block alvt Posttioft Indicator hanntl 1noptrab1t, one or more valves: a) Restore tht channel o OPEii.Ail[ status prior o tht next startup fro* COLO SHUTDOWN, a
b) .1 f the PORY path ts rt~ fot' LTOP or u a PC ~*nt, and tht lvt postt1on lights tnop ablt, vertfy PORY Bloc Valve Is open 12 hours.
3.17.5.11 Wtth tht
&) ERAILE status prior to th* next c:hanntl inoperable:
f:AT. st1rtup. ~ 3,3, 1-Zl&>
Amtnd1111nt Ho. J 1 11 1 9'1 91 1 1111 1111 1211 1241 1291 136, 162 3*75 October 26, 1994
I I
3,3 T<.£LOCA 7EIJ

l.17 !NSTRUMEHTATION sysTEM$
Tablt l.17.S (continued) M1n1 rm.ui RtQuf rtd OPERABLE Appl icablt Channt11 Ch1001J1 Condltlons At 111 tlmts n1tss tht l>tS 1s tprtssur1ztd a vented I accordanc1 w1t Specif at1on l.1.8. 0 ft~lux-~rCo~r ~ ~
lS~,)(COr*
15. AX evht1on Al&r11 O*tt~r Ah SHAPE INOEX
~
414 ~2
~ Abovt 251 ~POWER. . ~251 RATED ";ov~~
17 ../f6C Su ct 1on ~.wt"" 0 Abaft-ZOO ps1a /
/ Int1rlocky ~PCS Prusurt/
18. owtr Otpt~nt sert1on Alal'9...__

19. ~u1Y?Oo't Arn ~ 2 111 ~*0 ~i1 1s lo~
.-Rt'dtation Mo~ ~ . _...........-fuel pool art~
lb) ec1flc~3.~.~o.4-~t 1pp~bl1.~ Amtndllltnt No. J 1 171 911 91 1 lllu Ill; lit. 1241 1291 !Hi, 162 October 26, 1994 3*71 ..
3._; R£LOCA tE..b 4.17 ICSjBVCEHjAjtOD SY$iW j£Sj5 Tabl* 4.17., (contlnu1d) Instru111tntat !on su17l*ll1 anct haul r1m1nts for Othtr Lf UJ furict LOOS CHANNEL CHANNEL FUNCTIONAL
~~~~~ IEH
11. 11 IDOOthS eAI:
'3 ."3. \-ZG:>
I I Z.
13. d Grau~ S1qu1nc1 18 months ontrol/ 111..
BAT Lov L1v1\ Al"' NA Not R1qulr1d NA 11 IDOnths IS. ASI Ala,.. NA
17. NA
18.
19. zo. NA 31 days 1* ll days ll days (d)
4-82
-.endllent Ho. ~. ~. 171 April S, 1996
M.2 ATTAC1'ENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.3.1, RPS INSTRUMENTATION .
CTS 3 .17 .1. 6 requires specific actions be taken when the number of OPERABLE channels is less tha.'1 specified in the "Minimum OPERABLE Channels" column of Table 3.17.1. The actions are to place the reactor in HOT SHUTDOWN within 12 hours; and place the reactor in a condition where the affected equipment is not required, within 48 hours. In the ITS, the actions when the "number of OPERABLE Channels" is less than the CTS minimum required is to enter LCO 3.0.3. The actions of ITS LCO 3.0.3 are to initiate action within 1 hour to place the plant, as applicable, in MODE 3 within 7 hours; in MODE 4 withiri 31 hours; and in MODE 5 within 37 hours. Since this change will require the plant to be in a lower MODE in a shorter time frame, this change is considered more restrictive. This change continues to assure that a plant shutdown can be achieved in a controlled manner without challenging plant systems. This change is an additional restriction on plant operations and is consistent with NUREG-1432. M. 3 CTS 3 .17 .1. 6b requires the reactor to be in a condition where the affected equipment is not required within 48 hours. ITS 3.3.1, ACTION I requires the same condition be achived within 6 hours. Since the plant is required to be in a lower MODE in a shorter
. time frame, this change is considered more restrictive. With the plant required to be in MODE 3 in 6 hours, de-energizing the clutch power supplies or borating to the

REFUELING BORON CONCENTRATION can also be performed within the same time frame. These actions will not challenge plant systems, since the reactor is subcritical in MODE 3. This change is consistent with NUREG-1432.
N(J* u~e.. ~ RM 1."?.1- z6 M.4 M.5 CTS Table 3.17.1 Footnote (c) provides an allowance to change the setpoint of the low power bypass setpoint from 104 % RTP to 10-1 % RTP during LOW POWER PHYSICS TESTING. Since this allowance is not used, it is not included in the ITS. Since this change deletes an allowance and the plant will no longer be able to change this low power setpoint during the ITS PHYSICS. TEST (see Discussion of Changes for ITS Section 1.0), this change is considered more restrictive. This change is consistent with
NUREG-1432 .
Palisades Nuclear Plant Page 6of9 01/20/98
ATTACH1\1ENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.3.1, RPS INSTRUMENTATION M.2 CTS 3.17.1.6 requires specific actions be taken when the number of OPERABLE channels is less than specified in the "Minimum OPERABLE Channels" column of Table 3 .17 .1. The actions are to place the reactor in HOT SHUTDOWN within 12 hours; and place the reactor in a condition where the affected equipment is not required, within 48 hours. In the ITS, the actions when the "number of OPERABLE Channels" is less than the CTS minimum required is to ~nter LCO 3.0.3. The actions of ITS LCO 3.0.3 are to initiate action within 1 hour to place the plant, as applicable, in MODE 3 within 7 hours; in MODE 4 within 31 hours; and in MODE 5 within 37 hours. Since this change will require the plant to be in a lower MODE in a shorter time frame, this change is considered more restrictive .. This change continues to assure that a plant shutdown can be achieved in a controlled manner witlJ.out challenging plant systems. This change is an additional restriction on plant operations and is consistent with NUREG-1432. M. 3 CTS .3 .17 .1. 6b requires the reactor to be in a condition where the affected equipment is not required within 48 hours. _ITS 3. 3 .1, Condition G Actions require the same condition be achieved within 6 hours. Since the plant is required to be in a lower MODE in a shorter time frame, *this change is considered more restrictive. With the
plant required to be in MODE 3 in.6 hours, de-energizing the clutch power supplies or borating to the REFUELING BORON CONCENTRATION can also be performed within the same time frame. These actions will not challenge plant systems, since the reactor is subcritical in MODE 3. This change is consistent with NUREG-1432.
M.4 Not Used M.5 CTS Table 3.17.1 Footnote (c) provides an allowance to change the setpoint of the low power bypass setpoint from 10~ % RTP to 10-1 % RTP during LOW POWER PHYSICS TESTING. Since this allowance is not used, it is not included in the ITS. Since this change deletes an allowance and the plant will no longer be able to change this low power setpoint during the ITS PHYSICS TEST (see Discussion of Changes for ITS Section 1.0), this change is considered more restrictive. This change is consistent with NUREG-1432. M.6 New Conditions Band C and new Required Actions B.l and C.1 have been added for
/
the Loss of Load and High Startup Rate Functions. These actions require restoration of the inoperable channel(s) prior to re-entering operational conditions during which the Functions are required to be Operable. CTS Actions applicable to single channel inoperability for other RPS trip Functio~ are not required for High Startup Rate and Loss of Load. In ITS, some Action must be specified to avoid an LCO 3.0.3 entry upon channel inoperability. These added Actions are unique to Palisades, since, to the best of our knowledge, Palisades is the only CE plant with non-safety grade High Startup Rate and Loss of Load trips which are not credited in the safety analyses. Palisades Nuclear Plant Page 7o(11 05/30/99 a8-
M.6 ATTACI'ENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.3.1, RPS INSTRUMENTATION CTS Table 3.17.1 permits operational bypasses to be used for certain Functions. ITS Table 3.3.1-1 requires that the operational bypasses for these Functions be automatically removed when the Function variable reaches the value at which use of the bypass was permitted to be placed in use. CTS Table 4.17.1 is also revised to include bypass removal channels in the CHANNEL CALIBRATION. This change imposes additional requirements which did not previously exist, and is considered more restrictive. This change will provide additional assurance that RPS Trip Functions are not bypassed when the Function is required to be OPERABLE. This change is consistent with NUREG-1432.
Uif.Lte-\lE..~)
LESS RESTRICTIVE CHANGES - REMOVAL OF DETAILS TO LICENSEE CONTROLLED DOCUMENTS (LA)
LA.1 CTS Table 4.17.1, Functional Unit 15, "Thermal Margin Monitor" requires that the constants be verified every 92 days. In the ITS, this surveillance is considered to be a part of SR 3.3.1.5, the CHANNEL FUNCTIONAL TEST for the Thermal Margin/Low Power (TM/LP) Function and other Functions. The constants of the Thermal Margin Monitor are specified in CTS Table 2.3.1 (ITS Table 3. 3 .1-2). These constants must be verified to ensure the trip setpoint of the TM/LP Function is set correctly. The details of the CHANNEL FUNCTIONAL TEST have been relocated to the Bases. The Bases for ITS SR 3.3.1.5 states that the thermal margin monitor constants must also be checked to be within tolerances. Removing the details of the CHANNEL FUNCTIONAL TEST from the CTS and placing them in the Bases of the ITS is acceptable since these details are not pertinent to the actual requirements. Placing these details in the Bases provides adequate assurance that they will be maintained since the Bases are ~ontrolled by the Bases Control Program proposed in ITS Chapter 5.0. This change is consistent with NUREG-1432 .
Palisades Nuclear Plant Page 7 c98-~
of~ 01/20/98
ATTACHMENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.3.1, RPS INSTRUMENTATION M.6 (continued) CTS allow continued operation for an unlimited period of time with one High Startup Rate of Loss of Load Trip inoperable. Therefore, LCO 3.0.4 would not require their restoration prior to re-entry of their applicable conditions, even after an extended shutdown. This open ended allowance is unnecessary and was not intended when the current CTS Actions were written. CTS contain other Actions for inoperable channels
. of supplementary equipment which allow continued operation for an unlimited period of time, but require restoration prior to the next startup. (See CTS Actions 3.17.6.8, 9, 10,. 11, *& etc.) The ITS Actions provide this same limitation.
The proposed Required Actions and Completion Times are considered to be acceptable because:
1. They are consistent with current practice, and do not impose unacceptable operational restriction,

2. They are more restrictive th.an the Actions required by CTS,
3.
4. With one channel inoperable, the trip Function is still capable of initiating a reactor trip with a two-out-of-three logic, and These trip Functions are not credited in the plant safety analysis. This is a More Restrictive change because the ITS require restoration of an inoperable trip channel where the CTS do not. M.7 Not used .
Palisades Nuclear Plant c98-Y Page 8of11 05/30/99

LESS RESTRICTIVE CHANGES (L)
ATTACHM:ENT 3 DISCUSSION OF CHAi'\fGES SPECIFICATION 3.3.1, RPS INSTRUMENTATION L.1 CTS 3 .17 .1. 3 requires entry when two RPS trip units or associated instrument channels are inoperable in one or more Functions. In the ITS, these conditions are addressed in ITS 3.3.1, Condition B. The difference is that the Condition B Required Actions include a Note which excludes the applicability of LCO 3.0.4. This provision was added to allow MODE changes even though two channels are inoperable, with one channel in trip. In this condition, with one channel in trip the protection system is in a one-out-of-two logic. Since the probability of a common cause failure affecting both of the OPERABLE channels during the 7 day Completion Time to restore one trip unit to OPERABLE status is remote, and the low probability of occurrence of an event during this interval, this change is considered acceptable. This change is consistent with NUREG-1432. L.2 The Frequency of the CHANNEL FUNCTIONAL TEST associated with RAJ 3.).1-Z' certain RPS Functions CTS Tables 4.17 .1 'i!i 1. ii} is 31 days. *In ITS SR 3.3.1.5, the proposed Frequency is 92 days. The proposed change revises the CHANNEL FUNCTIONAL TEST Frequency for certain RPS Functions from 31 days to 92 days. Justification for extending the test interval is provided in Combustion Engineering (CE) Topical Report CEN-327-A Supplement 1, "RPS/ESFAS Extended Test Interval Evaluation," dated January 1989. NRC acceptance of the CE topical report was provided in a letter to the CE Owner's Group (CEOG) dated November 6, 1989. As discussed in the NRC acceptance letter, Consumers Energy has reviewed the instrument drift information for each instrument channel involved and has determined that drift occurring in that channel over the extended surveillance interval will not cause the setpoint to exceed it's Allowable Value as cakulated by our setpoint methodology. Onsite records of the recorded instrument drift are available for audit. Therefore, it is* proposed that the CHANNEL FUNCTIONAL TEST be performed in accordance with ITS SR 3.3.1.5, at a Frequency of 92 days. This change is consistent with NUREG-1432. ,
Palisades Nuclear Plant Page 8 of 9 a8-(Y) 01/20/98
ATTAC1'ENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.3.1, RPS INSTRUMENTATION L.2 The Frequency of the CHANNEL FUNCTIONAL TEST associated with certain ' RPS Functions CTS Tables 4.17.1 is 31 days. In ITS SR 3.3.1.5, the proposed rn. Frequency is 92 days. The proposed change revises the CHANNEL FUNCTIONAL TEST Frequency for certain RPS Functions from 31 days to 92 days. Justification for extend~ng the test interval is provided in Combustion Engineering (CE) Topical Report CEN-327-A Supplement 1, "RPS/ESFAS . Extended Test Interval Evaluation," dated January 1989. NRC acceptance of the CE topical report was provided in. a letter to the CE Owner's Group (CEOG) dated November 6, 1989. As discussed in the NRC acceptance letter, Consumers Energy has reviewed the instrument drift information for each instrument channel involved and has determined that drift occurring in that channel over the extended surveillance interval will not cause the setpoint to exceed it's Allowable Value as calculated by our setpoint methodology. Onsite records of the recorded instrument drift are available for audit. Therefore, it is proposed that the CHANNEL FUNCTIONAL TEST be performed in
.accordance with ITS SR 3. 3 .1. 5, at a Frequency of 92 days. This change is consistent with NUREG-1432.
L.3 CTS Table 4.11:1, Footnote (c), requires that the excore channels be calibrated with a test signal every 31 days. ITS SR 3.3.1.6 requires the surveillance to be performed every 92 days. The proposed change revises the calibration Frequency for the excore power range channels from 31 days to 92 days. This test leaves the channel adjusted to account for instrument drift between successive calibrations to ensure that the channel remains OPERABLE between
( calibrations. Other surveillances are performed on the excore power range channels more frequently to account for overall gain and to ensure the upper and lower subchannel amplifiers are calibrated correctly to correspond to the incore detectors. These surveillances are considered sufficient to ensure the excore power range channels are functioning properly. This change is consistent with NUREG-1432. L.4 CTS Table 4.17.1 Footnote (b) requires calibration of the Variable High Power Function with heat balance when power is > 15% RTP. ITS 3.3.1 will add a Note to SR 3.3.1.3 (heat balance) which states that the SR is not required to be performed until 12 hours after THERMAL POWER is > 15% RTP. ~he allowance to delay performance of the SR for 12 hours after power is
> 15% RTP provides time for the plant to achieve stable operating conditions to calibrate the instruments at a power at w4ich the heat balance is accurate. This will provide more accurate results, and thereby providing assurance that the
RPS Functions will actuate at the required setpoints. The 12 hours interval is within the 24 hour delay allowed for a missed SR. This change is consistent with NUREG-1432.
Palisades Nuclear Plant Page 10of11 05/30/99 as-o
R.3 ATTACHMENT 3 DISCUSSION OF CHANGES SECTION 3.3, INSTRUMENTATION CTS 3.17.6, Table 3.17.6, items 8, 9, 10, and 11, and associated Note(a), and Table 4.17.6, items 8, 9, 10, ii, and associated Note(c), contain requirements for primary safety valve position indicator (CTS 3.17.6.8), PORV position indicators (CTS 3.17.6.9), PORV block valve position indicator (CTS 3.17.6.10), and for the service water break detector (CTS 3.17.6.11). These requirements are proposed to be relocated to the Operational Requirements Manual (ORM). These instruments provide indications to the operator in the event of an abnormal condition associated with the specific monitored parameters. These instruments do not provide inputs to safety systems in order for these systems to mitigate Design Basis Accidents (DBAs). These instruments are not required to mitigate any DBAs, nor do they provide input into any system required to mitigate DBAs. These instruments monitors do not meet any criteria in 10 CPR 50.36(c)(2)(ii). Therefore, per 10 CPR 50.36(c)(2)(ii), these Specifications can be relocated out of the Technical Specifications. Any changes to these requirements will be made under the provisions of 10 CPR 50.59. This change is consistent with NUREG-1432.
R.4 CTS Table 3.17.6, item 19, and associated Note(b), requires two fuel *pool area
radiation monitors to be operable at HOT STANDBY condition and above .
CTS 3.17.6.19 requires the plant to stop moving fuel within the fuel pool area and to restore the monitor to OPERABLE status or provide equivalent monitoring capability within 72 hours. CTS Table 4.17.6, item 19, requires periodic surveillances on these monitors. These requirements are being relocated to the Operational Requirements Manual (ORM). These instruments do not provide inputs to safety systems in order for
*these systems to mitigate DBAs. The fuel pool area radiation monitors are not required to mitigate any DBAs, nor do they provide input into any system required tq mitigate DBAs. These radiation monitors do not meet ariy criteria in 10 CPR 50.36(c)(2)(ii).
Therefore, per 10 CPR 50.36(c)(2)(ii), this Specification can be relocated out of the Technical Specifications. Any changes to these requirements will require a 10 CFR 50.59 evaluation. This change is consistent with NUREG-1432 .
Palisades Nuclear Plant Page 2 of 2 d8-o 01120/98
ATTAC1'ENT 3
R.4 DISCUSSION OF CHANGES SECTION 3.3, INSTRUMENTATION CTS Table 3.17.6, item 19, and associated Note(b), requires two fuel pool area radiation*
monitors to be operable at HOT STANDBY condition and above. CTS 3.17.6.19 requires the plant to stop moving fuel wit_hin the fuel pool area and to restore the monitor to OPERABLE status or provide equivalent monitoring capability within 72 hours. CTS Table 4.17.6, item 19, requires periodic surveillances on these monitors. These requirements are being relocated to the Operational Requirements Manual (ORM). These instruments do not provide inputs to safety systems in order for these systems to mitigate DBAs. The fuel pool area radiation monitors are not required to mitigate any DBAs, nor do they provide input into any system required to mitigate DBAs. These radiation monitors do not meet any criteria in 10 CFR 50.36(c)(2)(ii). Therefore, per 10 CFR 50.36(c)(2)(ii), ,this Specification can be relocated out of the Technical Specifications. Any changes to these requirements will require a 10 CFR 50.59 evaluation. This change is consistent with NUREG-1432. R.5 CTS Table 3.17.6, item 12 (the Flux - LiT alarm), the associated Action Statements I .::} (3.17.6.12.1and2), and Surveillance Requirements 4.17.6.12, have been relocated to the I <\) ORM. The Flux - LiT alarm does not provide any inputs to safety systems or initiate any I I automatic actions. This alarm monitors the two input signals to the Variable High Power I .
Trip auctioneer circuit and alarms if these signals differ by more than a pre-determined amount. The Flux - LiT alarm do not meet any criteria in 10 CFR 50.36(c)(2)(ii).
Therefore, per 10 CFR 50.36(c)(2)(ii), this Specification can be relocated out of the Technical 'specifications. Any changes to these requirements will require a 10 CFR 50.59 evaluation. This change is consistent with NUREG-1432 . I rO I c{) I I <:;L I Ck"
Palisades Nuclear Plant Page 2of2 05/30/99 d8-p
CONVERSION TO l,JIPROVED TECHNICAL SPECIFICATIONS RESPONSE TO JANUARY 6~ 1999 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.3, INSTRUMENTATION
NRC REQUEST:
3.3.1-27 CTS Table 3.17.6 DOC L.3 CTS Table 3.17.6 Functions 12 & 20 are modified by footnote (a) eliminating the applicability of Specification 3.0.4. Comment: This change is marked up as if it were included in ITS 3.3.1, Condition D as a Note. ITS 3.3.1 Condition D does not include a note or any provision for eliminating the applicability of Specification 3.0.4. Provide clarification including all applicable documentation for the change. Consumers Energy Response: The markup of CTS Table 3.17.6 has been corrected, and is no longer associated with ITS LCO 3.3.1. See the response to RAI 3.3.1-26. Affected Submittal Pages: See the response to RAI 3.3.1-26 .
29
CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO JANUARY 6, 1999 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.3, INSTRUMENTATION
NRC REQUEST:
3.3.1-28 CTS Table 3.17.6 CTS Table 3.17.6, Function 19 is modified by footnote (b) deleting the applicability of Specifications 3.0.3, 3.0.4, and 4.0.4. Footnote (b) is marked up as a deletion in the CTS Mark-up and two annotations are provided "See 3.3.6" and "See also 3.2". Comment: It is unclear where Function 19 requirements are located in the ITS and where the footnote (b) exceptions are applied. Provide clarification and appropriate documentation for the change. Consumers Energy Response: The markup of CTS Table 3.17.6 has been corrected, and is no longer associated with ITS LCO 3.3.1. See the response to RAI 3.3.1-26. Affected Submittal Pages: See the response to RAI 3. 3.1-26 .
30
CONVERSION TO IMPROVED TECHNICAL ;)PECIFICATIONS RESPONSE TO JANUARY 6, 1999 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.3, INSTRUMENTATION
NRC REQUEST:
3.3.1-29 ITS 3.3.1 Table 3.3.1-1 CTS Table 2.3-1 The Allowable Value for the Variable High Power Trip Function in CTS Table 2.3-1 is stated as a percentage above "core power with a minimum of less than or equal to 30% "Rated Power and a maximum of less than or equal to 106.5% Rated Power. ITS Table 3.3.1-1 changes the stated percentage above "core power, to "RTP", and stated minimum and maximum percentage values in "Rated Thermal Power ,versus the CTS "Rated Power. Comment: No discussion or justification is provided for the changed terms in the ITS. This change also results in a change in STS presentation for the Variable High Power Trip Function. that is not discussed or justified. The change from the STS presentation appears to be unnecessary. Provide discussion and justification for the changed terms, describing the difference between the CTS and ITS terms. Consumers Energy Response: The three presentations of the required trip setting, spaced to accentuate the similarities and differences, are: CTS: ~ 15% above core power, with a minimum of ~ 30% RATED POWER and a maximum of~ 106.5% RATED POWER. ITS: ~ 15% RTP above current THERMAL POWER with a minimum of ~ 30% RTP and a maximum of ~ 106.5% RTP STS: ~[10]% RTP above current THERMAL POWER but not <[30]% RTP nor >[107]% RTP In the ITS, the CTS term "RATED POWER" was replaced with the ITS (and STS) term "RTP" (RATED THERMAL POWER). The CTS term "RATED POWER" and the ITS term "RATED THERMAL POWER (RTP)" are both defined; their meanings are equivalent: CTS: RATED POWER shall be a steady state. reactor core output of 2530 MW1* ITS: RTP shall. be a total reactor core heat transfer rate to the primary coolant of 2530 MWt. The CTS wording "core power, an undefined term taken to mean core power, was changed to "current THERMAL POWER", using the ITS (and STS} defined term THERMAL POWER) with the same intended meaning. The power levels specified in ITS were clarified as being percentages of RTP; in CTS, that was assumed. The remaining differences. between the ITS and STS presentations are due to the retention of CTS terminology familiar to the Palisades staff. The use of a setting requirement presentation using "a minimum of' and "a maximum of' was approved in Amendment 118 to the Facility Operating License. As can be seen in the presentation above, the presentations are equivalent. No further change to the presentation of this requirement is requested.
\
Affected Submittal Pages: No page changes. 31
CONVERSION TO IMF.. ROVED TECHNICAL SPECIFICATIONS RESPONSE TO JANUARY 6, 1999 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.3, INSTRUMENTATION
NRC REQUEST:
3.3.1-30 Comment: ITS Table 3.3.1-1 Function 2, High Startup Rate Allowable Value The High Startup Rate function is proposed for ITS with the ALLOWABLE VALUE set at "not applicable." The ITS also proposes channel check, channel functional test and channel calibration SRs. Provide the plant procedures for performing the channel calibration on this RPS function. Identify how the Palisades setpoint methodology relates to the procedure calibration limits. Consumers Energy Response: Plant procedures for calibration of the High Startup Rate trip function record the as found and final (as left) values for the trip bistable. The nominal setpoint is 2.6 DPM; specified tolerances are:
"As Found" 2.5 to 2.7 DPM "As Left" 2.55 to 2.65 DPM Palisades setpoint methodology selects setting limits which provide assurance that setpoints will not drift beyond their allowable values between calibrations, yet provide sufficient operating I margin so that monitored parameters do not approach trip setpoints or pre-trip alarms during .
I expected operational transients and sufficient setting tolerance so that frequent adjustment is necessary. The high startup rate trip setting limits use this same methodology, with the exception that the 2.6 DPM nominal 'setpoint is not based on any analytical assumptions. That value has been used throughout plant life. Affected Submittal Pages: No page changes .
32
CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO JANUARY 6, 1999 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.3, INSTRUMEhlTATION
NRC REQUEST:
3.3.1-31 NRC REQUEST: 3.3.2-1 Resolved during 10/27/98 meeting; Consumers Energy not required to respond. Resolved during 10/27/98 meeting; Consumers Energy not required to respond. NRC REQUEST: 3.3.3-1 ITS 3.3.3 CTS 3.17.2, 3.17.3 JFD 10 Operational bypasses are deleted from the CTS ESF applicability and operational bypass channel TS are deleted from ITS ESF applicability. The ITS proposes to include actions (Required Action C.1) for the SIRWT function which require placing the inoperable SIRWT channel in bypass within 8 hours. JFD 10 states that most ESF functions do not have bypass capability. It would appear that the SIRWT channels can be bypassed, yet CTS Function 2.c (T3.17.2) shows that this function has no operational bypass. Comment: Explain the difference between the CTS operational bypass and channel bypass. Explain the use of the bypass corrective action in ITS Action C.1. Provide a revised LCO Applicability which includes a requirement to have bypass channels operable or operational bypasses operable for those ESF functions that have bypass TS requirements in CTS 3.17.2 and CTS 3.17.3 . Consumers Energy Response: The term "operational bypass" as used in CTS LCOs 3.17.2 and 3.17.3 refers to circuitry that is routinely used during normal plant startup and shutdown to prevent occurance of automatic features which, although required during actual plant operation, are not desired during all startup and-shutdown conditions. Examples are the bypass circuitry which can be manually enabled, when the plant is below 1700 psia, to prevent SIS actuation on low pressurizer pressure or can be manually enabled, when the plant is below 565 psia, to prevent actuation of the low steam generator isolations. The term "channel bypass" is only used as part of the RPS term "trip channel bypass". It does not properly apply to ESF Functions or instrumentation. The correct use of these terms is explained, in detail, in the Background section of the bases for LCO 3.3.1. The CTS 3.17.2.4 and ITS 3.3.3 8.1 Action requiring the SIRWT low level (RAS actuation) channels to be bypassed would be accomplished by circuit modification (i.e., jumper installation). There are no built in bypass capabilities for defeating these circuits. ITS LCO 3.3.3 has been revised to address the portions of the bypass removal circuitry that is part of the instrument channels. However, since at Palisades the bypass it-self is affected by opening contacts in the logic circuitry and not in ifldividual instrument channels, as is implied in the STS, the revised wording still differs from the STS .
33 (continued)
CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO JANUARY 6, 1999 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.3, INSTRUMENTATION NRC REQUEST: 3.3.3-1 Consumers Enerqv Response: (continued) The following changes were made:
1. Bypass Removal Functions v.;ere added to Table 3.3.3-1; with their Applicabilities, Allowable Values, and SRs specified in that table. The bypass removal requirements were added to the table, rather than to the LCO wording because in the STS, the footnotes providing the bypass removal setpoints are presented in the ESFAS Instrumentation LCO Table. In the ITS, the equivalent footnotes are presented in the ESF Logic LCO Table, therefore simply adding a reference to bypass removal channels in
.the LCO (as is done in the STS) would not result in the required setpoints being provided in the LCO. This treatment is consistent with the treatment of bypass removal functions in ITS LCO 3.3.1. "Bypass removal" functions are not addressed in 3.3.4 Condition C, because Functions 5 and 6 do not have any associated bypass removal function.
2. The Function titles in Table 3.3.3-1 were editorially revised to more clearly address actuation bistables (or functions) and bypass removal bistables (or functions).

3. The order in which the functions are listed in ITS Tables 3.3.3-1 and 3.3.4-1 was revised.
so one shutdown Condition addresses Functions 1, 2, 3, and 4 and the following Condition addresses 5 and 6. This also groups actuation functions with similar applicabilities together within the table.
4.
5. 6. The headings of Table 3.3.3-1 have been editorially revised to be consistent with Tables 3.3.1-1 and 3.3.4-1. Conforming changes were made to the Bases. JFD 10 has been revised to more clearly explain the ITS/STS differences. (continued)
34
CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO JANUARY 6, 1999 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.3, INSTRUMENTATION
NRC REQUEST: 3.3.3-1 Consumers Energy Response: (continued)
Affected Submittal Pages: Att 1, ITS 3.3.3, page 3.3.3-1 Att 1, ITS 3.3.3, page 3.3.3-2 Att 1, ITS 3.3.3, page 3.3.3-3 Att 1, ITS 3.3.3, page 3.3.3-3 insert Att 1, ITS 3.3.4, page 3.3.4-1 Att 1, ITS 3.3.4, page 3.3.4-2 Att 3, CTS, page 3-63 Att 3, CTS, page 3-67 Att 3, CTS, page 3-69 Att 3, CTS, page 4-77 Att 3, CTS, page 4-78 Att 5, NUREG, page 3.3-18 Att 5, NUREG, page 3.3-19 Att 5, NUREG, page 3.3-19 insert Att 5, NUREG, page 3.3-20 Att 5, NUREG, page 3.3-20 insert Att 5, NUREG, page 3.3-22 Att 5, NUREG, page 3.3-22 insert (2 pages) Att 5, NUREG, page 3.3-23 Att 5, NUREG, page 3.3-24 Att 5, NUREG, page 3.3-25 Att 6, JFD 3.3.4, page 2 of 2 (former submittal) Att 6, JFD 3.3.4, page 2 of 2 *
35
ESF Instrumentation
3.3 INSTRUMENTATION 3.3.3 Engineered Safety Features (ESF) Instrumentation 3.3.3 LCO 3.3.3 Four ESF bistables and issociated instrument channels for each Function in Table 3.3.3-1 shall be OPERABLE.
APPLICABILITY: As specified in Table 3.3.3-1. ACTIONS
-----~-------------------------------NOTE-------------------------------------
Separate Condition entry is allowed for each Function. CONDITION REQUIRED ACTION COMPLETION TIME Place affected 7 days A. --------NOTE---------- A. l bistable in trip. 1l9mrl@ One or more Functions with one ESF bistable or associated instrument channel inoperable. B. -------------NOTE------------ LCO 3.0.4 is not applicable. B.1 Place one bistable in 8 hours trip. One or more Functions AND A.SSOCIA.'TC-D with two ESF bistables
-or_a_s_s_o_c_i-ated------- Ar:m -'-~-'f>:rJ~~~~-'-1-f~_e_ -
instrument channels B.2 Restore one bistable 7 days inoperable . to OPERABLE status.
Palisades Nuclear Plant 3.3.3-1 Amendment No. 01/20/98
~5-o-
ESF Instrumentation
ACTIONS CONDITION REQUIRED ACTION 3.3.3 COMPLETION TIME
c. c.1 8 hours c anne ino erable. t(A1:.
B\S\A.'6L'E' AND o ~ ASS"oc:. I A.TE-I:> 3,3,3-1 1"-ISTll:!UN\E-)J\ C.2 Restore channel to 7 days OPERABLE status. BISTA.BLE: AJJt:> ASX>c.IAT'e'P 11.JST~ t:AJI D. Required Action and D.1 Be in MODE 3. 6 hours associated Completion Time not met for AND
~nction@, ~or . 2, 3, D.2 Be in MODE 4. 30 hours
E. Required Action and associated Completion Time not met for Functions~or~
E.1 AND E.2 Be in MODE 3. Be in MODE 5. 6 hours 36 hours SURVEILLANCE REQUIREMENTS
-------------------------------------NOTE-------------------------------------
Refer to Table 3.3.3-1 to determine which SR shall be performed for each Function. SURVEILLANCE FREQUENCY SR 3.3.3.1 Perform a CHANNEL CHECK. 12 hours SR 3.3.3.2 Perform a CHANNEL FUNCTIONAL TEST. 92 days
\\
SR 3.3.3.3 Perform a CHANNEL CALIBRATION.
Palisades Nuclear Plant 3.3.3-2 18 months Amendment No. 01/20/98 35-b
ESF Instrumentation 3.3.3 Table 3.3.3-1 (page 1 of 1) Engineered Safety Features Instrumentation (A.PPl..\LASLE-) SURVEILLANCE Al.LOWAILE FUNCTION MOUES ltfQU!REMENTS VALUE
l. Safety Injection Signal (SIS)

a. Pressurizer Low Pressure 1,2,3 SR 3. 3. 3.1 , 1593 psh SR 3.J.J.2 SR J.J.J.3 Containment High Pressure~(CHP)
~ a. Contaf111Mtnt High Pressure - Left Train 1,2,3,4 SR 3.3.3.2 , J . 7 psi g and SR 3.3.J.J s 4~sig
b. Cont1in11ent High Pressure - Right Train 1,2,3,4 SR 3.J.3.2 2 J. 7 ps f g and SR J.J.3.3 ~ 4~sf g c;J Cont41nment High Radiation Signal (CHR)
&. Containment High Radiation 1,2,3,4 SR 3.J.J.l SR J.3.3.2 s 20 R/hour SR 3.3.3.3 RAI
St*** 61nerator Low Pressure Signal (SGLP)
a. St*** 61n1rator A Low Pressure 1, 2<*1' 3<11 SR 3.3.3.l SR 3.J.3.2 SR 3.3.3.3 2 500 psh 3.3.~-I

b. Ste .. 61nerator B Low Pressure SR 3.3.3.1 ~ 500 psh SR J.3.3.2 SR 3.J.3.3 Recirculation Actuation Signal (RAS)

a. SIRWT Low Level 1,2,3 SR 3.3.J.J ~ 21 inches and s 27 inches above hnk bottom Auxflf1ry F11dw1t1r Actuation Signal (AFAS)

a. Stea* 61n1r1tor A Low Level 1.2,3 SR 3.J.J.l ~ 25.9% narrow SR J.J.3.2 range SR 3.J.J.3

b. Ste.. 61ner1tor B Low Level l,2,3 SR J.J.3.1 2 25.9% narrow
_____________________________________ SL3_._~_,_3_4 ____ ran_g!_ __________ _ -- -------(])-(ADD lt..1'5E:"iZ.-r 1) SR 3.3.3.3
\I (a) Hot required to be OPERABLE when all Main Stea* Isol1tion Valves (MSIVs) are closed and de1ctfv1t1d, 1nd all Hain F11dw1t1r Regulating Valves (HFRVs) and HFRV bypass v1lv1s ire either closed and deactivated, or isolated by closed .. nual valves.
Palisades Nuclear Plant 3.3.3-3 Amendment No. -01/20/98 35-(_
7. Automatic Bypass Removals:
a.. Pressur1 zer* Low Pressure Bypas.s 1.2 .. 3 SR 3.3 .. 3.3 . > 1700_ psi a b.. Steam. Genera tor A tow SR 3.3~3.3 > 565 psi a Press.ure Bypass SR 3.3.3.3 >565 psia c.. Steam Generator B Low Press.ure Bypass
\ \
ESF Logic and Manual Initiation
3.3 INSTRUMENTATION 3.3.4 Engineered Safety Features (ESF) Logic and Manual Initiation 3.3.4 LCO 3.3.4 Two ESF Manual Initiation and two ESF Actuation Logic channels and associated bypass removal channels shall be OPERABLE for each ESF Function specified in Table 3.3.4-1.
APPLICABILITY: According to Table 3.3.4-1. ACTIONS
-------------------------------------NOTE-------------------------------------
Separate Condition entry is allowed for each Function. CONDITION REQUIRED ACT ION COMPLETION TIME
A. One or more Functions with one Manual Initiation, Bypass Removal, or Actuation Logic channel inoperable.
A.1 Restore channel to OPERABLE status. 48 hours B. One or more Functions B.1 Be in MODE 3. 6 hours with two Manual Initiation, Bypass AND Removal, or Actuation Logic channels B.2 Be in MODE 4. 30 hours or . i!i i noperab 1e for ~ Fu~i ans 1, Z 3 1-i OR
~l. --- -Required--Ac-tion and- -- --- -- - - - - - - --- - - - - ---- - - - --~-----
5,3.-3--1-- -- associated Completion Time of Condition A I no~et for ~nctions 1, , ~or 2 3 Palisades Nuclear Plant 3.3.4-1 Amendment No. 01/20/98 35- e...
ESF Logic and Manual Initiation
ACTIONS CONDITION REQUIRED ACTION 3.3.4 COMPLETION TIME C. One or more Funct i ans C.1 Be in MODE 3. 6 hours with two Manual Initiation,-li14 ___il4....*l
....Qi AliD.
l8'm4v#. or Actuation Logic channels C.2 Be in MODE 5. 36 hours inoperable for Functions &or~ QR Required Action and associated Completion Time of Condition A not ~ for Functions
~or .
(2) ft> - - - -- - ~ - - -- -- - -- - - - - - -----------
Amendment No. 01/20/98 Palisades Nuclear Plant 3.3.4-2 3S-.f'
3. 16 ENG NE RED SAF TY FE 3.16 he Engineered S~fety eatures (ESF) syste instrumentati l 'mi-ts shall be as sta d in Table 3.16.
Specif ation 3.16 is applic le when associated F or Function instrumentation is r uired to be OPERABLE Specifica ion 3.17.2 or 3.17.3. 3.16.1 I an ESF instrumen setting is not with1 the allowable settin Ta e 3.16, innediate declare the instru nt inoperable and co corr tive action as d ected by specificat1 n 3.17.
~
TABLE~ Engineered Safetv Features Svstem Instrument Settings
n.
- :;l...1_,
l5.a.,5.b] 1> . .., 1. 2. 3. Instrument Channel Pressurizer Low Pressure Containment High Pressure Containnient High Radiation Allowable Value
~ 1593 psia 3.70 - 4. ~
s 20 R/h
~
ig M.
~- ..... j ~Al: ...., 3:s:s-& 'Z.c... ,2.b_j 4. Steam Generator Low Pressure . ~ 500 psia \'-\ .a.,4.~ !5. Steam Generator Low Level ~ 25.~
Narrow Range
,..-:_. I (J.O...J 6. SIRW Tank Low Level 21 - 27 inches Above Tank Bottom
7. E~red afeguards P~o~ ~ s2~
Ven lation h Radiatio -~---- -~-----------------------* -- Amendment No. 89, 162 October 26, 1994 3-63
lf'S" fr3.3.3-I]
3.17 . INSTRUMENTATION SYSTEMS Table~~
~
Instrymentat1on Ooerat1ng Begy1rements for Engineered Safety Features 4m11"' '\ ~
~~!*~'*!~*
(Mj Funct1 on al Unit t=r s~!~*~ Safety InJect1on Signal (SIS) 1 1
c. l Q.0..1 d. Pressurizer Pressure Q.cot.3.3] 4 Instrument Channels

2. Recirculation Actuation Signal (RAS)
(?. a-J c.
~~:u~itiation ~
SI RWT Level Switches fu,oJ,'3.3j 4 Non~ Non
3. Auxiliary Feedwater Actuation Signal (AFAS) a.
b.
~. ei...] c. "A" Steam Generator Level t,l. b] d. "B" Steam Generator Level Amendment No. ~
3-67 35-~
3. 17 INSTRUMENTATION SYSTEMS ~ ~
nJ:J,;-J)
- il:S
(!"3.3'. 3 - 0 Table' Instrumeatat1ga Oper1t1ng Begy1rements fgr lsglatjon Fynct1oas Functional Un1t
1. ~gct1jcm~ct ~jgb ec1ssYtl (CHP)

a. (CHP l~c Trains ~ 1 \' None:r-@
cs a..J
~
b. Containment Pressure ~ ~
Switches - Left Train
~.b] c. Containment Pressure J["*'-'J ~ ~.~.'3-1 Switches - Right Train
z. ~2ct1icm~at Higb B1~i1tiga (CHR)

a. Man 1 Initiation 1 b.
f!>, "-] c. Containment Area Radiation Monitors
3. Steam Geaer1tgr Lgw PressYtl (SGLP) 4 (1c.o 3.3.;']
~ ~* ~~,1
a. Manu Actuation 1 set

b. 1

c. "A" Steam Generator Pressure

d. "B" Steam Generator Pressure 0 No~
~ West Room nitor 1 0 None.
Amendment No. 162 October 26, 1994
35 -_A_
3-69
4.17 INSTRUMENTATION SYSTEMS TESTS Table~
Instrumentation Syrye1
~ *Engineered Safetv Featyres (5(,.J.3', D; Reay1rements for CHANN~,J,!.1J G1t.!.l.S'.:3]
CHANNEL, FUNCTIONAL CHANNEL Functional Un1t _ CHECK TEST CALIBRATION
1. Safety InJect1on Signal (SIS}

a. Initiation NA

b. SIS gic (Init tion, Actuation, and low pre sure block auto reset

c. 18 months NA Pressurizer Pressure Instrument Channels 12 hours ~ days 18 months

2. Recirculation Actuation Signal (RAS}
a.
~
18 months NA
b. 18 months NA 11-'9 c. 3:,,,_,
K'Al:
~
SIRWT Level Switches NA 18 months
3. Auxiliarl Feedwater Actuati~n Signal (AFAS}

a. [:anu~nitiation

b. AFAS L ic ~
18 months 92 da;'.S ~ }@ f9 ,a.] c. "A" SG Level 12 hours~ days 18 months ~1'."I.. 3,3.~-1 B bJ d. "B" SG Level 12 hours ~z; days 18 months L. a} Test n3!al and emergency power unctions using test circuit~each 92 days. Verify a l automatic actuations d automatic resetting of lo~pressure bloc ___ e.ach_l8_m nths-.---- - - - - - - ---.--- - - -
4-77 Amendment No. ~. '* H-1,
\T'S LJ3,:J,3-1]
4.17 IHSTRUt1E!fIATION SYSTEMS TESTS Ttbll~ -
~
Instryrnentat1on Surye111ance Beay1rernents for
~
Isoltt1on Eynct1ons ~ n ;' ~l!S,J,j,~ ~ M
~f.U,3,L CHANNEL ,r-1-~t..J;J:l$.J \CHANNEL FUNCTIONAL CHANNEL fynct1onal Un1t CHECK TEST CALIBRATION
1. ~gctliDIDIDt Hfgb fCISSYCI (CHP) 1.(£HP ~gic Trains ~ 18 months \ NA ~ -'*:!.4-
~.a-] b. Contain.. nt Pressure Switches
Left Train NA
~days 18 1110nths ~AI: ~.'33-1
(}". b] c. Conta1nnient Pressure NA days 18 months Switches
Right Train
2. ~gct1icamct Hfgb B1df 1tf gc (CHR)

a. Manual~tttatton MA 18 llOnt NA
/fa.a.] I
b. CHR Lo i Trains

c. ContainllM!nt Area Radiation Monitors 12 hours 6
18 1110nths
~days NA 18 months ~
3,3.1-1
3. Stlll ~IDICltQC La* ECISSYCI (SGLP)

a. Hanua Actuation 18 1110nths NA

b. SGLP Log c Trains 18 110nths NA (1.~ I c. *A* Stea11 Gtn*rator 12 hours ~ days 18 iaonths Pressun L."l eta.!
153-1
~.b]I d. *s* St1111 Gtnerator 12 hours ll days 18 11anths Pressure 110nths
35-4 4-78 Allendllent No. ~. ~. 171
. ~../ ./) ,
3.3 INSTRUMENTATION ESF~ Instrumentation 3.3.if{ ~nqinee~ed Safety Features(ActU::ati'btl Sy'ijilii) (ESP~) Instrumentation
)~ \!Aft4l'o9t~'0 LCO 3.3.~ Four ESF@trfi;Ou£i~a~~~~ated <(jiii(iiib channels for each Function in Table 3.3 instrument ss 1 shall be OPERABLE. ~
APPLICABILITY: ACTIONS
-------------------------------------NOTE-------------------------------------
Separate Condition entry is allowed for each!~ U?.ft) ot;D'Ypa§S:'!e~ ~ Function. CONDITION REQUIRED ACTION COMPLETION TIME
A_.1
~tin~
Place 'affected trip l31STA l 3. 11 *"\.. :z.J ~.
~1 One or more Funct 1ons Place affected~
with one ES~
'.:*'*-_;~or ,J..;; ,1 :: .,-1 ~ in 'Q:a'S$ 'O"'
assorut trip.
- ' " instrument channel '[hijp~D inopera:
B. 2. f"'. Restore channe1 to OP,E_~LE stltf.l~*
'~ '*.,"" ',,""'
2.2""'flace af'fected trip un_it in trip.
\ \
CEOG STS 3.3-18 Rev 1, 04/07/95 35-/
fe (({{( rJ/{_1<-r)
ACTIONS continued CONDITION REQUIRED ACTION ES~ Instrumentation <(Anah)i]
COMPLETION TIME 3.3.~
-- -- /JOT~ - - - -
Not- CJ..prltt:t.-.b le. +o R~S. Mil
-------------i--------------- --*--*-------*--*-...
D. l Di sab1e bypass \ 1 hour .\ channel. \
'\'\
0.2.1
"""'\'"Ma "\
Place affe~ted trip units in byp~ss or trip. *."**,, 1 ho~ i i. i D. 2. 2.i\Restore bypass [48] hours removal channel and aff!!cted trip units \"
\ , to OPERABLE status.
r~~ace~~ed L'"'~its in tri}>-., trip 48 hours (continued) CEOG STS 3.3-19 Rev 1, 04/07/95 3 5-fYJ
SECTION 3.3
CONDITION INSERT 1 REQUIRED ACTION COMPLETION TIME I
C
One RAS bistable or associated C. l Bypass affected 8 hours instrument channel inoperable.
Bistable. MID C.2 Restore bistable and associated instrument 7 days channel to OPERABLE status .
35-n 3.3-19
ESF~ Instrumentation ~ ,....._
~ ~ ACTIONS continued CONDITION REQUIRED ACTION COMPLETION TIME e or more Fun ons ------------NOTE---~~--------
w two automatic LCO 3.0.4 is not appltcable. bypa removal ------------------------ channe inoperable. Disable bypass 1 hour channels. E.2.1 Plac ,one affected 1 hour trip u~ t in bypass and plac the other
~ in trip fo each \ affected ESF ~Function.
Alil2 \. E. 2. 2 Resto.re one bypass ~hours. channel. and the associated trip unit to OPERABC~st~tus for each a!!~ed trip Function. [3,n.7...s] ~Required Action and associated Completion Be in MODE 3. 6 hours Time not met
~ fu~c...+1*01'\S I 2 1 '3,Y oe1 Be in MODE 4.
CEOG STS 3.3-20 Rev 1, 04/07/95 35-o
SECTION 3.3
CONDITION INSERT 1 I COMPLETION TIME E. Required Action and .t.l He in MODE 3. 6 hours associated Completion Time not met for Functions 5 or 6.
E.2 Be in MODE 5. 36 hours
35-p 3.3-20

E,..fnnred S.ftty Tlblt J,J,4*1 C~of 2)
'*tl..r'9S Acti.mti 1)1t*
ESF~ Instrumentation *~ lrwt~nfan \
~-~ ~
SIJtV( I LL.Ult! Al.LCWAaLE
\ \
Qt:S IEQUl~llTS VALUE
~ \
1. S.fety Inject fan Actl.llt!an II rel CllAI)
'X*I} ** CGM1f~ l'rtuure- NI .. ,1 ,Z,J $1.
SI J.J.4.1
*J.J.4.2
s [19.0] psf1 /
SI SI 3.J.4.4 J.J'.4.5 0
b. ,.,......! zer l'resaure - La.1< 1 > 1,Z,J SI J.J.4.t, sa J.3.4.2 t [16'57] s-!1 ,
SI 3.J.4.3 Sii 3.J.4.4 SI J.J.4.5
z. Conuf,_,t ,,....,, Acti.mtfan Sll"ll
** CGM1l,_,t l'reuure - Nllft 1,2,l Sii J.J.4., s [19.0] s-11 Sii 3.J.4.2 SI 3.J.4.4 II J.J.4.5 J. Cant1I~ lloleclan Miultfon ll1Nl ** CGntlf~ ,.,........ -11111t 1,Z,J Sii 3.J.4. 1 s [19.0] s-11 R- J.J.4.Z SI 3.J.4.4 \ SI l.3.4.5
[ b. CGM1f~ lldlttlan - Mlllt
\
J 1,Z,J SI SI J.J .... 1 J.l.4.2 SI J.J.4.4 '-. Sii J.J.4.5
4. lllefn I t . . ltoletlan lflNl 1, I t - &ener1tor ,.,.._.. - Lw<c:> 1,z<d> ,l<d> SI 3.J.4.1 t (495] Piii Sii J.J.4.2 SI J.J.4.J Sii J.J.4.4
\ Sii J.J.4.5 \ \ \
1,2,J '
[SI 3.J.4. 1] B 24 fndln era SI 3.3.4.2 s 301 Ind!*
SI 3.J.4.4 ~t. . llCltt* SI J.J.4.5
\ \ *, \ \
(c:antf ""9d) i
~
1z.,. ,.......,. - L* ..., be --.-1\y ~ lil9'1 ~fzer ~ ll be 111tamtlcally rmMlll ~ .......,1ur """'°" h c [1800'] psf1. fa ~C1SOOl psf1.
\
Tiie 1 CCb) SIAI !1 11 r..,fr.d 11 1 Plf'9flllw co lnt'tl,~tl cant*I~ IPf'IY*l \
\ (c:) ltl* G9wr1tcii'~r..-.-Lw _,be mrwlly ~ ~~-n~~-t~ -~*-~-c J7a51 pi1f1_. ___ i
- - - - - - -- - - -- -Th* ~-11111-1--.-ana111ii-can-}'.....,. liil~ifi* 19Nf'ltor preN1re 1*\t C785J psl1. I "- . '
; Cd) Only me lea!n Ste* tlolltfan Sl1Nl (llSIS) fll'Ctlan *erct tlle S t - ~r1tor ,,......,... - Le.. end / Canui,_,t l'r-e-Rllh 1!rel1 are not r~ired to* be QllOAIU ~ell auocftted velws faol1ud \ by me llSIS f\l'Ctfan ire cloeed Ind [drectfvetllQ. ___./ --~--*
e CEOG STS 3.3-22 Rev l, 04/07/95
SECTION 3.3
INSERT 1 Table 3.3.3-1 (page 1 of 2)
Engineered Safety Features Instrumentation APPLICABLE SURVEILLANCE ALLOWABLE FUNCTION MODES REQUIREMENTS V.1\LUE
1. Safety Injection Signal (SIS)

a. Pressurizer Low Pressure 1,2,3 SR 3.3.3.l ~ 1593 psia SR 3.3.3.2 SR 3.3.3.3

2. Steam Generator Low Pressure Signal (SGLP)

a. Steam Generator A Low SR 3.3.3.1 ~ 500 psia Pressure SR 3.3.3.2 SR 3.3.3.3

b. Steam Generator B Low SR 3.3.3.1 ~ 500 psia Pressure SR 3.3.3.2 SR 3.3.3.3
3. Recirculation Actuation Signal
* (RAS)
a. SIRWT Low Level 1,2,3 SR 3.3.3.3 ~ 21 inches and

s; 27 inches above tank bottom

4. Auxiliary Feedwater Actuation Signal (AFAS)

a. Steam Generator A Low Level 1,2,3 SR 3.3.3.1 ~ 25.9%
SR 3.3.3.2 narrow range SR 3.3.3.3
b. Steam Generator B Low Level 1,2,3 SR 3.3.3.1 ~ 25.9%
SR 3.3.3.2 narrow range SR 3.3.3.3
===::::;:================================================-===-==*=============== --~-
(a) Not required to be OPERABLE when all Main Steam Isolation Valves (MSIVs) are closed and deactivated, and all Main Feedwater Regulating Valves (MFRVs) and MFRV bypass valves are either closed and deactivated, or isolated by closed
manual valves .
35-s 3.3-22
SECTION 3.3
INSERT 1 (continued)
Table 3.3.3-1 (page 2 of 2) Engineered Safet~ Features Instrumentation APPLICABLE SURVEILLANCE ALLOWABLE R'At.
'3.~."'3-t FUNCTION MODES DC" o*U.lr\
l'.L IT n Ea1r r*TC-l'C:.11.J VALUE
5. Containment High Pressure (CHP)

a. Containment High Pressure 1,2,3,4 SR 3.3.3.2 ~ 3.7 psig
- Left Train SR 3.3.3.3 and
; 4. 34- psi g

b. Containment High Pressure 1,2,3,4 SR 3.3.3.2 ~ 3.7 psig
- Right Train SR 3.3.3.3 and
; 4.34- psig

6. Containment High Radiation Signal (CHR)

a. Containment High Radiation 1,2,3,4 SR 3.3.3.l : ; 20 R/hour SR 3.3.3.2 SR 3.3.3.3

7. Automatic Bypass Removals

a. Pressurizer Low Pressure 1,2,3 SR 3.3.3.3 > 1700 psia Bypass 1,2Ca),3(a) SR 3.3.3.3 ) 565 psia

b. Steam Generator A Low Pressure Bypass l,2(a) ,3Ca) SR 3.3.3.3 >565 psi a

c. Steam Generator B Low Pressure Bypass (a) Not required to be OPERABLE when all Main Steam Isolation Valves (MSIVs) are closed and deactivated, and all Main Feedwater Regulating Valves
--- ------ - - ---(MFRVs)--and-MrRV-bypa_s_s___ va-1-ve-s a-re-eTthe-r-cTos-ed- and-deactivated,-or___ ---- ----- - isolated by closed manual valves.
\ \
35-;t-3.3-22

ES~ Instrumentation <(k.ihii'£i Telllt 3.3.4*1 Cs:i-te Z of Zl *--*-------------
E"1lrwerld leftt'f futur" Acn.tton Sywt* lrwtnaw1t1tlon
~
3.3.~-.2..J SUIVI I LI.Alla ALLOWAaLl UCIJllU.llTS VALUE 6, Aullltery , . . . .,.,. Actl.9tlon lllNI (A,AI) .. ~--
** u . . ...,_raior A L...,.,,_ Low 1',Z,l SI l.l.4.1 t [45.n I SI :S.l.4.Z SI'. l.l.4.4 'Y I b .....st- Generator I L...,.I - Low 1,Z,l .. :S.l.4.5
l.:S.4.1 t [45.n I 0
SI l .l .4 .l-.. *' I SI SI l.l.4,4 . ',
].l.4.5 "-,. )
c, u - """*~rnor ,,...~ Differ-* .. Migl! f",Z,:S SI Sit l.l.4.1 l.l.4.Z S ~.lJ paid
/ CA > I> or ct..!' Al Sit 3.3.4.4 ~
SI J.l.4.5
~---------------~-------------------------------------
CEOG STS 3.3-23 35-LJ Rev 1, 04/07/95
~
3.3 INSTRUMENTATION ESF@ Logic and Manual @ ~-G)
<9cw ~~~~ -
(111< f"-~es) Logic and Kinui1 Two ESFQt Manual and two ES~Actuation Logic channels shall be OPERABLE for each ESF~ Function specified in Table 3.3 1. cuid tlf.roc.ia f.u{ h'(f'A5.J re11-1.,va.1 APPLICABILITY: According to Table 3.3. 1. t-'1~HH~/J ACTIONS
-------------------------------------NOTE-------------------------------------
Separate Condition entry is allowed for each Function.
-------------------------------------------------------------------~----------
CONDITION REQUIRED ACTION COMPLETION TIME
* [3.17. LU \).11.3. Q IYI r'ha...ho"'>
A. A.I Restore channel to OPERABLE status .. One o c- 11-1.ot-e. Fv.."-t.-t:rM~ wl+~
+wo "'10. ... w:.\ :Lv\i+-~c..~;oiA> Erf'.
Re.,.,.. o v~.1 1 ot- Acf &<AL t; o->< Lo!l
48 hours S.5 B1p11.~S R_Q.""1o-J;l.(J ch~"'"'t!Ls 1-'.oper,J..\a.. +or Fv.."-d;o>\S 11?. ,~ 1 or~. RA,!:.
'5,-;.~- \
I
~
[!. 11. z .sJ B. Required Action and associated Completion B.1 Be in MODE 3. 6 hours Time of Condition A !fill
~s--0 not met er B.2 Be in HOOE 4. R~!. --to~ Func::...+io~s 1i2) '3 J bt' '-1. '3:'S."3 - I ~ne o~e Functio s C.l R~~ channe~
with one anual Trip OPE LE status.
~tuati~gic
- - --- - --- ----- --~xcept~ i_n~~e- ~- - - -- ----- ----- --- --~- - - --- ~ -- - (continued)
CEOG STS 3.3-24 35'-v Rev 1, 04/07/95
llrtt or Merl. f:".._ ... l t;"""~
~ t'lo.~~1 :t. ... ~+: .... +~r.A, or ~,+IU\.1:'"" Lo~;t. ~. . -0 RA't ~hi:.i-lo; J.5.'~*I t.l.o.\\"tl~ \v..c~~r;1..~\..a.. f.er:-
F. . .~t."'(~s. 5 O'" lD ESF~Log1c and Manual~~ _ I
-~
C...TS REQUIRED ACTION COMPLETION TIME CJ, 17,J, s-] ~Required Action associated Completion and 9 Be 1n MODE 3. 6 hours Time of Condition~ AHD W ~
©-
not met'
~ ~1.4c._{10~ 5 Or' g Be in MODE 5. 36 hours J<AI. ~.3.!-1 I
SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY
~
SR Perform a CHANNEL FUNCTIONAL TEST on each q{gz@ days -0 _{i) AS logic channel. ~~Ar A r.1,t.j .,
~lE@months~ © ---------------~*-~.~~~~~~--~~~-------i--~~-------
5R :S.3.4,1 Per{on'w\ ~1..1.."-~..\-;cna..\ +"s1: o~ t::-1>.!.I-\ $1S 1:rz, d.o..vs e~l:. ACIUATlot-l C~t.JN'i:-L Uo~M~L
/o..>-.\D ~TA~t:>6Y '?OWE-~ FulJC.ltONS. 5.~.'{-'Z. --~~-~~1~~------1~~~------ i ----------**--*-*-----------l-------- I
CEOG STS 3.3-25 Rev 1. 04/07/95

Change ATTAC1'ENT 6 JUSTIFICATION FOR DEVIATIONS SPECIFICATION 3.3.4, ESFAS INSTRUMENTATION Discussion
9. The Applicability has been revised since certain Functions are applicable in MODES 1, 2, and 3, while others are applicable in MODES 1, 2, 3, and 4. The Applicability has been changed to "As specified in Table 3.3.3-1," and a "MODES" column has been included in Table 3. 3. 3-1. The proposed Applicability is consistent with the current licensing basis.
10 10.
11. ITS Table 3.3.3-1 identifies the applicable SRs for each ESP Instrumentation Function.
The appropriate generic SR Note is included to address this format and the wording of the SRs has also been modified to reflect this change.
12. ISTS SR 3.3.4.5, the ESP RESPONSE TIME test, is not included in the ITS. This test is not within the current licensing basis since the conclusions of NUREG-0820, "Integrated Plant Safety Assessment Systematic Program, Palisades Plant" - Final Report USNRC 10/82, concluded that backfitting this testing into the current program was not required.

13. The specific wording which discusses other plants is being deleted. The Bases are specific to Palisades and will contain, where possible, only Palisades specific information.

14. The text in the Bases concerning the justification for surveillance frequencies with the use of topical reports has been deleted since it was not intended to be included in the ITS .
Palisades Nuclear Plant Page 2of2 35-x 01/20/98
ATTACHMENT 6
Change 10.
Discussion JUSTIFICATION FOR DEVIATIONS SPECIFICATION 3.3.4, ESFAS INSTRUMENTATION Bypass removal channels are addressed in both LCO 3.3.3 and LCO 3.3.4; the bistabies and instrument channels in 3.3.3, and the Logic channels in 3.3.4. The bypass removal bistables and their instrument channels are subject to a Channel Functional Test as part of the Channel Calibration required by ITS SR 3.3.3.3; the bypass removal Logic channels are subject to a Channel Functional Test by ITS SR 3.3.4.3. and most ESF Functions do not have "trip units". Due to these design
*differences:
The LCO wording "trip units and associated instrument and bypass removal channels," has been changed to "bistables and associated instrument channels." The related ISTS SR 3.3.4.3, i.e., the CHANNEL FUNCTIONAL TEST on each automatic bypass removal function, is not included in the ITS. The Completion Time for placing a channel in trip or bypass has been changed to 8 hours, consistent with the current licensing basis/Technical Specifications, as approved in Amendment 162 of the Palisades operating license .
11. ITS Table 3.3.3-1 identifies the applicable SRs for each ESP Instrumentation Function.
The appropriate generic SR Note is included to address this format and the wording of the SRs has also been modified to reflect this change.
12. ISTS SR 3.3.4.5, the ESP RESPONSE TIME test, is not included in the ITS. This test is not within the current licensing basis since the conclusions of NUREG-0820, "Integrated Plant Safety Assessment Systematic Program, Palisades Plant" - Final Report USNRC 10/82, concluded that backfitting this testing into the current program was not required.*

13. The specific wording which discusses other plants is being deleted. The Bases are specific to Palisades and will contain, where possible, only Palisades specific information.

14. The text in the Bases concerning the justification for surveillance frequencies with the use of topical reports has been deleted since it was not intended to be included in the
--~--~I=TS=. -~-----------
Palisades Nuclear Plant Page 2 of 2 05/30/99 35-Lj
CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO JANUARY 6, 1999 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.3, INSTRUMENTATION
NRC REQUEST:
3.3.3-2 NRC REQUEST: 3.3.3-3 Resolved during 10/27/98 meeting; Consumers Energy not required to respond. Resolved during 10/27/98 meeting; no Consumers Energy response required. NRC REQUEST: 3.3.3-4 Resolved during 10/27/98 meeting; no Consumers Energy response required. NRC REQUEST: 3.3.3-5 Resolved during 10/27/98 meeting; Consumers Energy not required to respond. NRC REQUEST: 3.3.3-6 ITS 3.3.3 DOC none CTS Table 3.17.2, function 1.d The CTS requires four pressurizer pressure instrument channels to be operable with an operational bypass at less than or equal to 1700 psia PCS pressure. This requirement is translated as Pressurizer Low Pressure in ITS Table 3.3.3-1. The FSAR contains two SIS actuation logic input functions; Pressurizer Low Pressure and Pressurizer Low Low Pressure. Comment: Provide discussion for deleting the low, low pressurizer pressure input to SIS.
Consumers Energy Response:
The requirement for the low, low pressurizer pressure input to SIS has not been deleted. It is required to be Operable as item #1 of Table 3.3.3:1. Plant bistables are typically labeled "low pressure" and "low-low pressure" to differentiate between two pressure settings on the same instrument loop. Plant usage, as reflected in CTS Tables 3.16 and 3.17 .1 is to refer to the actuations which occur on reducing pressure as Low Pressure SIS or Low Pressure MSIV closure, etc. The specified Allowable Values should allow the reviewers to correlate the individual bistable labeling on the plant Logic Drawings with the Function specified in the CTS and ITS. Affected Submittal Pages: No page changes .
36
CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO JANUARY 6, 1999 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.3, INSTRUMENTATION
NRC REQUEST:
3.3.3-7 NRC REQUEST: 3.3.4-1 Resolved during 10/27/98 meeting; no Consumers Energy response required. ITS 3.3.4 JFD/DOC none CTS Table 3.17 .2 requires two SIS logic (Initiation, Actuation, and low pressure block reset) two channels to be operable. ITS 3.3.4 specifies the SIS logic as "pressurizer pressure-low." Comment: Provide discussion for this change. Consumers Energy Response: CTS Table 3.17.2, Item 1, requires two channels of:
a. SIS Manual Initiation,

b. SIS Logic (Initiation, Actuation, and low pressure block auto reset),

c. CHP Signal SIS Initiation (SP Relay Output),
and four channels of:
d. Pressurizer Pressure Instruments to be operable.
ITS LCO 3.3.4 and Table 3.3.4-1, Item 1, require two ESF manual initiation channels for SIS .
That is equivalent to item "a" above. ITS LCO 3.3.4 and Table 3.3.4-1, Item 1, require two ESF actuation logic channels for SIS. That is equivalent to the Initiation and Actuation requirements of item "b" above. ITS LCO 3.3.4, requires the bypass removal channels associated with the required ESF channels to be Operable. That requirement includes the "low pressure block auto reset" requirement of item "b" above. ITS LCO 3.3.4 and Table 3.3.4-1, Item S (as renumbered in this revision), require two ESF actuation logic channels for CHP. That provides an equivalent requirement to item "c" above. The "SP" relays referred to in CTS Table 3.17.2 are the output relays for the CHP 2-out-of-4 logic channels. One function of these relays is to initiate an SIS signal. This is illustrated on Palisades P&ID E-17 Sheets 3, 4, and 6. ITS LCO 3.3.3 and Table 3.3.3, Item 1, require four ESF bistables and associated instrumentation for SIS initiation on Pressurizer Low Pressure. That is equivalent to item "d" above. _ _ _ _ _ -~~~e'!_~b!!!~f:!_a_!_!'!!_g~: _____ ________________________ ~-- ________ _ No page changes .
37
.\ CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO JANUARY 6, 1999 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.3, INSTRUMENTATION NRC REQUEST: 3.3.4-2 ITS 3.3.4 JFD/DOC none CTS Table 3.17.2 and 3.17.3 require two manual and two logic channels for SIS, RAS, AFAS, CHP, CHR, and SGLP. iTS 3.3.4 requires two ESF Manual Initiation and two ESF Actuation Logic channels and associated bypass removal channels for each ESF Function specified in the LCO table. ITS Table 3.3.4-1 lists six ESF Functions: SIS, RAS, AFAS, CHP, CHR, and SGLP. Comment: All functions, except CHP, list Manual Initiation Functions twice; once in the LCO and again as a separate function in Table 3.3.4-1. Provide additional justification to explain the translation of CTS LCO requirements into the proposed ITS. Consumers Energy Response: Consumers Energy Company agrees that the initial ITS presentation of Table 3.3.4-1 included details which were redundant to the LCO requirements. ITS Table 3.3.4-1 has been revised to deleting these details. The requirements are unchanged since both the actuation logic and manual initiation channels are required by the LCO. The SIS actuation on CHP is an integral part of the CHP actuation and is required to be operable as part of that actuation logic. SR 3.3.4.1 wording was revised to clarify that this test applied only to the SIS Actuation
Channels .
As part of the revision to Table 3.3.4-1, footnote (d) was corrected. The footnote in our January 26, 1998 submittal was inadvertently taken from Table 3.3.1-1. The correct Table 3.3.4-1 footnote is modeled after from STS Table 3.3.4-1. The CTS markups and DOC L.1 addressed the proper footnote. Affected Submittal Pages: Att 1, ITS 3.3.4, page 3.3.4-3 Att 1, ITS 3.3.4, page 3.3.4-4 Att 1, ITS 3.3.4, page 3.3.4-4 insert Att 5, STS 3.3.5, page 3.3-25 Att 5, STS 3.3.5, page 3.3-26 Att 5, STS 3.3.5, page 3.3-26 insert
38
ESF Logic and Manual In1t1at1on 3.3.4 SURVEILLANCE FREQUENCY ( E>..c.H SIS A.C.Tl)~TIOU C.11t:..N~l:-\-) ~ SR 3.3.4.1 Perfonn functional test ofJnonnal and 92 days J.J.'f-2. (~~t>BV)-1gp;nr power functions I usifi§ t§eftl l_c_t_. SR 3.3.4.2 92 days
SR 3.3.4.3 Perfonn a CHANNEL FUNCTIONAL TEST. 18 months
'2)8-a---
Amendment No. 01/20/98 Palisades Nuclear Plant 3.3.4-3
ESF Logic and Manual Intttat1on
Table 3.3.4-1 (page 1 of 1)
En ineered Safety Features Actuation Lo ic and Manual 3.3.4 FUNCTION MODES
1. Safety Injection Signal (SIS) (A.)
~ Containment High Pressure Signal p;. 0'ati;?L'og~ 1,2,3,4 Contain111ent Hfgh Radiation Sfgnal (CHR) be automatically
SIS A.C..TUA.TlOt-.1 Palisades Nuclear Plant
'B~ 38-b 3.3.4-4 Amendment No. 01/20/98
bC:C::.\\ Df'i '3 .'3
..r~'Se~ 2 l~'f:>~e'r A.S FOo\UO\E.. (d) \b i~~L.E: "S,5,y-1 * ---- --- -----~--- - - ------ - - --- --

~-

------- --- - - - - -- ----~ ---
Orit or- Mer~ F.,..._c +;O'\<.~ 1()1~ 4wo t'lo.~VJ..\ :t ... ;+:A.+;,......, or ~d..r~-t:~ Le~; l.
~ . . :.) RA'I t.kc~~el~ \\A.C~er .... ~~"2.. -hr- r11 ha.:ho; J, '$.~*I F. . . 1At.~~D'V\S 5 0'" lD ESF~ Log1c and Manual~ (~ _
I
-~
c!..TS REQUIRED ACTION COMPLETION TIME [J, 17,J, S"] ~Required Action and ~ Be in MODE 3. 6 hours associated Completion Time of Condition~ AHl2 not met~ W ~ (])-- ~ hllAc_i;o~ 5 or- filJ Be in MODE S. 36 hours J.2AI.
~.,3.!-1 I
SURVEILLANCE REQUIREMENTS ~At. 3.~.Li-2 I SURVEILLANCE FREQUENCY
~
SR f>A.I:
~~.'/*(p 2.'"', e ays assoc ate w p ant eguipment ',~hat cann~t be operit~d dur1ng)>~t o)>eration ir:~nly req'tl-lred to be "
test~d during ach MODE S*,entry
excee"Q~g 24 hour-s unless tis-~d rin the previou~
~92@ days --@
Perform a CHANNEL FUNCTIONAL TEST on each A AS logic channel. c!RAZ 'r.1.~-, w~ Perform a CHA~NEL SR 3 .3 ~- FUNCT!ONAL TEST <bn"iid}l ~lSJl months~
.£'.Sf.AS Maoilil- Ic:i.p cbann }
5R 3.3.4,l ?er{orM ~u.~ct;cntt.\ +t9s.\: tl~ E:-A.C\-\ $IS qz. d.11.vs ACiUATlot-l Cl-Vl..r-J~t:-L UOe.Mb.L
"->JD STA~l)l3Y 'POWE'f:'.: FulJCT'tONS.
CEOG STS 3.3-25 38-ciJ Rev 1. 04/07/95
\ \
ESF@ Logic and Manual
. T.tlle 3.3.5*1 Cpqe 1 of 1)
E111hwered Safety FHtur* Acn.wtion\~t* Aetuetion Logic: rd\ c:11 .... 1 Applic:abi
\
I FIMCT I Oii APPL!~£ MODES lefety Inject! Actu.tlon lltnal n.wtfon lftnal 1,2,3, t4l ontlll,_,t lsolatfon
4. llaf I t - Isolation Si tuatlon lftnal 1,2,3,4 1,2,3,4 11 I ..._Cl 5.
6. lecfrcu A.uallfary tlon Ac:tuetion SllN 1,2,3,4 1,2,3
, I l ..is~ '
CEOG STS 3.3-26 38-e_,
Rev 1. 04/07/95
SECTION 3.3 f?Ar.
INSERT 1 Table 3.3.4-1 (page 1 of 1) 3.3.'-/-Z
3. !.4- !
Engineered Safety Features Actuation Logic and Manual Initiation APPLICABLE FUNCTION MODES
1. Safety Injection Signal (SIS) 1 ~

2. Steam Generator Low Pressure Signal (SGLP) !bl !cl

3. Recirculation Actuation Signal (RAS)

4. Auxiliary Feedwater Actuation Signal (AFAS)

5. Containment High Pressure Signal (CHP) !el

6. Containment High Radiation Signal (CHR)
(a) SIS actuation by Pressurizer Low Pressure may be manually bypassed when pressurizer pressure is ~ 1700 psia. The bypass shall be automatically removed whenever pressurizer pressure is > 1700 psia.
(b) SGLP actuation may be manually bypassed when SG pressure is ~ 565 psia. The bypass shall be automatically removed whenever steam generator pressure is
> 565 psia.
(c) Manual Initiation may be achieved by individual component controls. (d) Not required to be OPERABLE when all Main Steam Isolation Valves (MSIVs) are closed and deactivated, and all Main Feedwater Regulating Valves (MFRVs) and MFRV bypass valves are either closed and deactivated, or isolated by closed manual valves. (e) Manual Initiation channels not required . 3.3-26
CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO JANUARY 6, 1999 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.3; INSTRUMENTATION
NRC REQUEST:
3.3.4-3 ITS 3.3.4 JFD/DOC none T 3.3.4-1 is titled, "Engineered Safety Features Actuation Logic and Manual Channel Appiicability." Comment: A more appropriate title is", "Engineered Safety Features Actuation Logic and Manual Initiation" Consumers Energy Response: The title of Table 3.3.4-1 has been revised as suggested. Affected Submittal Pages: Att 1, ITS 3.3.4, page 3.3.4-4 Att 5, NUREG, 3.3-26 Att 5, NUREG, 3.3-26 insert
39
\ ESF Logic and Manual Initiation
Table 3.3.4-1 (page 1 of 1)
En ineered Safety Features Actuation Lo ic and Manual APPLICABLE 3.3.4 FUNCTION HODES
1. Safety Inject 1on S1gna1 (SIS) (a.)
1,2,3
~ Contain~ent High Pressure Signal p;. 7fuat1i<og~ 1,2,3,4 (CHR)
be automatically

SG:.LP Ac.TvA.T\of...l SIS A.C..TUA.TlO~ *e;,'/
Palisades Nuclear Plant 3lt-o._ 3.3.4-4 Amendment No. 01/20/98
,1->t ~
ES~ Logic and Manual@~
3.3Srt--(D FUllCTICll APPLlcAaiE MODES
'\.
lefety JnJeetl Actu.tlon lllNL 1 ,Z,3, [4] CGnt*f,_,t Spray tu.tlon ll1Nl 1,Z,3, C4J
~
3. 1,Z,3,4

4. 1,Z,3,4

5. 1,Z,3,4
.~ ** Auaf l l*r"Y 1,Z,3 l~S~ \
CEOG STS 3.3-26 Rev 1. 04/07/95
-~ SECTION 3.3
** INSERT 1 Table 3.3.4-1 (page 1 of 1) f?AI:
3.3.'-l-Z
3. !.&.4- !
Engineered Safety Features Actuation Logic and Manual Initiation APPLICABLE FUNCiION MODES
1. Safety Injection Signal (SIS)'~ 1,2,3

2. Steam Generator Low Pressure Signal (SGLP) (bl <cl

3. Recirculation Actuation Signal (RAS) 1,2,3

4. Auxiliary Feedwater Actuation Signal 1,2,3 (AFAS)

5. Containment High Pressure Signal 1,2,3,4 (CHP) <el

6. Containment High Radiation Signal (CHR) 1,2,3,4
* (a) SIS actuation by Pressurizer Low Pressure may be manually bypassed when pressurizer pressure is ~ 1700 psia. The bypass shall be automatically removed whenever pressurizer pressure is > 1700 psia.
(b) SGLP actuation may be manually bypassed when SG pressure is ~ 565 psia. The bypass shall be automatically removed whenever steam generator pressure is
> 565 psia.
(c) Manual Initiation may be achieved by individual component controls. (d) Not required to be OPERABLE when all Main Steam Isolation Valves (MSIVs) are closed and deactivated, and all Main Feedwater Regulating Valves (MFRVs) and MFRV bypass valves are either closed and deactivated, or isolated by closed manual valves. (e) Manual Initiation channels not required . 3.3-26
CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO JANUARY 6, 1999 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.3, INSTRUMENTATION
NRC REQUEST:
3.3.4-4 ITS SR 3.3.4.1, SR 3.3.4.3 CTS Tables 3.17.2 and 4.17.2 DOC A.1 CTS Tables 3.17.2 and 4."17.2 Function ib, SIS Logic, CHANNEL FUNCTIONAL TEST references footnote (a), requiring verification of all automatic actuations and automatic resetting of the low pressure block each 18 months. This CTS SR is shown to be ITS SRs 3.3.4.1 and 3.3.4.3. However, the requirements of footnote (a) to verify automatic actuation and automatic resetting of the low pressure block is deleted. Comment: Additional discussion and justification for the change is required in the conversion documentation to verify the current licensing basis is maintained. Explain how the automatic low pressure block actuation and reset are tested by ITS SR 3.3.4.1 and ITS SR 3.3.4.3 as required for ITS Table 3.3.4-1 Function 1b. Consumers Energy Response: The intent of our January 26, 1998 submittal was that the requirement to "Verify all automatic actuations and automatic resetting of the low pressure block" would be accomplished as part of the Channel Functional Testing required by ITS SR 3.3.4.3. Since the bistables which provide the bypass removal signals are physically part of the instrument loops required by LCO 3.3.3, their calibration would be performed by SR 3.3.3.3 . LCO 3.3.3 has been revised to explicitly address the bypass removal bistables and associated instrument channels in response to RAI 3.3.3-1. The bypass removal bistables and associated instrument channels will be calibrated under SR 3.3.3.3. ITS LCO 3.3.4 requires the bypass removal channels to be Operable, -SR 3.3.4.3 requires a Channel Functional Test of all channels specified in the LCO. Affected Submittal Pages: No page changes .
40
CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO JANUARY 6, 1999 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.3, INSTRUMENTATION NRC REQUEST: 3.3.4-5 ITS 3.3.4 DOCA.6 CTS Tables 3.17.2 and 3.17.3 CTS Tables 3.17.2 and 3.17.3 aie ievised to incorporate ITS Tabie 3.3.4-1 Footnote (c), allowing manual initiation achieved by any individual component controls for RAS and AFAS (DOC A.6 states SGLP also, although neither the CTS mark-up nor the ITS apply Footnote (c) to the SGLP Function). Comment: It is not clear from the discussion of change what "any individual controls" includes. It is not clear that these ESF Functions include more than 2 Manual Initiation trains initiated by one control switch each. Provide additional clarification of the hardware configuration including any additional actuation devices not included in LCO 3.3.4. This is a SCOPE issue because it is a change to the STS and a change to the ITS. Provide justification for not adopting current TS requirements for these functions. Consumers Energy Response: There is no intent that ITS requirement~ for manual initiation capabilities differ from CTS requirements. The Palisades design does not include a single manual push-button to initiate each train of each ESF function. SIS, AFAS, and CHR functions are provided with such a single pushbutton manual initiation capability for each train. Initiation of SGLP and CHP may be accomplished manually, but only through use of the normal controls for individual components. SGLP, for instance, is manually initiated by manually closing the MSIVs (either of two switches closes both MSIVs}, MFRVs and bypass valves. Both the CTS markups and ITS Table 3.3.4-1 have been corrected to apply footnote (c) to only SGLP. DOC A.6 has been revised. A new footnote, (e}, has been applied to the CHP signal stating that manual initiation channels are not required (they are not installed in the plant, nor required by CTS). Conforming changes have been made to the Bases. These changes more precisely represent the CTS requirements in the ITS format. Affected Submittal Pages: Att 1, ITS 3.3.4, page 3.3.4-4 Att 1, ITS 3.3.4, page 3.3.4-4 insert (marked up) Att 1, ITS 3.3.4, page 3.3.4-4 (As revised) Att 3, CTS 3.3.4, page 3-67 Att 3, CTS 3.3.4, page 3-69 Att 3, DOC 3.3.4, page 2 of 5 (former submittal) Att 3, DOC 3.3.4, page 2 of 8
41
ESF Logic and Manual Initiation
Table 3.3.4-1 (page 1 of 1)
Engineered Safety Features Actuation Logic and Manual Channel Applicability APPLICABLE 3.3.4 FUNCTION MODES
1. Safety Injection Signal (SIS) (~)
1,2,3
~ Containment High Pressure Signal p;. )?U'atii<og~ 1,2,3,4 (CHR) be automatically
Palisades Nuclear Plant Amendment No. 01/20/98

6e:c.11ot-.t 3.3
.I~ S-e:g.\ 2
{d) Mot ~!~cl' to be OPERABLE when all Main Stearn I so 1at ion I/ al ves (MS I Vs} a re c:lQffcf:::iid\ditcthated. and 11.ll Main feedwater Regulating Valves (HFRVs} and NFflV'IUifva1ves are either closed and deacthilted. or isolated by closed JMftUil?:'**l~~~: .
11-b

ESF Logic and Manual Initiation 3.3.4 Table 3.3.4-1 (page 1 of 1}
Engineered Safety Features Actuation Logic and Manual Initiation APPLICABLE FUNCTION MODES
1. Safety Injection Signal (SIS)Cal 1,2,3

2. Steam Generator Low Pressure Signal (SGLP) CblCcl

3. Recirculation Actuation Signal (RAS) 1,2,3

4. Auxiliary Feedwater Actuation Signal 1,2,3 (AFAS)

5. Containment High Pressure Signal 1,2,3,4 (CHP) Cel
6. Containment High Radiation Signal (CHR)
(a) 1,2,3,4 SIS actuation by Pressurizer Low Pressure may be manually bypassed when pressurizer pressure is ~ 1700 psia. The bypass shall be automatically removed whenever pressurizer pressure is > 1700 psia. (b) SGLP actuation may be manually bypassed when SG pressure is ~ 565 psia. The bypass shall be automatically removed whenever steam generator pressure is
> 565 psia.
(c) Manual Initiation may be achieved by individual component controls. (d) Not required to be OPERABLE when all Main Steam Isolation Valves (MSIVs) are closed and deactivated, and all Main Feedwater Regulating Valves (MFRVs) and MFRV bypass valves are either closed and deactivated, or isolated by closed manual valves. (e) Manual Initiation channels not required .
Palisades Nuclear Plant 3.3.4-4 zf 1-c_
Amendment No. 03/31/99
lil- 3.17 I! 5'.3. 4- -.D INSTRUMENTATION SYSTEMS
~~9.4.-.D Table : ;
lnstrymentation Ooerating Regyirements for Engineered Safety Featyres Perm1ss1ble Operational functional Un1t Byoasses [1] 1. Safety lniection Signal {SIS) a .. Manual Initiation 2 frp,e,rmJ*r f'iururc*L;c..J])
b. SIS Lo ict ""== 2 (IM.ti at i'e{I, Act\l_at 1on, '-.nd low 0-,.essu~bloclt--.auto r~et

c. CHP Signal SIS Initiation ~
@"Rel i)---.Du~f})'-----
r{j-tf.,3.J.'f-11
d. Pres rizer Pressure 4 2 s 1700 psi a ~o+e (~
Instr ent Channels .. PCS pressure. [3] 2. Recirculation Actuation Signal {RAS)
a.
b. Manual Initiation RAS Logic
c. @RWT '§vel Switches 2
2 4\ 3 None.
~ \. NonY-@
[YJ 3. Auxiliary Feedwater Actuation Signal {AFAS}
a. Manual Initiation

b. AFAS Logic 2
2 h}@ 2
d. enerator Level 2
Amendment No. ~ *
~* 3,4--J) 3 .17 INSTRUMENTATION SYSTEMS Tab 1e<:J)'lON *Permissible Operat 1ofla i Fynct1onal Un1t Byoasses
[!?~ 1. Containment H1gh Pressyre (CHP)
a. CHP 1og1c Trains 2

b. Con a ment Pressure None.
Swttche - Left Train
c. 4 None.
[~] 2. Containment High Radiation (CHR) ~~i: 1.'5.'4-~ I
.~
a. Manual Initiation

b. CHR Logic Trains 2
2 (}@ ~
[2]
c.
3. Stem ~enerator L2! eressuce 4
.<SGLP)~
2
~A't 3.,,"1*?
(!O""set/~in
~
a. Manual Actuat;onCc) to@

b. SGLP Log;c Tra;ns 2

c. "A" team Generator 2 < 550 ps;g Press re Steam Pressure.

d. Generator 2 < 550 ps;g Steam Pressure.
"-S~ p.r ; I)_
1 0 West R om Hon;tor 1 0 Amendment No. 162 October 26, 1994
3-69 9-/- e_
A-dd l~bl.e 33A-I
,L1J6 To..lol e.
tJ 6+e C.c) 3,3.'l -I ,\Jote {d) Pll.Jf e 4- of. 7
ATTAC1'ENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.3.4, ESF LOGIC AND MANUAL INITIATION A.4 A Note was added to the Actions of CTS 3.17.2 and 3.17.3 which allows separate Condition entry for each Function. The Note in ITS 3.3.4 provides explicit instructions for proper application of the Actions for Technical Specification compliance. In conjunction with the proposed Specification 1.3 - "Completion Times, " this Note provides direction consistent with the intent of the existing Actions for the ESF instrumentation. This change is consistent with NUREG-1432. A.5 CTS 3.17.3 requires the instrumentation channels to be OPERABLE except as allowed by the permissible operations bypasses. Reference to operational bypasses is revised such that they are specifically addressed in the ITS. In addition, the overall Applicability of this Specification is when the PCS is above COLD SHUTDOWN. The Applicability of ITS 3.3.4 is "According to Table 3.3.4-1." ITS Table 3.3.4-1 includes a*MODES column where the Applicable conditions are included for each Function. The ITS Applicability associated with CTS Table 3.17.3 is MODES 1, 2, 3, and 4. The differences in the applicability between the C_TS and ITS are negligible. For the CTS' COLD SHUTDOWN versus the ITS MODE 5, the temperature requirement is being less than 210°F versus being less than 200°F in the ITS . This difference which is common between the CTS terms HOT SHUTDOWN and COLD SHUTDOWN and the corresponding ITS MODES 3 and 5 is the reactivity condition. The ITS MODE 3 and 5 are defined, as a reference point, by a reactivity condition of Keff < .99. However, in ITS Section 3 .1, the equivalent amount of SHUTDOWN MARGIN is required as that specified in the CTS definitions of HOT SHUTDOWN and COLD SHUTDOWN. Therefore, the amount of SHUTDOWN MARGIN is considered to be same when the requirements of proposed ITS 3 .1 are considered. These changes with respect to the Applicability are all considered to be administrative in that there is no significant impact to operation of the plant and they reflect the terminology and usage rules of NUREG-1432. A.6 ! CTS Table 3.17.2 and 3. .3 are revise to incorporat Footnote (c) which states manual
tiation may b . achieved by i ividual comg nent controls and is applic le to Ma al Initiation fi Functions 4. (SGLP), 5.a S), and 6. (AFAS). T *

foot ote is added to larify that use f any indivi al componen ontrols to a ate a F ction is adequ e to demonstr e that the Fu tion can be anually initi d. This ange is consi red to be a * *strative, in at it only pro ides clarifica ion and does not alter any chnical requir RAI 3.3.4-5

Palisades Nuclear Plant Page 2 of 5 111 -f 01/20/98
I) ATTACHMENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.3.4, ESF LOGIC AND MANUAL INITIATION A.5 CTS 3 .17. 3 requires the instrumentation channels to be OPERABLE except as allowed by the permissible operations bypasses. Reference to operational bypasses is revised such that they are specifically addressed in the ITS. In addition, the overall Applicability of this Specification is when the PCS is above COLD SHUTDOWN. The Applicability of ITS 3.3.4 is "According to Table 3.3.4-1." ITS Table 3.3.4-1 includes a MODES column where the Applicable conditions are included for each Function. The ITS Applicability associated with CTS Table 3.17.3 is MODES 1, 2, 3, and 4. The differences in the applicability between the CTS and ITS are negligible. For the CTS COLD SHUTDOWN versus the ITS MODE 5, the temperature requirement is being less than 210°F versus being less than 200°F in the ITS. This difference which is common between the CTS terms HOT SHUTDOWN and COLD SHUTDOWN and the corresponding ITS MODES 3 and 5 is the reactivity condition. The ITS MODE 3 and 5 are defined, as a reference point, by a reactivity condition of Keff < .99. However, in ITS Section 3 .1, the equivalent amount of SHUTDOWN MARGIN is required as that specified in the CTS definitions of HOT SHUTDOWN and COLD SHUTDOWN. Therefore, the amount of SHUTDOWN MARGIN is considered to be same
when the requirements of proposed ITS 3.1 are considered. These changes with respect to the Applicability are all considered to be administrative in that there is no significant impact to operation of the plant and they reflect the terminology and usage rules of NUREG-1432.
A.6 The requirements of CTS Tables 3 .17. 2 and 3 .17. 3 are revised to incorporate Footnote l\n (c), which states that manual initiation may be achieved by individual component I;_ controls, and is applicable to Manual Initiation for Function 4.a (SGLP). This footnote I \V' is equivalent to the CTS entry requiring "1 set [of controls]/train, and is added to 1~ clarify that use of any individual component controls to actuate the SGLP Function is I adequate. This change is considered to be administrative, in that it only provides I* clarification and does not alter any technical requirements. I~ A.7 CTS Table 4.17.2 has been revised by adding a Note to the 92 day CHANNEL FUNCTIONAL TEST which states that testing of the Actuation Logic shall include verification of the proper operation of each initiation relay. This change is considered to be administrative, in that it only provides clarification and does not alter any technical requirements .
Palisades Nuclear Plant 05/30/99
CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO JANUARY 6, 1999 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.3, INSTRUMENTATION
NRC REQUEST:
3.3.4-6 ITS 3.3.4 J FD/DOC none CTS Table 4.17.2 CTS includes requirements to perform an AFAS Logic CHANNEL FUNCTIONAL TEST at a 92 day interval. The ITS translates the CTS requirement as SR 3.3.4.2 which specifies testing is to be performed using test circuits. Comment: This is a SCOPE issue because it is a change to the STS and a change to the ITS. Provide justification for not adopting current TS requirements for these functions. Consumers Energy Response: The wording of ITS SR 3.3.4.2 has been revised to more clearly adopt CTS requirements in the ITS. The revised wording omits the reference to use of test circuits and the note requiring verification of each initiation relay. The note (STS Note 1) requiring verification of each initiation relay is unnecessary because the allowance provided by STS Note 2 was not included. Neither note is necessary since these circuits are testable during power operation. JFDs 3.3.5-8 and 3.3.5-10 have been revised. Affected Submittal Pages: Att 1, ITS 3.3.4, page 3.3.4-3 Att 5, NUREG, page 3.3-25 Att 6, JFD 3.3.5, page 1 of 2 (former submittal) Att 6, JFD 3.3.5, page 1 of 2
42
ESF Logic and Manual In1t1at1on 3.3.4 SURVEILLANCE FREQUENCY ( EM..H SIS A.C.Tl.lA.TIOU C.11ti..1'H.\C:-L.)
~
SR 3.3.4.1 Perfonn functional test ofJnonnal and 92 days J.~.'i-7. (STANt>BV)-;gf1{nf,1 power functionslusifi§ te§'tl l_Lt_.
SR 3.3.4.2 92 days

SR 3.3.4.3 Perfonn a CHANNEL FUNCTIONAL TEST. 18 months Palisades Nuclear Plant
'I~ -
3.3.4-3 Amendment No. 01/20/98
i i c:'.Jl\t er t-\c<"c! F.._ ... d,O"Y..~ lc.)1~ 4wo r'to."v.L\ 1 ...;t:d;"'-", or AL+..,.~1:~ lc~~' t.l.o.~i<els '1v.o~~rjl.l..~.Q.. -hr-Fii.~'--tl ~5. 5 o~ lD
continued CONDITION REQUIRED ACTION COMPLETION TIME
[J, 17,J, s-J ~Required Action and ~ Be in HOOE 3. 6 hours associated Completion Time of Condition~ Alil2 not met~ @ ~ Be in MODE 5. 36 hours (])-+ ~ hll.\C..{10~ 5 Or' 'iD SURVEILLANCE REQUIREMENTS ~At. 3,'5.Y-2 I SURVEILLANCE FREQUENCY
--+
SR J:>AJ:
~~.'/-(,,
2.'-.. e ays assoc ate w p ant e~ipment
*,~hat cannot be operitt!d during ~
operation ar-~nly reqU-tred to be tesled during ach HOOE S**,entry exceeoi.!19 24 hour unless tes*~d rin _the previo~
Perform a CHANNEL FUNCTIONAL TEST on each A
AS logic channel.
~gz@ days --{!) ~A~ r:S,"f -~
SR 3.3~~ (SFAS Perform a CH~~NEL FUNCTIONAL TEST<ti~ Hiou.il- Icip cbaone1.
~la} months~ © Per.{on"'- .Q.u..~~-t;cv-.~\ +'1s-t: "~ 'E:-~""" ~1S '11.. dei.vs AeiUATlot-l. C'-'Ar-!~t:-L tJo'f:.MA.L ~t.J'D t;TA>-ll)8Y 'Powe-~ Fu1JCllONS.
CEOG STS 3.3-25 Rev 1. 04/07 /95

ATTACHMENT 6 JUSTIFICATION FOR DEVIATIONS SPECIFICATION 3.3.5, ESFAS LOGIC AND MANUAL TRIP 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 word Analog was removed from the title of each Specification in Section 3.3.
Palisades is an Analog plant and specifically listing this in the title is unnecessary. Analog was placed in NUREG-1432 to distinguish between Analog and Digital Specifications.
7. ISTS 3.3.2, is not included in the ITS, therefore the following specificatipns have been renumbered, where applicable to reflect this deletion. ...,) FI 8 8.
Palisades Nuclear Plant Page 1of2 01/20/98
ATTACHMENT 6 JUSTIFICATION FOR DEVIATIONS SPECIFICATION 3.3.5, ESFAS LOGIC AND MANUAL TRIP 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 word Analog was removed from the title of each Specification in Section 3.3.
Palisades is an Analog plant and specifically listing this in the title is unnecessary. Analog was placed in NUREG-1432 to distinguish between Analog and Digital Specifications.
7. ISTS 3.3.2, is not included in the ITS, therefore the following specifications have been renumbered, where applicable to reflect this deletion.

8. The SRs have been modified to reflect the current licensing basis. Proposed SRs 3.3.4.1 and 3.3.4.2 represent explicit CTS functional test requirements.
SR 3.3.4.3, the Channel Functional Test requirement, applies to all channels
'°I specified in the LCO. ..
9. Bypass removal logic channels are addre~sed in LCO 3.3.4, rather than LCO 3.3.3,
because they bypass the logic channels rather than individual instrument channels, and are tested along with the ESF logic, consistent with current design and licensing basis.
Palisades Nuclear Plant Page 1of2 05/30/99
CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO JANUARY 6, 1999 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.3, INSTRUMENTATION
NRC REQUEST:
3.3.4-7 ITS 3.3.4 JFD 11 CTS Tables 3.17.2 and 4.17.2, Functional Units 1.b and 1.c JFD 11 is not used in ITS 3.3.4. Comment: Provide a revised submittal markup. Consumers Energy Response: Reference to JFD 11 for ITS 3.3.4 (STS 3.3.5) was inadvertently omitted from the markup of STS page 3.3-26. The changes made to ITS Table 3.3.4-1 in response to RAls 3.3.4-2 and 3.3.4-5 eliminate the need for JFD 11. Affected Submittal Pages: Att 6, JFD 3.3.5, page 2 of 2 (former submittal) Att 6, JFD 3.3.5, page 2 of 2
43

ATTACHMENT 6 JUSTIFICATION FOR DEVIATIONS SPECIFICATION 3.3.5, ESFAS LOGIC AND MANUAL TRIP Discussion
9. Bypass removal channels are addressed in LCO 3.3.4, rather than LCO 3.3.3.
because they are tested along with the ESF logic, consistent with current design and licensing basis.
10. ISTS SR 3.3.5. l Note 2 has been deleted. Circumstances described by the Note are adequately addressed in Table 3. 3 .4-1.
NO"t" t)~~~. f?Pt \ ).1. &.of , lL Details related to des~riptiQR aRd s~gpe gf lggi~ ~ir~Yits assg~iated \Jlitl:l £rn iRitiatieA f'Hnetions added, eonsistent with details f'eloeated ff'om CTS Table3 3.17.2 and 4 .17.2, Ftinetional Units 1.8 aAa l.e (see DOC LA. l).
12. TSTF-187 is incoq>orated to add a Condition for two inoperable Actuation Logic Channels to ESF Logic and Manual Trip. There is currently no Condition for cwo inoperable Actuation Logic channels. This change adds the Condition of two Actuation Logic channels inoperable with the Required Action to shutdown the plant.
This change puts the appropriate Actions in the Specification and eliminates any confusion that may arise from not addressing the AFAS actuation logic and manual trip functions in the Actions. Paiisades Nuclear Plant Page 2 of 2 01120798 13-()-/
ATTACHMENT 6 JUSTIFICATION FOR DEVIATIONS SPECIFICATION 3.3.5, ESFAS LOGIC AND MANUAL TRIP Change Discussion
10. ISTS SR 3. 3 .5 .1 Notes have has been deleted. Circumstances described by the Notes I £!-
are not applicable to Palisades. I~
11. Not used. II .""

12. TSTF-187 is incorporated to add a Condition for two inoperable Actuation Logic
~
Channels to ESP Logic and Manual Trip. There is currently no Condition for two inoperable Actuation Logic channels. This change adds the Condition of two Actuation Logic channels inoperable with the Required Action to shutdown the plant. This change puts the appropriate Actions in the Specification and eliminates any confusion that may arise from not addressing the AFAS actuation logic and manual trip functions in the Actions .
Palisades Nuclear Plant Page 2 of 2 05/30/99
CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO JANUARY 6, 1999 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.3; INSTRUMENTATION NRC REQUEST: 3.3.4-8 ITS 3.3.4 JFD 9 CTS Tables 3.17.2 and 4.17.2 Bypass channels are included in the ITS LCO as "associated bypass removal channels." Comment: Revise the ITS SRs to include testing requirements for bypass removal channels and modify Table 3.3.4-1 to specify those Functions that have bypass capability. Consumers Energy Response: ITS LCO 3.3.4 requires the bypass removal channels to be Operable; Footnotes (a) and (b) to Table 3.4.4-1 denote which channels have bypass removal requirements; ITS SR 3.3.4.3 requires performing a Channel Functional Test on each channel required by the LCO. The Channel Functional Test SRs in STS 3.4.5 include enhanced wording only because neither of those SRs apply to all required channels. Affected Submittal Pages: No page changes .
44
CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO JANUARY 6, 1999 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.3, INSTRUMENTATION NRC REQUEST: 3.3.4-9 ITS 3.3.4, DOC A.3 ITS 3.3.3, DOCs M.1 & M.3 CTS 3.17.2.5 and 3.17.3.5 The "Minimum OPERABLE Channels" column and TS actions based on this requirement are deleted from CTS 3.17.2.5 and 3.17.3.5. These CTS specifications are included in ITS LCOs 3.3.3 and 3.3.4. In the CTS mark up for ITS 3.3.3 the changes are discussed in DOC M.1 and in CTS 3.17.3.5 the changes are discussed in DOC M.3. The DOCs contain identical justifications. In the CTS mark up for ITS 3.3.4 these same changes are discussed in DOC A.3. The staff notes that Mode changes that follow in subsections a) and*b) of CTS 3.17.2.5 and 3.17.3.5 are also identical and that the DOC justifications are the same for the CTS mark up for ITS 3.3.3 and 3.3.4. Comment: The staff questions the use of both a more restrictive and administrative change categorization for the same CTS change. Evaluate the differences between the CTS markup M-DOCs and A-DOCs and provide a consistent category justification. Consumers Energy Response: The markups of CTS 3.17 .2 and 3.17. 3 associated with ITS LCOs 3. 3. 3 and 3. 3.4 and associated DOCs have been rewritten to be more consistent. Affected Submittal Pages: Att 3, CTS 3.3.3, page 3-66 Att 3, CTS 3.3.3, page 3-68 Att 3, CTS 3.3.4, page 3-66 Att 3, CTS 3.3.4, page 3-68 Att 3, DOC 3.3.3, page 3 of 9 (former submittal) Att 3, DOC 3.3.3, page 4 of 9 (former submittal) Att 3, DOC 3.3.3, page 5 of 9 (former submittal) Att 3, DOC 3.3.3, page 6 of 9 (former submittal) Att 3, DOC 3.3.3, page 7 of 9 (former submittal) Att 3, DOC 3.3.3, page 3 of 8 Att 3, DOC 3.3.3, page 4 of 8 Att 3, DOC 3.3.3, page 5 of 8 Att 3, DOC 3.3.3, page 6 of 8 Att 3, DOC 3.3.4, page 3 of 5 (former submittal) Att 3, DOC 3.3.4, page 4 of 5 (former submittal) Att 3, DOC 3.3.4, page 3 of 8 Att 3, DOC 3.3.4, page 4 of 8 Att 3, DOC 3.3.4, page 5 of 8 Att 3, DOC 3.3.4, page 6 of 8
45
3.3.3
~------- CA.D-if
ITS 3.17 INSTRUP£NTATION SYSTEMS Engineered Sof ety Feotures CESF> Instrumentotion SEE 3.3.~
Applicability Action ...,._ ___,( Add ACTIONS Note )--@ 3.17.2.1 [Cond Al 3.17.2.2 With one ESF instrU11ent channel inoperable for one or more functions, [NOTD-~xcept SIRWT Leve 11: [RA A.ll a) Place the trip unit for each affected ESF function in the tripped condition within 7 days. [Cond Bl 3.11:2.3 With two ESF instrument channels inoperable for one or more functions, [NOTEJ--lexcept SIRWT Levell: (Add Cond. B Required Actions Note)-@ [RA B.ll a) Place one channel trip unit for each affected ESF function in the tripped condition within 8 hours, and [RA B.2l b) Restore one channel to OPERABLE status within 7 days. [Cond Cl3.17.2.4 With one SIRWT Level channel inoperable: [RA C.ll a) Bypass the 1evel switch within 8 hours, and [RA C.2l b) Restore the channel to OPERABLE status within 7 days. [Cond Dl 3 .17
2

5 rnA D.ll a) The reactor shall be placed in ic..::..~=..;..::.z.:..:=.i and
[RA 0.2] b) M.2
3-66
~5 -Q._,
Amendment No. ~. 180 Page 1 of 8
3.3.3
ITS 3.17 I NSTRUHENTATION SYSTEMS Engineered Sof ety Feotures CESFl Instrumentot1on Applicability Action
......--.....(<Add ACTIONS Note >-8 3.17.3.l
[Cond AJ 3.17.3.2 . With one Isolation Function instrument channel inoperable for one or 1DOre functions: rnA.A.lJ a). Place the trip unit for each affected Isolation Function in the tripped condition within 7 days. ~ond BJ 3.17.3.3
With two Isolation Function instrument channels inoperable for one or
*ore functions: ( Add Cond. B Required Actions \ ~
Note~ [RA B.lJ a) Place one channel trip unit for each affected Isolation Function in the tripped condition within 8 hours, and [RA B.2J b) Restore one channel to OPERABLE status within 7 days. 3.17.3.4 [Cond EJ 3.17.3.5 [RA D.1 & E.lJ a) RAI [RA D.2 & E.2J b) 3.3.4-9 M.4
9-5-b 3-68 Amendment No. 162 Page 3 of 8
3.3.4
ITS 3.17 INSTRU!1£NTATION SYSTEMS 3.3.3)
Action . ._,.___-1(
. Add ACTIONS Note >-8
[Cond AJ 3.17.2.1 With one ESF .anual control channel or ESF lo 1c channel inoperable for one or 11are functions: ERA A.lJ a) Restore the channel to OPERABLE status within 48 hours *
(sEE 3.3.3
[Cond Bl 3 .17
2

5 ERA B.ll a) The reactor shall be placed in =..:.......:~..:....::..i= and RAJ 3.3.4-9
[RA 8.2] b) M.2
3-66 Y5 -<:_
Amendment No. -162. 180 Page 1 of 7
3.3.4
ITS 3.17 Two Manual Ini t1at1on, bypass l'"emoval, and two actuation (sEE 3.3.3 1
(3.3.41 3.17.3 e nc on logic chann.els -~d;:.-::~~*==*~~~~:::.:..:wL:...::;.!,., the functions listed in Table 3.17.3 shall be OPERABLE
~ th/epe2'i ss j¥1 e pper¢1 one' byi)assey co l)ifiln .1 "----'-"-__.___ _,_____.
Applicabilitv Action Add ACTIONS Note)--@ ( ESF ~ A.5 oss remova r{~-ES_F_) [Cond Al 3 .17. 3 .1 With one i.l'SorU"tonl Function manual control or !Bo1a;t1on] Function logic channel inoperable for one or more functions: [RA A.ll a) Restore the channel to OPERABLE status within 48 hours. 3.17.3.2
3.17.3.3 (sEE 3.3.3 3.17.3.4 3.17.3.5
[Cond Bl CSGLPI [Cond Cl [RA 8.1 & C.lJ a) The reactor sha 11 be p1aced 1n i<..:..::...::......=:=...:..== and RAI * [RA 8.2 & C.2J b) 3.3.4-9 M.4
~5~J 3-68 Amendment No. 162 Page 3 of 7

A.6 ATTACHMENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.3.3, ESF INSTRUMENTATION CTS 3 .16 includes the requirement that the Engineered Safety Features (ESP) System instrumentation setting limits shaii be as stated in Table 3 .16. The CTS 3 .16 Applicability is when the associated ESP or Isolation Function instrumentation is required to be OPERABLE by Specification 3.17.2 or 3.17.3. CTS 3.16.1 requires that the instrumentation be declared inoperable when the settings are not within the allowable values of Table 3 .16 and complete the corrective action as directed by Specification 3 .17. ITS 3 .3. 3 includes all of these requirements in the associated LCO
*and ACTIONS. Therefore, this cross reference to CTS 3.17 is not necessary and is deleted. Since this change is simply a change in format it is considered administrative.
A.7 CTS Table 4.17 .2, Functional Unit 2.c, requires both a CHANNEL FUNCTIONAL TEST and a CHANNEL CALIBRATION to be performed on the SIRWT Level Switches every 18 months. In the ITS, a CHANNEL CALIBRATION (SR 3.3.3.3) is only required because the definition of CHANNEL CALIBRATION explicitly encompasses the CHANNEL FUNCTIONAL TEST. Since the requirements for CHANNEL CALIBRATION include the CHANNEL FUNCTIONAL TEST, this change is considered administrative .
MORE RESTRICTIVE CHANGES (M)
M.l cS ee ~ev I ~ e cl lXx::., /VI, I
Palisades Nuclear Plant Page 3of9 01/20/98

M.1 (continued)
ATTACHMENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.3.3, ESF INSTRUMENTATION ith respect to PCS temperature requirement differences between the CTS and ITS, the pro ed ITS MODE 3 is specified as being greater than 300°F, while the CTS HOT SHUT WN is specified as being greater than 525°F. While the ITS covers ab ader range, for hutdown the ITS Completion Time to achieve this condition is s rter, and therefore, change is more restrictive. In addition, the current re
ement to place the reactor* condition where the affected equipment is not re Ired (the Applicability is when S temperature is greater than or equal to °F) within 48 hours has been changed to in MODE 4 in 30 hours. This c ge is also more restrictive since the MODE 4 S PCS temperature cutoff of 300°F.
The other parameter which is common be COLD SHUTDOWN and the correspondin S MODES 3, 4, and 5 is the reactivity condition. ITS MODES 3, 4, and 5 are efined, a reference point, by a reactivity condition of ~ff < 0. 99. However,
ITS Section 3. the equivalent amount of SHUTDOWN MARGIN is requ* d as that specified in CTS definitions of HOT SHUTDOWN and COLD S TDOWN. Therefore, the nt of SHUTDOWN MARGIN is considered e same when the requirements of pr osed ITS 3 .1 are considered.
e will require the plant to be in a lower MODE in a shorter
e frame, 1s considered more restrictive. This change is appropriate because cont s to provide adequate time for an orderly plant shutdown without challeng1 pl t systems. This change imposes additional restrictions on plant operations and is consistent with NUREG-1432.
M.2
Palisades Nuclear Plant Page 4of9 01/20/98

M.2 (continued)
ATTACHMENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.3.3, ESF INSTRUMENTATION ith respect to PCS temperature requirement differences between the CTS and ITS, the pro sed ITS MODE 3 is specified as being greater than 300°F, while the CTS HOT SHU OWN is specified as being greater than 525 °F. While the ITS covers a b ader range, for shutdown the ITS Completion Time to achieve this condition is s tter, and therefore, e change is more restrictive. In addition, the current req
ement to place the reactor a condition where the affected equipment is not re ired (the Applicability is when CS temperature is greater than or equal to 3 °F) within 48 hours has been chang to be in MODE 4 in 30 hours. Thi hange is also more restrictive since the MODE S PCS temperature cutoff of 300°F.
The other parameter which is common the corresponding ITS MODE 3 is the reac
ty condition. ITS MODES 3 is defined, as a reference point, by a reactivity con

10n K:ff <0.99. However, in ITS Section 3 .1, the equivalent amount o HUTDO MARGIN is required as that specified in the CTS definition o OT SHUTDOW . Therefore, the amount of SHUTDOWN MARGIN is c idered to be same when t requirements of proposed ITS 3 .1 are considered.
Since this change ill require the plant to be in a lower MODE in a orter time frame, this change is nsidered more restrictive. This change is appropriate ti ause it continues provide adequate time for an orderly plant shutdown without c lenging plant s terns. This change imposes additional restrictions on plant operations d is co 1stent with NUREG-1432 .
Palisades Nuclear Plant Page 5of9 01/20/98

M.3 ATTACHMENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.3.3, ESF INSTRUMENTATION CTS 3.17.3.5 requires specific Actions be taken when the "number of OPERABLE c_ 11..nels is less than specified in the "~.1inimum OPERABLE Chaw1els" column of Tafi 3.17.3. These Actions are to place the reactor in HOT SHUTDOWN wit

12 ho sand place the reactor in a condition where the affected equipment is t required, :within 48 hours. Based on the current Applicability in CTS 3 .17. the condition ere the affected equipment is not required is when the PCS i at or below COLD SHU OWN. In the ITS, the Required Action when the "nu er of OPERABLE Ch els" is less than the CTS minimum requirements *s to enter LCO 3.0.3. The uired Action of ITS LCO 3.0.3 is to initiate ction within 1 hour to place the plant, as licable, in MODE 3 in 7 hours, in M E 4 in 31 hours, and in MODE 5 in 37 hours. The differences in the applicabili~ etween the CTS and ITS are negligible. With respe to the differences in tempera re requirements between the CTS and ITS, the proposed I MODE 3 is specified a eing greater than 300°F, while the CTS HOT SHUTDO . is specified as be* g greater than 525°F. While the ITS covers a broader range, for a s utdown the IT Completion Time to achieve this condition is shorter, and therefore the hange is ore restrictive.

For the CTS COLD SHUTDOWN, the te erature requirement is less than 210°F, versus the ITS MODE 5 which has ate era re requirement of less than 200°F. The difference of 10 degrees is negligible d has no ignificant impact on operations. The other parameter which is common tween the CT HOT SHUTDOWN and COLD SHUTDOWN modes and the co esponding ITS MO ES 3, 4 and 5 is the reactivity condition. The ITS MODE 3 ~. and 5 are defined, as eference point, by a reactivity condition of ~ff < 0. 99. wever, in ITS Section 3 .1, th equivalent amount of SHUTDOWN MARGIN *s required as that specified in the S definitions of HOT SHUTDOWN and CO D SHUTDOWN. Therefore, the amou of SHUTDOWN MARGIN is consid ed to be same when the requirements ?f prop ed ITS 3 .1 are considered.
Since this c nge will require the plant to be in a lower MODE in a shorte time frame, this chan is considered more restrictive. This change is appropriate becau it contin s to provide adequate time for an orderly plant shutdown without chall ing plan systems. This change imposes additional restrictions on plant operations an s c istent with NUREG-1432 .
Palisades Nuclear Plant Page 6of9 01/20/98

M.4 ATTACHMENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.3.3, ESF INSTRUMENTATION TS 3.17.3.5 requires specific Actions when "any action required by CTS 3.17.3 is no et AND the associated completion time has expired." These A.ctions are to pl e the r ctor in HOT SHUTDOWN within 12 hours, and in a condition where the affected guipment is not required within 48 hours. Based on the current App

ability of CTS 3. .3, the condition where the affected equipment is not required i when the PCS is at or B ow COLD SHUTDOWN. In the ITS, the Required Acti , for Containment Hi Pressure and Containment High Radiation, when "Required Action and associaf Completion Times are not met" is to enter I 3.3.3 Condition E, which re uires that the plant be in MODE 3 in 6 h rs and in MODE 5 in 36 hours, and the Requir Action, for Steam Generator Low ressure, when the "Required Action and assoc* ted Completion Times are no et" is to enter ITS 3.3.3 Condition D, which requires t the plant be in MODE in 6 hours and in MODE 5 in 30 hours (see DOC L.3 below).
For the CTS COLD SHUTDOWN, th emper re requirement is less than 210°F, versus the ITS MODE 5 which has a tern rature requirement of less than 200°F. The difference of 10 degrees is negligible andJia o significant impact on operations. The other parameter which is common betW;een the TS HOT SHUTDOWN and COLD SHUTDOWN modes and the corre~onding ITS ODES 3, 4 and 5 is the reactivity
condition. The ITS MODE 3, 4,)ind 5 are defined, a reference point, by a reactivity condition of ~ff < 0. 99. How,ev'er, in ITS Section 3 .1, e equivalent amount of SHUTDOWN MARGIN is quired as that specified in th CTS definitions of HOT SHUTDOWN and COL HUTDOWN. Therefore, the am nt of SHUTDOWN MARGIN is considere to be same when the requirements of pr osed ITS 3 .1 are considered.
Since this cha ge will require the plant to be in a lower MODE in a sho r time frame, this change *s considered more restrictive. This change is appropriate beca e it continue to provide adequate time for an orderly plant shutdown without cha nging plant stems. This change imposes additional restrictions on plant operations a istent with NUREG-1432.
~ee.. rev16e LESS RESTRICTIVE CHANGES - REMOVAL OF DETAILS TO LICENSEE CONTROLLED DOCUMENTS (LA)
There were no "Less Restrictive" changes associated with this specification .
Palisades Nuclear Plant Page 7of9
.y5;_
01/20/98
ATTACHMENT 3
DISCUSSION OF CHANGES SPECIFICATION 3.3.3, ESF INSTRUMENTATION A.7 CTS Table 4.17.2, Functional Unit 2.c, requires both a CHANNEL FUNCTIONAL TEST and a CHANNEL CALIBRATION to be performed on the SIRWT Level Switches every 18 months. In the ITS, a CHANNEL CALIBRATION (SR 3.3.3.3) is only required because the definition of CHANNEL CALIBRATION explicitly encompasses the CHANNEL FUNCTIONAL TEST. Since the requirements for CHANNEL CALIBRATION include the CHANNEL FUNCTIONAL TEST, this

change is considered administrative.
MORE RESTRICTIVE CHANGES (M) M.1 CTS 3.17.2.5 contains Required Actions to be taken when the "number of OPERABLE channels is less than specified in the "Minimum OPERABLE Channels" column of Table 3.17.2. In CTS, these Required Actions are to place the reactor in HOT SHUTDOWN within 12 hours and place the reactor in a condition where the affected equipment is not required, within 48 hours. In ITS, these Required Actions have been replaced by LCO 3.0.3. Both CTS 3.17.2 and ITS 3.3.3 contain Required Actions for conditions where the
plant equipment does not meet the LCO, but still is able to provide the required safety function. In CTS, an explicit "shutdown" Required Action is also provided for equipment failures beyond this acceptable level. In ITS, since no condition is provided to address conditions beyond the acceptable level of degradation, the shutdown requirement is provided by LCO 3.0.3. Both of these Required Action treatments require the plant to be taken to a condition where the affected LCO is not applicable.
The times specified, to reach similar conditions, in ITS are shorter and therefore more restrictive. c-(
':l-The CTS "Operating Condition" definitions differ from the ITS "Mode" definitions of ~
the same name. ~ CTS Required Action 3.17.2.Sa requires the plant to be in Hot Shutdown ~ (i.e., subcritical) within 12 hours; ITS LCO 3.0.3a requires the plant to be in MODE 3 (subcritical) within 7 hours. When starting from power operation, the entry conditions for CTS Hot Shutdown and ITS MODE 3 are essentially identical; when the plant is subcritical. Since the required action is the same, and the completion time is shortened, this change is considered to be More Restrictive .
Palisades Nuclear Plant 05/30/99
ATTACHMENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.3.3, ESF INSTRUMENTATION M.1 (continued) CTS 3.17.2 is applicable when the plant is above 300°F. CTS Required Actions 3. i 7.2.Sb requires the piant to be "in a condition where the affected equipment is not required within 48 hours." For CTS LCO 3.17.2, that would be below 300°F. ITS LCO 3.0.3 requires "the plant shall be placed in a MODE or other specified condition in which the LCO is not applicable." ITS LCO allows 31 hours to be in MODE 4 (i.e., below 300°F). Again, since the required action is the same, and the completion time is shortened, this change is considered to be More Restrictive. Therefore the replacement of CTS 3.17.2.5 wording "or if the number of operable channels is less than specified in the "Minimum Operable Channels" and reliance upon the requirements of ITS LCO 3.0.3 will result in a More Restrictive Change. This change is consistent with NUREG-1432. M.2 Both CTS 3.17.2.5 and ITS 3.3.3 Condition D contain Required Actions to be taken when any Required Action is not completed and the associated completion time has expired. In CTS, these Required Actions are to place the reactor in HOT SHUTDOWN within 12 hours and place the reactor in a condition where the affected
*equipment is not required, within 48 hours. In ITS LCO 3.3.3 these Required Actions are to be in MODE 3 within 6 hours, and to be in MODE 4 within 12 hours (for functions which are addressed by CTS LCO 3.17.2).
The CTS "Operating Condition" definitions differ from the ITS "MODE" definitions of the same name. CTS Required Actions 3.17.2.5a requires the plant to be in Hot Shutdown (i.e., subcritical) within 12 hours; ITS Required Action 3.3.3 D. l requires the plant to be in MODE 3 (subcritical) within 6 hours. When starting from power operation, the entry conditions for CTS Hot Shutdown and ITS MODE 3 are essentially identical; when the plant is subcritical. Since the required action is the same, and the completion time is shortened, this change is considered to be More Restrictive. CTS 3.17.2 is applicable when the plant is above 300°F. CTS Required Actions 3.17.2.5b requires the plant to be placed "in a condition where the affected equipment is not required within 48 hours." For CTS LCO 3.17 .2, that would be below 300°F. ITS Required Action 3.3.3 D.2 requires to be in MODE 4 (i.e., below 300°F) within 30 hours. Again, since the required action is the same, and the completion time is shortened, this change is considered to be More Restrictive. Therefore the replacement of the CTS 3. l7.2.5 requirements to be taken "when any Required Action is not completed and the associated completion time has expired" by
ITS Required Actions 3.3.3 D. l and D.2 will result in a More Restrictive Change .
This change is consistent with NUREG-1432. Palisades Nuclear Plant 15-Ju Page 4 of 8 05/30/99
ATTACHMENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.3.3, ESF INSTRUMENTATION M.3 CTS 3.17 .3.5 contains Required Actions to be taken when the "number of OPERABLE channels is less than specified in the "Minimum OPERABLE Channels" column of Table 3.17.3. In CTS, these Required Actions are to place the reactor in HOT SHUTDOWN within 12 hours and place the reactor in a condition where the affected equipment is not required, within 48 hours. In ITS, these Required Actions have been replaced by LCO 3.0.3. Both CTS 3.17.3 and ITS 3.3.3 contain Required Actions for conditions where the plant equipment does not meet the LCO, but still is able to provide the required safety function. In CTS, an explicit "shutdown" Required Action is also provided for equipment failures beyond this acceptable level. In ITS, since no condition is provided to address conditions beyond the acceptable level of degradation, the shutdown requirement is provided by LCO 3.0.3. Both of these Required Action treatments require the plant to be taken to a condition where the affected LCO is not applicable. The times specified, to reach similar conditions, in ITS are shorter and therefore more restrictive. The CTS "Operating Condition" definitions differ from the ITS "MODE" definitions of the same name .
CTS Required Action 3.17 .3.5a requires the plant to be in Hot Shutdown (i.e., subcritical) within 12 hours; ITS LCO 3.0.3a requires the plant to be in MODE 3 (subcritical) within 7 hours. When starting from power operation, the entry conditions for CTS Hot Shutdown and ITS MODE 3 are essentially identical; when the plant is c:>--
subcritical. Since the required action is the same, and the completion time is shortened, this change is considered to be More Restrictive. ' CTS 3.17.3 is applicable when the plant is above Cold Shutdown (i.e., above 210°F). CTS Required Actions 3.17.3.5b requires the plant to be "in a condition where the affected equipment is not required within 48 hours." For CTS LCO 3.17.3, that would be below 210°F. ITS LCO 3.0.3 requires "the plant shall be placed in a MODE or other specified condition in which the LCO is not applicable." ITS LCO allows 37 hours to be in MODE 5 (i.e., below 200°F). Again, since the required action is essentially the same, and the completion time is shortened, this change is considered to be More Restrictive. Therefore the replacement of CTS 3.17.3.5 wording "or if the number of operable channels is less than specified in the "Minimum Operable Channels" and reliance upon the requirements of ITS LCO 3.0.3 will result in a More Restrictive Change. This change is consistent with NUREG-1432 .
Palisades Nuclear Plant Page 5 of 8 05/30/99
ATTACHMENT 3
M.4 DISCUSSION OF CHANGES SPECIFICATION 3.3.3, ESF INSTRUMENTATION Both CTS 3.17.3.5 and ITS 3.3.3 Condition E contain Required Actions to be taken when any Required Action is not completed and the associated completion time has expired. In CTS, these Required Actions are to place the reactor in HOT SHUTDOWN within 12 hours and place the reactor in a condition where the affected equipment is not required, within 48 hours. In ITS LCO 3.3.3 these Required Actions are to be in MODE 3 within 6 hours, and to be in MODE 5 within 12 hours (for functions which are addressed by CTS LCO 3 .17. 3).
The CTS "Operating Condition" definitions differ from the ITS "MODE" definitions of the same name. CTS Required Actions 3.17.3.5a requires the plant to be in Hot Shutdown (i.e., subcritical) within 12 hours; ITS Required Action 3. 3. 3 E.1 requires the plant to be in MODE 3 (subcritical) within 6 hours. When starting from power operation, the entry conditions for CTS Hot Shutdown and ITS MODE 3 are essentially identical; when the plant is subcritical. Since the required action is the same, and the completion time is shortened, this change is considered to be More Restrictive. CTS 3.17.3 is applicable when the plant is above 210°F. CTS Required Actions 3. l 7.3.5b requires the plant to be placed "in a condition where the affected equipment is not required within 48 hours." For CTS LCO 3 .17. 3, that would be below 210°F. ITS Required Action 3.3.3 E.2 requires to be in MODE 5 (i.e., below
.200°F) within 36 hours. Again, since the required action is essentially the same, and the completion time is shortened, this change is considered to be More Restrictive.
Therefore the replacement of the CTS 3.17.3.5 requirements to be taken "when any Required Action is not completed and the associated completion time has expired" by ITS Required Actions 3.3.3 E.1 and E.2 will result in a More Restrictive Change. This change is consistent with NUREG-1432. M.5 The Allowable Value for the upper limit on CHP actuation has been reduced from 4.4 psig to 4.3 psig to assure that there is adequate margin for instrument tolerances. During review of the allowed values for the Containment High Pressure (CHP) Engineered Safety Feature (ESF) actuation setpoint for the Improved Technical Specification project, it was determined that the upper limit of 4.4 psig specified in current Technical Specification Table 3.16-Item 2 was not consistent with the assumptions of the FSAR Chapter 14.18 containment response analyses. The containment response analyses assume that the CHP ESF actuation occurs prior to containment pressure exceeding 4.3 psig,. when allowance is made for the allowed "as found" calibration tolerances for the actual CHP ESF pressure switches. This change
is more restrictive with respect to requiring a more rapid actuation following postulated LOCA or MSLB event. The 4.3 psig setting limit is that currently used by the plant, and is the setting required by the Operating Requirements Manual.
Palisades Nuclear Plant Page 6 of 8 05/30/99 Z/5-ff)
A.7 SPECIFICATION 3.3.4, ESF LOGIC AND MANUAL INITIATION CTS Table 4.17 .2 has been revised by adding a Note to the 92 day CHANNEL ATTAC1'ENT 3 DISCUSSION OF CHANGES FUNCTIONAL TEST which states that testing of the Actuation Logic shall include verification of the proper operation of each initiation relay. This change is considered to be administrative, in that it only provides clarification and does not alter any technical requirements.
MORE RESTRICTIVE CHANGES (M) M.l S 3.17.2.5 requires specific Actions be taken when the "number of OPERABLE cha els is less than specified in the "Minimum OPERABLE Channels" column of Table 7.2," and when "any action required by CTS 3.17.2 is not met AND th associate ompletion time has expired." These Actions are to place the reac in HOT SHUT WN within 12 hours, and in a condition where the affecte quipment is not required wi
48 hours. Based on the current Applicability of C 3.17. 2, the condition where the ected equipment is not required is when the S temperature less than 300°F. In th TS, the Required Action when the "Re rred Action and

associated Completion T are not met" is to enter ITS 3. . Condition B, which requires that the plant be in DE 3 in 6 hours and in E 4 in 30 hours.
CTS 3.17.3.5 requires specific Actio channels is less than specified in the "........ ,..,.......... Table 3.17.3, and when "any action requir y CTS 3.17.3 is not met AND the associated completion time has expired." he Actions are to place the reactor in HOT SHUTDOWN within 12 hours nd in a con "tion where the affected equipment is not required within 48 hours. Ba Cl on the current licability of CTS 3 .17. 3, the condition where the affected e ipment is not required i hen the PCS is at or below COLD SHUTDOWN. In e ITS, the Required Action, fo ontainment High Pressure and Containm t High Radiation, when the "Require ction and associated Completion Times not met" is to enter ITS 3.3.4 Condition C, hich requires that the plant be in DE 3 in 6 hours and in MODE 5 in 36 hours, and e Required Action, for S am Generator Low Pressure, when the "Required Action d associated Completi Times are not met" is to enter ITS 3.3.4 Condition B, which re ires that the pl be in MODE 3 in 6 hours and in MODE 5 in 30 hours (see DOC L.1 low) .
Palisades Nuclear Plant Page 3 of 5 01/20/98

ATTACHMENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.3.4, ESF LOGIC AND MANUAL INITIATION e CTS COLD SHUTDOWN, the temperature requirement is less 210°F, versus ITS MODE 5 which has a temperature requirement of less an 200°F. The difference o degrees is negligible and has no significant imp on operations. The other parameter "ch is common between the CTS HOTS TDOWN and COLD SHUTDOWN modes the corresponding ITS MOD 3, 4 and 5 is the reactivity condition. The ITS MOD 4, and 5 are define , as a reference point, by a reactivity condition of ~ff < 0. 99. Howe in ITS S on 3 .1, the equivalent amount of SHUTDOWN MARGIN is required specified in the CTS definitions of HOT SHUTDOWN and COLD SHUTD erefore, the amount of SHUTDOWN MARGIN is considered to be s e when the r irements of proposed ITS 3 .1 are considered.
Since this chang ill require the plant to be in a lower MO in a shorter time frame, this change
onsidered more restrictive. This change is approp

e because it continu o provide adequate time for an orderly plant shutdown with challenging pl systems. This change imposes additional restrictions on plant operat1 and is onsistent with NUREG-1432.
-See revise.cl M. \
LESS RESTRICTIVE CHANGES - REMOVAL OF DETAILS TO LICENSEE CONTROLLED DOCUMENTS (LA) LA.I CTS Tables 3.17.2 and 4.17.2 Function l.b and l.c contain details of the scope of logic circuits associated with these Functions. Specifically, the CTS Function
1. b lists "Initiation, Actuation, and low pressure block auto reset," while Function l.c lists "5P Relay Outputs." In the ITS, proposed ITS Table 3.3.4-1 Function l.b requires that the SIS Actuation Logic (Pressurizer Pressure - Low) be Operable while proposed Function l.c requires that the SIS Actuation Logic (CHP) be Operable. The details of what constitutes an Operable SIS Actuation Logic (Pressure - Low) and SIS Actuation Logic (CHP) are specified in the Bases. The Bases state that "an Actuation Logic Channel consists of all circuitry housed within the actuation logic circuits, including the initiating relays contacts responsible for actuating the ESF equipment." For the SIS Actuation logic (Pressurizer Pressure - Low) the automatic bypass removal feature is also required .
Palisades Nuclear Plant Page 4 of 5 11~-o 01/20/98
ATTACHMENT 3
DISCUSSION OF CHANGES SPECIFICATION 3.3.4, ESF LOGIC AND MANUAL INITIATION MORE RESTRICTIVE CHANGES (M)
M.1 . CTS 3.17.2.5 contains Required Actions to be taken when the "number of OPERABLE channels is less than specified in the "Minimum OPERABLE Channels" column of Table 3.17.2. In CTS, these Required Actions are to place the reactor in HOT SHUTDOWN within 12 hours and place the reactor in a condition where the affected equipment is not required, within 48 hours. In ITS, these Required Actions have been replaced by LCO 3.0.3. Both CTS 3.17.2 and ITS 3.3.4 contain Required Actions for conditions where the plant equipment does not meet the LCO, but still is able to provide the required safety function. In CTS, an explicit "shutdown" Required Action is also provided for equipment failures beyond this acceptable level. In ITS, since no condition is provided to address conditions beyond the acceptable level of degradation, the shutdown requirement is provided by LCO 3.0.3. Both of these Required Action treatments require the plant to be taken to a condition where the affected LCO is not applicable. The times specified, to reach similar conditions, in ITS are shorter and therefore more restrictive.
The CTS "Operating Condition" definitions differ from the ITS "MODE" definitions of the same name.
CTS Required Action 3.17.2.5a requires the plant to be in Hot Shutdown (i.e., subcritical) within 12 ho~rs; ITS LCO 3.0.3a requires the plantto be in MODE 3 (subcritical) within 7 hours. When starting from power operation, the entry conditions for CTS Hot Shutdown and ITS MODE 3 are essentially identical; when the plant is subcritical. Since the required action is the same, and the completion time is shortened, this change is considered to be More Restrictive. CTS 3.17.2 is applicable when the plant is above 300°F. CTS Required ~ Actions 3.17 .2.5b requires the plant to be "in a condition where the affected equipment is not required within 48 hours." For CTS LCO 3.17.2, that would be below 300°F. ITS LCO 3.0.3 requires "the plant shall be placed in a MODE or other specified condition in which the LCO is not applicable." ITS LCO allows 31 hours to be in MODE 4 (i.e., below 300°F). Again, since the required action is the same, and the completion time is shortened, this change is considered to be More Restrictive. Therefore the replacement of CTS 3.17 .2.5 wording "or if the number of operable channels is less than specified in the "Mipimum Operable Channels" and reliance upon the requirements of ITS LCO 3.0.3 will result in a More Restrictive Change. This change is consistent with NUREG-1432 . Palisades Nuclear Plant 05/30/99
ATTACHMENT 3
M.2 DISCUSSION OF CHANGES SPECIFICATION 3.3.4, ESF LOGIC AND MANUAL INITIATION Both CTS 3.17.2.5 and ITS 3.3.4 Condition B contain Required Actions to be taken when any Required Action is not completed and the associated completion time has expired. In CTS, these Required Actions are to place the reactor in HOT SHUTDOWN within 12 hours and place the reactor in a condition where the affected equipment is not required, within 48 hours. In ITS LCO 3.3.4 these Required Actions are to be in MODE 3 within 6 hours, and to be in MODE 4 within 12 hours (for functions which are addressed by CTS LCO 3.17.2).
The CTS "Operating Condition" definitions differ from the ITS "MODE" definitions of the same name. CTS Required Actions 3 .17 .2.5a requires the plant to be in Hot Shutdown (i.e., subcritical) within 12 hours; ITS Required Action 3.3.4 B.1 requires the plant to be in MODE 3 (subcritical) within 6 hours. When starting from power operation, the entry conditions for CTS Hot Shutdown and ITS MODE 3 are essentially identical; when the plant is subcritical. Since the required action is the same, and the completion time is shortened, this change is considered to be More Restrictive. c-l
r CTS 3.17.2 is applicable when the plant is above 300°F. CTS Required Actions 3.17 .2.5b requires the plant to be placed "in a condition where the affected equipment is not required within 48 hours." For CTS LCO 3.17 .2, that would be below 300°F. ITS Required Action 3.3.4 B.2 requires to be in MODE 4 (i.e., below 300°F) within 30 hours. Again, since the required action is the same, and the completion time is shortened, this change is considered to be More Restrictive.
Therefore the replacement of the CTS 3.17.2.5 requirements to be taken "when any Required Action is not completed and the associated completion time has expired" by ITS Required Actions 3.3.4 B. l and B.2 will result in a More Restrictive Change. This change is consistent with NUREG-1432 .
Palisades Nuclear Plant 05/30/99
ATTACHMENT 3
M.3 DISCUSSION OF CHANGES SPECIFICATION 3.3.4, ESF LOGIC AND MANUAL INITIATION CTS 3.17 .3.5 contains Required Actions to be taken when the "number of OPERABLE channels is less than specified in the "Minimum OPERABLE Channels" column of Table 3.17.3. In CTS, these Required Actions are to olace the reactor in HOT
- ~
SHUTDOWN within 12 hours and place the reactor in a condition where the affected equipment is not required, within 48 hours. In ITS, these Required Actions have been replaced by LCO 3.0.3. Both CTS 3.17.3 and ITS 3.3.4 contain Required Actions for conditions where the plant equipment does not meet the LCO, but still is able to provide the required safety function. In CTS, an explicit "shutdown" Required Action is also provided for equipment failures beyond this acceptable level. In ITS, since no condition is provided to address conditions beyond the acceptable level of degradation, the shutdown requirement is provided by LCO 3.0.3. Both of these Required Action treatments require the plant to be taken to a condition where the affected LCO is not applicable. The times specified, to reach similar conditions, in ITS are shorter and therefore more restrictive. The CTS "Operating Condition" definitions differ from the ITS "MODE" definitions of the same name . CTS Required Action 3.17.3.5a requires the plant to be in Hot Shutdown (i.e., subcritical) within 12 hours; ITS LCO 3.0.3a requires the plant to be in MODE 3 (subcritical) within 7 hours. When starting from power operation, the entry conditions for CTS Hot Shutdown and ITS MODE 3 are essentially identical; when the plant is subcritical. Since the required action is the same, and the completion time is shortened, this change is considered to be More Restrictive. CTS 3.17.3 is applicable when the plant is above Cold Shutdown (i.e., above 210°F). -
~
CTS Required Actions 3.17.3.5b requires the plant to be "in a condition where the affected equipment is not required within 48 hours." For CTS LCO 3 .17. 3, that would be below 210°F. ITS LCO 3.0.3 requires "the plant shall be placed in a MODE or other specified condition in which the LCO is not applicable." ITS LCO allows 37 hours to be in MODE 5 (i.e., below 200°F). Again, since the required action is essentially the same, and the completion time is shortened, this change is considered to be More Restrictive. Therefore the replacement of CTS 3.17.3.5 wording "or if the number of operable channels is less than specified in the "Minimum Operable Channels" and reliance upon the requirements of ITS LCO 3.0.3 will result in a More Restrictive Change. This change is consistent with NUREG-1432 .
Palisades Nuclear Plant Page 5 of 8 05/30/99
ATTACHMENT 3
M.4 DISCUSSION OF CHANGES SPECIFICATION 3.3.4, ESF LOGIC AND MANUAL INITIATION Both CTS 3.17.3.5 and ITS 3.3.4 Condition C contain Required Actions to be taken when any Required Action is not completed and the associated completion time has expired. In CTS, these Required Actions are to place the reactor in HOT SHUTDOWN within 12 hours and place the reactor in a condition where the affected equipment is not required, within 48 hours. In ITS LCO 3.3.4 these Required Actions are to be in MODE 3 within 6 hours, and to be in MODE 5 within 12 hours (for functions which are addressed by CTS LCO 3 .17. 3).
The CTS "Operating Condition" definitions differ from the ITS "MODE" definitions of the same name. CTS Required Actions 3.17.3.5a requires the plant to be in Hot Shutdown (i.e., subcritical) within 12 hours; ITS Required Action 3.3.4 C.1 requires the plant to be in MODE 3 (subcritical) within 6 hours. When starting from power operation, the entry conditions for CTS Hot Shutdown and ITS MODE 3 are essentially identical; when the plant is subcritical. Since the required action is the same, and the completion time is shortened, this change is considered to be More Restrictive. CTS 3.17.3 is applicabl~ when the plant is above 210°F. CTS Required
Actions 3.17.3.5b requires the plant to be placed "in a condition where the affected equipment is not required within 48 hours." For CTS LCO 3 .17. 3, that would be below 210°F. ITS Required Action 3.3.4 C.2 requires to be in MODE 5 (i.e., below 200°F) within 36 hours. Again, since the required action is essentially the same, and the completion time is shortened, this change is considered to be More Restrictive.
Therefore the replacement of the CTS 3.17.3.5 requirements to be taken "when any Required Action is not completed and the associated completion time has expired" by ITS Required Actions 3.3.4 C. l and C.2 will result in a More Restrictive Change. This change is consistent with NUREG-1432 .
Palisades Nuclear Plant Page 6 of 8 05/30/99
COl'<l_VERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO JANUARY 6, 1999 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.3, INSTRUMENTATION
NRC REQUEST:
3.3.4-10 JFD 9 CTS Table 3.17.3 CTS Table 3.17.3 indicates a permissible operational bypass for a Steam Generator Low Pressure Function of "less than 550 psig Steam Pressure". ITS Table 3.3.4-1, Footnote (c) changes the allowed bypass pressure to "less than 565 psia Steam Generator Pressure". Comment: It is not clear that the CTS bypass pressure is measured by Steam Generator Pressure instrumentation or Steam Line Pressure instrumentation. Should the ITS assume that the ITS bypass pressure is measured by the Steam Generator Pressure instrumentation? The changed values are not discussed or justified. Provide discussion and justification for the change including any difference in measured parameters, monitoring instrumentation, scaling factors, and why the engineering units of measure are changed from psig to psia. Consumers Energy Response: Both the SGLP actuating bistable and the Bypass permissive/removal bistable, for a given channel, monitor the same instrument channel. That instrument channel measures steam generator pressure in psia. The CTS and ITS specified setpoints, 550 psig and 565 psia respectively, are equivalent. The change was made for consistency between specified settings and to specify the required settings in the units presented on the instrument.
Affected Submittal Pages:
No page changes .
46
I I CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO JANUARY 6, 1999 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.3, INSTRUMENTATION
NRC REQUEST:
3.3.4-12 Resolved during 10/27/98 meeting; Consumers Energy not required to respond. See Comment 3.3.4-6 NRC REQUEST: 3.3.5-i iTS 3.3.5, DOC M.1 JFD 9 The ITS proposes to adopt the STS LCO for LOP DG start. The M.1 DOC discusses the applicability for this LCO as MODES 1, 2, 3, 4 and when the associated DG is required to be operable. Comment: The ITS markup changes the STS Applicability requirements from MODES 1, 2, 3, 4 and when associated DG is required to be OPERABLE by LCO 3.8.2, "AC Sources-Shutdown." The proposed ITS represents a generic change to the STS and is not acceptable. Revise the ITS Applicability for LCO 3.3.5 to be consistent with NUREG-1432, Rev. 1. Consumers Energy Response: ITS proposes LCO 3.3.5, DG-Undervoltage Start, as a new LCO. There is no explicit CTS requirement for this equipment to be operable. Since this equipment is necessary for the DG to perform its specified function, CTS would require the Undervoltage Start circuits to be operable whenever the associated DG is required to be operable. That is the proposed ITS applicability .
ITS LCO 3.3.5 is not unique in specifying the applicability of support equipment as: whenever
[the supported equipment] is required to be operable:
1) Both STS and ITS LCOs 3.8.3, which also addresses DG support equipment, specify their applicability as: "When associated DG is required to be OPERABLE."

2) Both STS and ITS LCOs 3.8.6, which addresses DC electrical power source support equipment specify their applicability as: "When associated DC electrical power source(s) are required to be OPERABLE."

3) CTS LCOs 3.7.3 and 3.7.6, which are the equivalent of ITS 3.8.3 and 3.8.6, use similar applicability wording.
Since ITS LCO 3.8.1 requires both DGs to be operable in Modes 1, 2, 3, and 4, and only LCOs 3.8.1 and 3.8.2 address DG operability, the two ways of stating the applicability: STS 3.3.6: MODES 1, 2, 3, and 4, [and] When associated DG is required to be OPERABLE by LCO 3.8.2, "AC Sources-Shutdown." ITS 3.3.5: When associated DG is required to be OPERABLE. are equivalent. The variation which is proposed in ITS 3.3.5 is consistent with other ITS support system LCOs, is familiar to the Palisades Plant staff due to its use in CTS, and is more concise.
Therefore, no change has been made to the proposed applicability.
Affected Submittal Pages: No page changes. 47
CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO JANUARY 6, 1999 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.3, INSTRUMENTATION
NRC REQUEST:
3.3.5-02 Comment: STS SR 3.3.6.3 states:" Perform CHANNEL CALIBRATION with setpoint (emphasis added) Allowable Values as follows:" ITS SR 3.3.5.1 deletes the word "setpoint" without any discussion or justification. Provide discussion and justification for the change to STS wording.
**Discuss with Comment 3.3.5-03 Consumers Energy Response:
See response to RAI 3.3.5-03 Affected Submittal Pages: See response to RAI 3.3.5-03 48
CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO JANUARY 6, 1999 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.3, INSTRUMENTATION
NRC REQUEST:
3.3.5-03 Comment: STS SR 3.3.6.3 provides Allowable Values for the DG-LOVS Degraded Voltage, Loss of Voltage and Time Delay Functions. The Allowable Values are stated with an upper and lower value (i.e. Degraded Voltage Function ~ 3180 V and :.,; 3220 V). The corresponding SR in the ITS is SR 3.3.5.1. The Allowable Value in ITS 3.3.5.1 for the Degraded Voltage Function is
~2184 V, providing no setting tolerance around the Allowable Value. Provide Allowable Values for DG-UV Start Degraded Voltage, Loss of Voltage and Time Delay functions with upper and lower setting limits, consistent with the STS SR 3.3.6.3 presentation, and provide a basis for the Allowable Values.
Consumers Energy Response: ITS SR 3.3.5.1 has been revised in response to comments in RAls 3.3.5-02 and 3.3.5-03. The proposed Channel Calibration SR has been revised to more closely match the STS equivalent. The term "Allowable Value" is not used because the required settings for the Loss of voltage function have not been determined in accordance with "Allowable Value" methodology. Affected Submittal Pages:
Att 1, ITS 3.3.5, page 3.3.5-1 Att 5, NUREG, page 3.3-29 49
DG - UV Start 3.3.5 3.3 INSTRUMENTATION 3.3.5 Diesel Generator (DG) - Undervoltage Start (UV Start) LCD 3.3.5 ~annelS)of Loss of Voltage Function a~~~~anne\5)lf Degraded Voltage Function auto-initiation instrumentation DG shal 1 be OPERABLE. AND ASSOC.IA.TCD LDE,1c. c~A.N~El.S Fo'e EAC.t+ APPLICABILITY: When associated DG is required to be OPERABLE. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more Functions A.l Enter applicable Immediately with one channel per Conditions and , DG inoperable. Required Actions for the associated DG made inoperable by
DG - UV Start instrumentation.
PEKfbeW\ to.. C&:-\Mit.lEL FutJc.TlOt'1AL TEST\ 0#.l EAC.t-i Db-U\} srAgr LoC::>IC:. CHANNEL. SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 18 months b. Time delay:r~ 8~15 seconds at 1400 VIJ. an (~ 5,45 SEC:OtJl)S A>JD)
Palisades Nuclear Plant Amendment No. 01/20/98

SURVEILLANCE RE UIREHENTS .
SURVEILLANCE FREQUENCY J SR
>.* Perform CHANNEL FUNCTIONAL TEST(ou E:At:~)
(PG,-u\J -sr~er La;,1c. cH~..ifJa;) Ok.I E,;A.c.t\ ~ (%- VOt..~C:*- Aklt> ~~P VOLTA6'Ef- C~l'.~tJc\...
CEOG STS 3.3-29 Rev l, 04/07/95
CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO JANUARY 6, 1999 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.3, INSTRUMENTATION
NRC REQUEST:
3.3.5-04 ITS LCO 3.3.5 Bases Markup Bases insert 1 describes the two levels of undervoltage protection relays installed on each 2,400 volt bus (1 C and 1 D). The Bases state that each relay measures voltage on all three phases, and protects against sudden voltage loss on the corresponding bus using a three-out-of-three phases coincident logic. Once the logic is made up the actuation relay will trip its respective incoming bus circuit breakers, start its associated DG, initiate bus load shedding, and activate annunciators in the control room. The Bases describes a three channel detection system with a single actuation relay for each bus. Comment: Identify the number of installed calibrated devices that sense voltage loss for each bus. Identify the number of channels associated with these devices. Provide information regarding the reliability of the DG-UV Start instrument channels for the proposed 18 month calibration interval and relate channel drift characteristics to the need for a periodic channel functional test. Revise LCO 3.3.5 channels requirements to be consistent with TS defined channels. Consumers Energy Response: ITS LCO 3.3.5 and the associated Bases have been revised to clarify both the required circuit components and the required surveillance for those components.
Affected Submittal Pages:
Att 1, ITS 3.3.5, page 3.3.5-1 Att 5, NUREG, page 3.3-27 Att 5, NUREG, page 3.3-29 Att 6, JFD 3.3.6, page 2 of 2 (former submittal) Att 6, JFD 3.3.6, page 2 of 2 50
OG - UV Start
3.3 INSTRUMENTATION 3.3.5 Diesel Generator (DG) - Undervoltage Start (UV Start) 3.3.5 LCO 3.3.5 ~anne1$)of Loss of Voltage Function a~;"~~anne\§)>f Degraded Voltage Function auto-initiation instrumentation DG shall be OPERABLE. AND ASSOC.IA.Tf:D Lc6ac. C~A.tHJCLS Fo'e 'E:AC.t+
APPLICABILITY: When associated DG is required to be OPERABLE. ACTIONS CONDITION REQUIRED ACTION COMPLETiON TIME A. One or more Funct i ans A.1 Enter applicable Immediately with one channel per Conditions and ~ DG inoperable. Required Actions for the associated DG made inoperable by DG - UV Start instrumentation. PRR::>e~ t>... C'-\~t-lEl.. FutJc..T10"1AL OIJ EAC.t-\ Db-U\} srM~:r LoC.:::OIC.. CHANNEL. TEST\ SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY 18 months
de a : < seconds;~
,. 0,15' S'E-GOJJDS All.l'D 1780
b. Loss o Vo ta e Function ~ V lo/'oj(Qj{jil A~D ~ I gyo " .
Time delay:r~ 8~15 seconds at 1400 VVJ. an (~ ;,45 SEC.OIJt)S A~D )
Palisades Nuclear Plant 9
.:fD -(J_/
3.3.5-1 Amendment No. 01/20/98
iN:~::~~::::tor (DGH.o0~:r:~~2rt (~ .
P?J ~tz. ~-if,-~-~ LCO 3.3~-- channels of Loss of Voltage Function and c annels of Degraded Voltage Function auto-initiatan instrumentation DG shall be OPERABLE.
~..it:> 1>..s-soc.1~re-i:> t.061C- CM~..,.~~ ~oc ~~c.."4 APPLICABILITY: CMObES 1).,.2, 'h anib4) ~
When ~sociated DG 1stdOW~ed to be OPERABLtvlQNC6'J.&;Z) AC'--SDurc~hu n. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME
A. One or ore Functions with on channel per DG inope 1e.
A~
**-\_,
AND '" Place channel in bypass or trip. I hour~ A.2.1 R store channel to [48] hours O BLE status. in (continued}
CEOG STS 3.3-27 Rev 1, 04/07/95

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY J
SR Oki 5::A.ot l..o% O'F-VOL.T'Af:>li- MlD ~ll'EP VOLTAlo&=- Cl-\A.tJt.JE\-
CEOG STS 3.3-29
.5'0 -0 Rev 1, 04/07/95
Change Discussion ATTAC1'ENT 6 JUSTIFICATION FOR DEVIATIONS SPECIFICATION 3.3.6, DG- LOVS 8.
9. Applicability of LCO 3.3.5 is revised to read, "When associated DG is required to be OPERABLE." This terminology is consistent with NUREG-1432 Specification 3.8.3, "Diesel Fuel Oil, Lube Oil, and Starting Air." '

10. The word Analog was removed from the title of each Specification in Section 3. 3.
Palisades is an Analog plant and specifically listing this in the title is unnecessary. Analog was placed in NUREG-1432 to distinguish between Analog and Digital Specifications.
11. ISTS 3.3.2 is not included in the ITS, therefore subsequent Specifications have been renumbered accordingly .
Palisades Nuclear Plant Page 2of2 so-cJ 01/20/98
ATTACHMENT 6
Change Discussion JUSTIFICATION FOR DEVIATIONS SPECIFICATION 3.3.6, DG - LOVS
8. NUREG-1432, Specification 3.3.6, SR 3.3.6.1 has been deleted. The Palisades Nuclear Plant instrumentation associated with Diesel Generator (DG) - Loss of Voltage Start (LOVS)design does not include metering in the control room associated with the undervoltage sensing channels. The remaining ITS SRs have been editorially revised to indicate that the Channel Functional Test applies to the logic channels and the Channel Calibration, to the undervoltage sensors.

9. Applicability of LCO 3.3.5 is revised to read, "When associated DG is required to be OPERABLE." This terminology is consistent with NUREG-1432 Specification 3. 8. 3, "Diesel Fuel Oil, Lube Oil, and Starting Air."

10. The word Analog was removed from the title of each Specification in Section 3.3.
Palisades is an Analog plant and specifically listing this in the title is unnecessary. Analog was placed in NUREG-1432 to distinguish between Analog and Digital Specifications.
11. ISTS 3.3.2 is not included in the ITS, therefore subsequent Specifications have been
renumbered accordingly .

Palisades Nuclear Plant
$D-Page 2of2 05/30/99
CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO JANUARY 6, 1999 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.3, INSTRUMENTATION NRC REQUEST: 3.3.6-01 JFD 9 provides discussion and justification for deleting the automatic actuation logic and manual trip functions from STS 3.3.7. The discussion states that each of the two refueling CHR channels have an associated logic circuit, therefore, the licensee considers the channel to include the actuation circuit. The licensee also states that a manual trip feature is included in the design but its function is not assumed in the safety analysis. Comment 1: It is not clear why the manual trip function is not included in ITS 3.3.6, because this system is classified as a Safety System, according to IEEE 603. Additionally, certain instrumentation features are integral to the format and content of this and other LCOs in the STS, thus the manual function and automatic actuation logic should be included in ITS 3.3.6. Are there failure modes of the manual circuitry that could render one or both channels inoperable? Comment 2: The proposed SRs include CHANNEL CHECK, CHANNEL FUNCTIONAL TEST and CHANNEL CALIBRATION. Which channel test verifies the allowable value? Why isn't the allowable value included in the TS? Which channel test verifies the channel logic? Consumers Energy Response: ITS LCO 3.3.6 and the associated Bases have been revised to explicitly require manual initiation capability . While logic circuit failures could render either train of CHR inoperable, a failure of the manual initiation portion of the circuit would not affect the automatic initiation. The Channel Calibration verifies the instrument trip setpoints. No Allowable Value is specified since there is no analysis which determines any required setting. There is no specified setting in CTS. The instrument is set to initiate CHR if radiation levels exceed expected background. Since expected background may change with the planned activity, no universal setting is appropriate. Certain infrequent activities, such as moving the core barrel, are expected to create considerably higher than normal radiation fields in the vicinity of the detectors for short periods .
51 (continued)
CONVERSION TO IMPROVED TECHNICAL S~ECIFICATIONS RESPONSE TO JANUARY 6, 1999 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.3, INSTRUMENTATION
NRC REQUEST:
3.3.6-01 (continued) Affected Submittal Pages: Att 1, ITS 3.3.6, page 3.3.6-1 Att 1, ITS 3.3.6, page 3.3.6-1 insert (2 pages) Att 3, CTS, page 3-46 Att 3, DOC 3.3.6, page 2 of 4 (former submittal) Att 3, DOC 3.3.6, page 4 of 4 (former submittal) Att 3, DOC 3.3.6, page 2 of 4 Att 3, DOC 3.3.6, page 4 of 4 Att 4, NSHC 3.3.6, page 3 of 4 Att 4, NSHC 3.3.6, page 4 of 4 Att 4, NSHC 3.3.6, page 2 of 4 Att 4, NSHC 3.3.6, page 3 of 4 Att 4, NSHC 3.3.6, page 4 of 4 Att 5, NUREG, 3.3.7, 3.3-30 Att 5, NUREG, 3.3.7, 3.3-31 Att 5, NUREG, 3.3.7, 3.3-32 Att 6, JFD 3.3.7, page 2 of 3 (former submittal) Att 6, JFD 3.3.7, page 3 of 3 (former submittal) Att 6, JFD 3.3.7, page 2 of 2
52

Palisades Nuclear Plant ltJS"E~
5q-o.- 3> 3.3.6-1 Amendment No. 01/20/98
3.3 INSTRUMENTATION INSERT #3 LCO 3. 3. 6 fi3.
3.3.6 Refueling Containment High Radiation (CHR} Instrumentation I o;: '2. LCO 3.3.6 Two Refueling CHR Automatic Actuation Function channels and two CHR Manual Actuation Function channels shall be OPERABLE. APPLICABILITY: During CORE ALTERATIONS, During movement of irradiated fuel assemblies within containment. ACTIONS CONDITION REQUIRED ACTION COMP LET ION TIME A. One or more Functions A. l Place the affected 4 hours with one channel channel in trip. inoperable. OR
A.2.1 AND.
Suspend CORE ALTERATIONS. 4 hours A.2.2 Suspend movement of 4 hours irradiated fuel asemblies within containment. B. One or more Functions B.l Suspend CORE Immediately with two channels ALTERATIONS . inoperable. AND B.2 Suspend movement of Immediately irradiated fuel assemblies within
containment .
SURVEILLANCE REQUIREMENTS
SR 3.3.6.1 SURVEILLANCE Perfonn a CHANNEL CHECK of each refueling FREQUENCY 12 hours CHR monitor channel.
SR 3.3.6.2 Perfonn a CHANNEL FUNCTIONAL TEST of each 31 days refueling CHR monitor channel. 60 mun SR 3.3.6~Perfonn a CHANNEL CALIBRATION~ueling 18 months CHR monitor channel. SR 3.3.6.~ Perfonn a CHANNEL FUNCTIONAL TEST of each 18 months CHR Manual Initiation channel.
==========================
(;,fl~:~f\o\tjo\..t ~*~~ ~1..6.~*+ro . . {cw l .. Nv11""e **:+r:...f?i.~ ITS 3.8 REFUELING QP}QAJ?OR~ tl,U\'"I..,~ .Lee~ A~A-11~1-.Jf tJ.Hd mot'l.-Jt<<IAI
, 4 -f1J l'val w1-l-'1 ...... e1ttnl ,.,.~.a.
Ob1ect1ye To 'm.tnimize \be poss\1>111ty o~an acci~nt occur)tjng dur operat-1.ons that- could~fect oUb-:lic heafti\ and sa~y. 3.8.1 Specifications
~ollowiltg-.conditio shal 1'U(_ sat1s oerittons *
a. One door o e emergency air oc s a e .
Whenever both doors of the personnel air lock are open during refueling operations, the equipment door shall be open and the ventilating system and charcoal filter in the fuel storage building shall be operating . d.
e. Whenever core geometry is being changed, neutron flux shall be continuously monitored by at least two source range neutron monitors, with each monitor providing continuous visual indication in the control room. When core geometry is not being changed, at least one source ran e neutron monitor shal

f. east one shutdown cooling pump and heat exchanger shall be operation. ----~-----------
3-46
5~-d Amendment No. 34 January 27, 1978

ATTACH1\.1ENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.3.6, REFUELING CHR INSTRUMENTATION A.3 CTS 3.8.lc requires two radiation monitors to be tested and verified to be operable immediately prior to refueling operations. In addition, Table 3 .17 .6 requires two Containment Refueling Radiation Monitors to be OPERABLE during REFUELING OPERATIO NS when irradiated fuel is in the containment.
In the ITS, the requirement that two radiation monitors be OPERABLE is placed in ITS LCO 3.3.6. The details of the CTS Applicability and testing requirements have been relocated to the Applicability and Surveillance Requirements of ITS 3.3.6. This change is simply a change in format placing the requirements of the CTS in the corresponding ITS LCO, and is therefore considered to be administrative. A.4 The Bases of the current Technical Specifications for this section have been completely replaced by revised Bases that reflect the format and applicable content consistent with NUREG-1432. The revised Bases are shown in the proposed Technical Specification Bases. A.5 CTS Table 3.17.6, Function 20, Footnote (a), contains an allowance that
Specifications 3.0.4 and 4.0.4 are not applicable. Footnote (b) contains an allowance that Specifications 3.0.3, 3.0.4, and 4.0.4 are not applicable. Since the applicability of Specifications 3.0.3, 3.0.4 and 4.0.4 is changed in the ITS to only apply in MODES 1, 2, 3 and 4, these footnotes have been deleted.
Since this change does not alter any technical requirements, it is considered to be administrative. MORE RESTRICTIVE CHANGES (M) M.1 CTS Table 4.17.6, Function 20, requires that a CHANNEL CHECK be performed on the containment refueling radiation monitor at a Frequency of 24 hours. ITS SR 3.3.6.1 requires that the CHANNEL CHECK be performed at a Frequency of 12 hours. This change is a reduction in surveillance Frequency and is considered more restrictive. The 12 hour Frequency is based on operating experience, and minimizes the chance of a loss of a protective function due to failure of redundant channels. This change is consistent with NUREG-143~.
Palisades Nuclear Plant Page 2 of 4 5d(- e__
01/20/98
ATTACHMENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.3.6, REFUELING CHR INSTRUMENTATION A.4 The Bases of the current Technical Specifications for this section have been completely replaced by revised Bases that reflect the format and applicable content consistent with NUREG-1432. The revised Bases are shown in the proposed Technical Specification Bases. A.5 CTS Table 3.17.6, Function 20, Footnote (a), contains an allowance that Specifications 3.0.4 and 4.0.4 are not applicable. Footnote (b) contains an allowance that Specifications 3.0.3, 3.0.4, and 4.0.4 are not applicable. Since the applicability of Specifications 3.0.3, 3.0.4 and 4.0.4 is changed in the ITS to only apply in MODES 1, 2, 3 and 4, these footnotes have been deleted. Since this change does not alter any technical requirements, it is considered to be administrative. MORE RESTRICTIVE CHANGES (M) M.1 CTS Table 4.17.6, Function 20, requires that a CHANNEL CHECK be performed on the containment refueling radiation monitor at a Frequency of 24 hours. ITS SR 3.3.6.1 requires that the CHANNEL CHECK be performed at a, Frequency of *
12 hours. This change is a reduction in surveillance Frequency and is considered more restrictive. The 12 hour Frequency is based on operating experience, and minimizes the chance of a loss of a protective function due to failure of redundant channels. This change is consistent with NUREG-1432.
() M.2 ITS LCO 3.3.6 contains a requirement for two CHR manual actuation function I channels to be operable in addition to the two refueling containment radiation monitor channels required by CTS LCOs 3.8 and 3.17. This requirement was added to assure that diverse actuation methods are available if open containment isolation penetrations need to be closed during Core Alterations or movement of irradiated fuel. The requirement for manual closure capability is consistent with NUREG 1432 .
~
Palisades Nuclear Plant 05/30/99

L.2 ATTACHMENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.3.6, REFUELING CHR INSTRUMENTATION

Palisades Nuclear Plant Page 4 of 4 5ci-i 01/20/98
ATTACHMENT 3
L.2 DISCUSSION OF CHANGES SPECIFICATION 3.3.6, REFUELING CHR INSTRUMENTATION CTS 3.8.2 requires that if any of the conditions of CTS 3.8.1 are not met, all refueling operations shall cease immediately. For ITS LCO 3.3.6, when one required channel of automatic (or ma.n.u~l) actuation is inoperable, that Required Action has been replaced with Actions which require placing the inoperable channel in trip (i.e., performing its safety function) or removing the plant from the applicable conditions of the LCO. The completion time has been extended from immediately to 4 hours. The extension of the completion time from "immediately" to 4 hours is offset by the additional requirement (see DOC M.2) to have two manual actuation channels Operable. The four hour completion time only applies when at least one channel of manual actuation and one channel of automatic actuation remain Operable.
With two channels inoperable, ITS 3.3.6 retains the CTS action of removing the plant from the applicable conditions of the LCO immediately. CTS 3.8.2 further requires that, "work shall be initiated to satisfy the required conditions and that no operations that may change the reactivity of the core shall be made." These requirements are not appropriate once the Applicability for the LCO has been exited. Therefore, they have been deleted .
The ITS retains the pertinent, and sufficient, requirements in LCO 3.3.6 to assure the OPERABILITY of the Refueling CHR instrumentation when it is needed to support the reduction of dose consequences following a fuel handling accident. Therefore, this deletion can be made with no impact to the health and safety of the public. This change is consistent with NUREG-1432 .

Palisades Nuclear Plant Page 4 of 4 05/30/99

LESS RESTRICTIVE CHANGE L.2 NO SIGNIFICANT HAZARDS CONSIDERATION SPECIFICATION 3.3.6, REFUELING CHR INSTRUMENTATION ATTACHMENT 4 S 3.8.2 requires that if any of the conditions of CTS 3.8.1 are not met, all refueling /
ope tions shall cease immediately. It further requires that, "work shall be initiated to satisfy the re ired conditions and that no operations that may change the reactivity of the core shall be made." These requirements are not appropriate once the Applicability for the LCO has been exite . Therefore, they have been deleted. The ITS retains the pertinent, and suffi ient, requirements 1 LCO 3.3.6 to assure the OPERABILITY of the Refueling CHR instrumentation en it is needed to support the reduction of dose consequences fo owing a fuel handling accid t. Therefore, this deletion can be made with no impact to e health and safety of the public. *s change is consistent with NUREG-1432.
1. Does the change in Ive a significant increase in the probab *
or consequence of an accident previous} valuated?
Analyzed everits are assume to be initiated by the failur of plant structures, systems or components. The proposed ange omits actions t not necessary to assure safety
once the Applicability for the LC has been exited Once the CORE ALTERATIONS and movement of irradiated fuel wi

the con

ent have ceased, there is no longer a need for the Refueling CHR Instrume tio o protect against a fuel handling accident. Therefore, additional Required tions are inappropriate. Restoration will still be required under the ITS before su a "vities can be restarted. Therefore, this relaxation will not alter the operation f any pla equipment, or otherwise increase its failure probability. The probabili that equipmen ailures will result in occurrence of an analyzed event is unrelated t a component which* "tiates a protective action. As such, the probability of occu nee for a previously ana ed accident is not significantly increased.
The consequences of previously analyzed event are dependent the initial conditions assumed for the ysis, and the availability and successful functi
g of the equipment ass ed to operate in response to the analyzed event, and e setpoints at tions are initiated. Performance of actions outside conditI under el handling accident can occur, i.e:, the Applicability, is not r uired to ensure at the assumptions of the safety analysis are met and does not affect e perfi ance of any credited equipment. As a result, no analyses assumptions v* ated. Based on this evaluation, there is no significant increase in the conseque f a previously analyzed event .

Palisades Nuclear Plant Page 3of4 01/20/98

ATTACHMENT 4 NO SIGNIFICANT HAZARDS CONSIDERATION SPECIFICATION 3.3.6, REFUELING CHR INSTRUMENTATION Does the change create the possibility of a new or different kind of accident fr y accident previously evaluated?
oposed change does not involve a physical alteration of the plant. o new equipm tis being introduced, and no installed equipment is being o rated in a new o different er. There is no alteration to the parameters within *ch the plant is normally ope ted or in the setpoints which initiate protective o mitigative actions. No change is being oposed to the procedures governing norm plant operation or those procedures relied u on to mitigate a design basis event. elaxing the action requirements while o ide the MODE of Applicabili Cloes not have a detrimental impact on the manner in. hich plant equipment o ates or responds to an actuation signal. As such, no new fa' ure MODEs are b
g introduced. In addition, the change does not alter assumptions ma in the safe analyses and licensing basis. Therefore, the change does not create the po ibilitY, f a new or different kind of accident from any accident previously evaluated.
Does this change involve a si 1cant r uction in a margin of safety?
The margin of safety is ermined by the des1 and qualification of the plant equipment, the operaf n of the plant within anal ed limits, and the point at which protective or miti 1ve actions are initiated. Relax1 the action requirements while outside the MO of Applicability does not significan impact these factors. There are no desi changes or equipment performance paramet changes associated with this chan . No setpoints are affected, and no change is be' proposed in the plant opera

nal limits as a result of this change. Therefore, this ch ge does not involve a ficant reduction in the margin of safety .

Palisades Nuclear Plant Page 4 of 4 sa-~
01/20/98
ATTAC1'ENT 4
NO SIGNIFICANT HAZARDS CONSIDERATION SPECIFICATION 3.3.6, REFUELING CHR INSTRUMENTATION
2. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?
The proposed change does not involve a physical alteration of the plant. No new equipment is being introduced, and no installed equipment is being operated in a new or different manner. There is no alteration to the parameters within which the plant is normally operated or in the setpoints which initiate protective or mitigative actions. No change is being proposed to the procedures governing normal plant operation or those
*procedures relied upon to mitigate a design basis event. Relaxing the requirement to confirm the OPERABILITY of the Refueling CHR channels prior to entering the MODE of Applicability does not have a detrimental impact on the manner in which plant equipment operates or responds to an actuation signal. As such, no new failure modes are being introduced. In addition, the change does not alter assumptions made in the safety analyses and licensing basis. Therefore, the change does not create the possibility of a new or different kind of accident from any accident previously evaluated.
3* Does this change involve a significant reduction in a margin of safety?
The margin of safety is determined by the design and qualification of the plant equipment, the operation of the plant within analyzed limits, and the point at which protective or mitigative actions are initiated. Relaxing the requirement to confirm OPERABILITY of the Refueling CHR channels does not significantly impact these factors. There are no design changes or equipment performance parameter changes associated with this change. No setpoints are affected, and no change is being proposed in the plant operational limits as a result of this change. Therefore, this change does not involve a significant reduction in the margin of safety.
LESS RESTRICTIVE CHANGE L.2 CTS 3.8.2 requires that if any of the conditions of CTS 3.8.1 are not met, all refueling I operations shall cease immediately. For ITS LCO 3.3.6, when one required channel of I automatic (or manual) actuation is inoperable, that Required Action has been replaced with IQ-Actions which require placing the inoperable channel in trip (i.e. , performing its safety I I function) or removing the plant from the applicable conditions of the LCO. The completion I~ time has been extended from immediately to 4 hours. The extension of the completion time I to" from "immediately" to 4 hours is offset by the additional requirement to have two manual I~ actuation channels Operable. The four hour COI\'lpletion time only applies when at least one 1-
. channel of manual actuation and one channel of automatic actuation remain Operable. 1 !i:t...
I~ I Palisades Nuclear Plant Page 2 of 4 05/30/99
ATTACHMENT 4
NO SIGNIFICANT HAZARDS CONSIDERATION SPECIFICATION 3.3.6, REFUELING CHR INSTRUMENTATION LESS RESTRICTIVE CHANGE L.2 (continued)
With two channels inoperable, ITS 3.3.6 retains the CTS action of removing the plant from the applicable conditions of the LCO immediately. CTS 3.8.2 further requires that, "work shall be initiated to satisfy the required conditions and that no operations that may change the reactivity of the core shall be made." These requirements are not appropriate once the Applicability for the LCO has been exited. Therefore, they have been deleted. The ITS retains the pertinent, and sufficient, requirements in LCO 3. 3. 6 to assure the OPERABILITY of the Refueling CHR instrumentation when it is needed to support the reduction of dose consequences following a fuel handling accident. Therefore, this deletion can be made with no impact to the health and safety of the public. This change is consistent with NUREG-1432.
1. Does the change involve a significant increase in the probability or consequence of an accident previously evaluated? --
G I Analyzed events are assumed to be initiated by the failure of plant structures, systems .
\...9 or components. The proposed change extends the allowed action time when one required automatic actuation channel is inoperable, and omits actions that are not ~
necessary to assure safety once the Applicability for the LCO has been exited. The four hour completion time extensio~ only applies when at least one channel of manual actuation and one channel of automatic actuation remain Operable, therefore, there is no potential for loss of isolation capability. Once the CORE ALTERATIONS and movement of irradiated fuel within the containµlent have ceased, there is no longer a need for the Refueling CHR Instrumentation to protect against a fuel handling accident. Therefore, additional Required Actions are inappropriate. Therefore, this relaxation will not alter the operation of any plant equipment, or otherwise increase its failure probability. The probability that equipment failure*s will result in occurrence of an analyzed event is unrelated to a component which initiates a protective action. As such, the probability of occurrence for a previously analyzed accident is not significantly increased .
Palisades Nuclear Plant Page 3 of 4 05/30/99
ATTACHMENT 4
NO SIGNIFICANT HAZARDS CONSIDERATION SPECIFICATION 3.3.6, REFUELING CHR INSTRUMENTATION LESS RESTRICTIVE CHANGE L.2
1. (continued)
The consequences of a previously analyzed event are dependent on the initial conditions assumed for the analysis, and the availability and successful functioning of the equipment assumed to operate in response to the analyzed event, and the setpoints at which these actions are initiated. The extended allowed outage time does not affect the consequence of any previously analyzed event, because the ability for both automatic and manual closure of containment penetrations is required for this extension to be used. Performance of actions outside conditions under which the fuel handling accident can occur, i.e., the Applicability, is not required to ensure that the assumptions of the safety analysis are met and does not affect the performance of any credited equipment. As a result, no analyses assumptions are violated. Based on this evaluation, there is no significant increase in the consequences of a previously analyzed event.
2. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?
The proposed change does not involve a physical alteration of the plant. No new
equipment is being introduced, and no installed equipment is being operated in a new or different manner. There is no alteration to the parameters within which the plant is normally operated or in the setpoints which initiate protective or mitigative actions. No change is being proposed to the procedures governing normal plant operation or those procedures relied upon to mitigate a design basis event. Extending the completion time for degraded isolation capability or relaxing the action requirements while outside the MODE of Applicability does not have a detrimental impact on the manner in which plant equipment operates or responds to an actuation signal. As such, no new failure MODES are being introduced. In addition, the change does not alter assumptions made in the safety analyses and licensing basis. Therefore, the change does not create the

possibility of a new or different kind of accident from any accident previously evaluated.
3. Does this change involve a significant reduction in a margin of safety?
The margin of safety is determined by the design and qualification of the plant equipment, the operation of the plant within analyzed limits, and the point at which protective or mitigative actions are initiated. Extending the completion time for degraded isolation capability or relaxing the action requirements while outside the MODE of Applicability does not significantly impact these factors. There are no design changes or equipment performance parameter changes associated with this change. No setpoints are affected, and no change is being proposed in the plant operational limits as a result of this change. Therefore, this change does not involve a significant reduction in the margin of safety. Palisades Nuclear Plant Page 4 of 4 05/30/99
(Refueling CAR lnstrumentotlon~
~~
ill. - @ 0-i' nstrumentotion Two Refueling CHR Automotic Actuotion Function chonnels ond t1to CHR Monuol Actuotion Function chonnels APPLI CAB IL ITV: During CORE ALTERATIONS, During 1M>vt11ent of irradiated fuel assellblies within containMnt. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME
One or more Functions with one chonnel A. l QB A.2.1 Place th* affected channel in trip.
Suspend CORE 4 hours RAI 3.3.6-01 moperoble. ALTERATIONS. A.2.2 Suspend 11avement of irradhted fuel assemblies within containment. One or more Qmmediotely) Functions with two chonnels inoperoble. (continued)
( Pohsodes Nucleor Plont) ldoG ifs I 5J-r\__
3.3-30
(Refuehnq CAR lnstrumentot1on~~
ACTIONS ill. CONDITION REQUIRED ACTION COMPLETION TIME
SURVEILLAHCE REQUIREMENTS SURVE ILLAHCE 1
FREQUENCY
~:tlb :1~
(T4.17.6<20l] SR 3.3~ Perfon1 a CHANNEL CHECK on Heh lZ hours
~ 110n1tor channel. @ID RAI 3.3.6-01 (T 4.17.6<20)]
SR 3.3~ ~days I@ (continued)
(Pohsodes Nucleor Plont)
[C(oG ifs I 5'~- 0 3.3-31 [Afv 1/ 04//B?/9~
© SURVEILLANCE RE UIREMENTS continued CTS SURVEILLANCE FREQUENCY
[T 4.17 .6<2~~ filtiRi
@On tafi11111! ragzr Perform a CHANNEL CALIBRATION on each (refue ingR 9111 llOn i tor channel .
i]IS(t1 months RAI 3.3.6-01
~
SR~ i]IB~ months
(Polisodes Nucleor Plont) l<<oG ifs I 5~-
3.3-32 . ? 1.Mv 1/ 04//Q1/9§J
Change SPECIFICATION 3.3.7, CONTAINMENT PURGE ISOLATION SIGNAL (CPIS)
Discussion ATTACHMENT 6 JUSTIFICATION FOR DEVIATIONS
8. Palisades includes two Refueling Containment High Radiation (CHR) instrument channels which are required to be OPERABLE during CORE ALTERATIONS, and during movement of irradiated fuel assemblies within containment. The purpose of these channels are to close the containment penetrations which have direct access from the containment atmosphere to the outside atmosphere and to initiate the Control Room Ventilation System in the emergency mode of operation. Since these channels perform a dual function which provide the protection as required by ISTS 3.3.7 and ISTS 3.3.8, the requirements have been combined and included into ITS 3.3.6, "Refueling Containment High Radiation (CHR) Instrumentation. ,
See ~,__,; i "::)e...~ J FD
9. here are two Refueling CHR c anne s m w c eac 1s connecte o 1 s ass
ted logic circuit. Each logic circuit will actuate when the associ instrume annel trips (one-out-of-one). Therefore in this Specifi ion a channel is co

red to include the actuation logic circuit as . Thus the term "automatic Ac *on Logic" has been deleted throu out the Specification. A Manual

channel is provided o ssociated with each Refuel_ing CHR channel, but al Trip cha s have not been included in the ITS, since the safety analysis doe t e credit for Manual Trip. This is acceptable since if any Refueling CH a 1 is inoperable the proposed actions are to immediately suspe the refueling "vities as indicated in ITS 3.3.6 ACTION A. AM Trip channel is availab allow operator action, but it is not required, therefore, not necessary to be inc Thus the require ts of automatic Actuation Logic and Manual are deleted fro e ISTS 3.3.7 LCO, ACTIONS and surveillance. TST - 4 and TSTF- are also not incorporated on the same basis.
10. The verification of the Allowable Value has been removed from the CHANNEL CALIBRATION surveillance since for consistency with* the current licensing basis.
This radiation monitor setpoint is not a specific assumption in the safety analysis and as such, an Allowable Value cannot be developed. The CHANNEL CALIBRATION will continue to check the setpoint, but changes to the setpoint will not require a License amendment. TSTF-186 is also not incorporated on the same basis .
Palisades Nuclear Plant Page 2of3 5d-tr-
. 01/20/98
Change SPECIFICATION 3.3.7, CONTAINMENT PURGE ISOLATION SIGNAL (CPIS)
Discussion ATTACHMENT 6 JUSTIFICATION FOR DEVIATIONS
11. Details related to location, associated logic, system ties, and other operational considerations associated with the containment radiation monitors is added, consistent with details relocated from CTS 3.8.l (see DOC LA.1).

12. Details related to ancillary requirements associated with inoperable containment radiation monitors is added, consistent with details relocated from CTS 3.8.2 (see DOC LA.2).

13. See r\ -e Lu ~\ F D '~,
Palisades Nuclear Plant Page 3of3 odJ-)'l_
01/20/98
ATTACHMENT 6
Change 9.
JUSTIFICATION SPECIFICATION 3.3.7, CONTAINMENT PURGE ISOLATION SIGNAL (CPIS) Discussion FOR ISTS LCO 3.3. 7 has been changed to reflect the Palisades equipment (ITS I DE VIA TIO NS LCO 3.3.6). Palisades utilizes two refueling radiation monitors, each of which directly actuates one train of the isolation functions actuated by the Containment High Radiation (CHR) signal used during plant operation, where the ISTS reflects use of one train four monitor in a two-out-of-four logic. ITS LCO 3.3.6 also requires two CHR manual actuation channels to be operable, where the ISTS requires only one. ITS Condition A provides four hours for restoration when the LCO is not met, but at least one channel of automatic actuation and one channel of manual actuation are available. If equipment inoperabilities reduce the actuation capability beyond that level, Condition B requires immediate action to place the plant in a condition outside the applicability of the LCO. There are two Refueling CHR channels in which each is directly connected (one-out-of-one logic) to actuate its associated CHR output relays. Thus the term "automatic Actuation Logic" has been deleted throughout the Specification.
10. The verification of the Allowable Value has been removed from the CHANNEL
CALIBRATION surveillance since for consistency with the current licensing basis .
This radiation monitor setpoint is not a specific assumption in the safety analysis and as such, an Allowable Value cannot be developed. The CHANNEL CALIBRATION will continue to check the setpoint, but changes to the setpoint will not require a License amendment. TSTF-186 is also not incorporated on the same basis.
11. Details related to location, associated logic, .system ties, and other operational considerations associated with the containment radiation monitors is added, consistent with details relocated from CTS 3.8.1 (see DOC LA. I).

12. Details related to ancillary requirements associated with inoperable containment radiation monitors is added, consistent with details relocated from CTS 3.8.2 (see DOC LA.2).

13. ISTS SR 3.3.7.6, RESPONSE TIME test is not included in the ITS. This test is not required by the current licensing basis since the conclusions of NUREG-082, "Integrated Plant Safety Assessment Systematic Program Palisades Plant" - Final Report USNRC 10/82, concluded that backfitting this testing into the current program was not required .
Palisades Nuclear Plant Page 2 of 2 05/30/99
CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO JANUARY 6, 1999 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.3, INSTRUMENTATION
NRC REQUEST:
3.3.6-02 JFD 9 STS (3.3. 7) Condition B addresses manual trip or actuation logic channels inoperable and requires isolation of the valves iendered inoperable by this instrumentation with subsequent entry into the affected LCOs. Comment: JFD 9 is annotated in the mark-up for deleting Condition B but does not specifically address eliminating the actions of Condition B. The deviation from the STS seems to be preference because the Palisades design would allow for STS required actions to apply; isolation of the specific valve(s) rendered inoperable by the instrumentation or require the safety function of the system to be initiated, thus preserving the plant analysis. Provide discussion and justification for these STS deviations.
**Discuss with Comment 3.3.6-01.
Consumers Energy Response: Following discussions with the NRC reviewer during the 10/27/98 meeting, Palisades has revised ITS LCO 3.3.6, to better reflect the installed CHR equipment which is used during Core Alterations and fuel handling in the containment. Since that equipment differs from that presumed in STS LCOs 3.3.7, the proposed LCO is patterned after conditions and actions in STS for other owner's groups. The equipment itself, and the intent of the proposed actions are described in the Bases provided. See also the response to RAI 3.3.6-01. Affected Submittal Pages: See the response to RAI 3.3.6-01 .
53
CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO JANUARY 6, 1999 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.3, INSTRUMENTATION NRC REQUEST: 3.3.6-03 JFD 8 JFD 8 states that the radiation channels have dual purposes; to isolate containment and to place the CRVS system in the emergency mode of operation. Comment: The proposed ITS delete STS LCO 3.3.8, Control Room Isolation Signal. When one or more of these channels become inoperable some or all capability to pressurize the CR is lost. Discuss the safety implications of not including an ITS LCO that provides remedial actions to place the control room in the emergency mode for one or more inoperable radiation isolation channels during Modes 1-4 and during refueling or core alteration operations. Justify these STS deviations.
** Discuss with Comment 3.3.6-01.
Consumers Energy Response: The proposed ITS omit STS LCO 3.3.8 because Palisades does not have any equivalent control signal in the Palisades design or. in CTS, as explained in JFD 6 for STS LCO 3.3.8. The automatic actuation of the control room ventilation system (i.e., place it in the emergency mode) is provided by the CHP, CHR, and Refueling CHR signals. No remedial actions to place the control room ventilation signal in the emergency mode are required in MODES 1 through 4 because the CHP and CHR signals are required to be Operable with no allowance for degradation to the point of loss of function. Similarly, during Core Alterations and during movement of irradiated fuel in the containment, Refueling CHR is required to be Operable with no allowance for degradation to the point of loss of function. Affected Submittal Pages: No page changes .
54
CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO JANUARY 6, 1999 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.3, INSTRUMENTATION NRC REQUEST: 3.3.6-04 Resolved during 10/27/98 meeting; Consumers Energy not required to respond. NRC REQUEST: 3.3.7-0i Resolved during 10/27/98 meeting; Consumers Energy not required to respond. NRC REQUEST: 3.3.7-02 Resolved during 10/27/98 meeting; Consumers Energy not required to respond. NRC REQUEST: 3.3.7-03 Action 8.1, CTS 3.17.1-3.17.3, Functions 1-15 CTS 3.17.4 (CETs) ITS Action 8.1 is to initiate action in accordance With Specification 5.6.6 if one channel of inoperable equipment cannot be restored to operable status. CTS 3.17.3 for PIVs has an alternate allowance which requires isolating and locking closed valves with inoperable position indication. If CTS repair actions (or in the case of CIVs, isolation) are not met, then CTS 3.17.4 for Functions 1-15 and CTS 3.17.4.7.a for CETs requires the reactor to be in Mode 3 within 6 hours and Mode 4 within 30 hours if inoperable functions are not restored to operable status. Comment: L.1 does not evaluate all CTS 3.17.4.1 and 3.17.4.a changes that result from
adopting the less restrictive ITS Action 8.1. Revise the submittal to provide justifications for each less restrictive CTS change that results from translating CTS into the STS format.
Consumers Energy Response: The CTS LCO 3.17.4 markup and DOCs L.1, L.2, L.3, and L.4 have been revised to more clearly identify and explain the proposed changes. Affected Submittal Pages: Att 3, CTS 3.3.7, page 3-70 Att 3, DOC 3.3.7, page 3 of 5 (former submittal) Att 3, DOC 3.3.7, page 4 of 5 Att 3, DOC 3.3.7, page 3 of 5 Att 3, DOC 3.3.7, page 4 of 5 Att 4, NSHC 3.3.7, page 1 of 11 (former submittal) Att 4, NSHC 3.3.7, page 1 of 11 Att 4, NSHC 3.3.7, page 2of11 Att 4, NSHC 3.3. 7, page 3 of 11 Att 4, NSHC 3.3.7, page 4of11 Att 4, NSHC 3.3.7, page 5of11
55
3,l,7
@\
-.:rr.s 3 .,11
[3.3.7] l. 17 '4 (f~o A] 3.17.4.l (cclJt> ~J 3.17.4.Z (c.01Jt> AJ 3.17.4.3 [~ 1. 3.11-1 1-1+-&. ~ (c.oNb B) 3.17.4.4
[cot.iC> F-J
[_coNt> A] 3.17.4.5 (;out;) l] 3.17.4.6 3.17.4.7 [corJ.b r] ** complete Act1on
b. IMlltl lllOnlt fcotJD B] c:. Submit a report to th* HRC in 1ccord1nc1 w1th Sptc1f1catton 6.6.7.
(lofJO c;.J
d. LE s
.-.endmlnt No. ~. ~. ~. 174 October 31, 1996 P~e. \ o~ 3
LESS RESTRICTIVE CHANGES (L)
ATTACHMENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.3.7, PAM INSTRUMENTATION L.1 e inoperable containment isolation valve (Function 15) to be restored to Operab sta s within 7 days, or place the reactor in Hot Shutdown within the following 12 ho s, and place the reactor in a condition where the affected equipment i not require , within the following 48 hours. ITS 3.3.7 Condition A is applica e to all accident m itoring functions, and increases the Completion Time from days to 30 days. IT .3.7 Condition B requires action be initiated imrnediat yin accordance with Specificati 5 .6.6 if the inoperable channel is not restored to perable status-~ within 30 days. e 30 day Completion Time is based on opera
g experience and takes into account th remaining Operable channel, the passive ature of the instrument (no critical automatic a ion is assumed to occur from these

struments), and the low probability of an event re iring accident monitoring inst mentation during this interval. The requirement t initiate action in accorda e with Specification 5.6.6 is appropriate in lieu of a shutdo requirement, given e likelihood of plant conditions that would require information p vided by this in rumentation. Also, alternative*
Required Actions are identified be~ e a loss of nctional capability condition occurs. This change is consistent with NURE -1432 L.2 CTS 3 .17.4.2 requires ori~ containment ogen concentration channel to be restored to operable status within 48 hours wh two c ntainment hydrogen concentration channels are inoperable. ITS 3 .3. 7 ct ion D in onjunction with Condition C Note requires one hydrogen monitor c nnel to be resto d to Operable status within 72 hours when two hydrogen nitor channels are in erable. The 72 hour Completion Time is based o the relatively low probab1 'ty of an event requiring hydrogen monitoring and e availability of alternative me s to obtain the required information. This cha e is consistent with NUREG-1432. L.3 CTS 3 .17 .4.2 and .17 .4.6 require one channel for functions 1 tli ugh 14 and functions 16 thr gh 21 respectively to be restored to Operable sta within 48 hours when two re red channels are inoperable. ITS 3.3.7 Action C requi s one channel 1 for all acci nt monitoring functions except for hydrogen monitor channe to be restored Operable status within 7 days with one or more Functions with o required chann inoperable. The Completion Time of 7 days is based on the relative low pro ility of an event requiring accident monitoring instrumentation operation d the a ilability of alternate means to obtain the required information. This change is ((1H 3.). ') -0'3
Palisades Nuclear Plant Page 3 of 5 55-b 01/20/98

L.4 ATTACHMENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.3.7, PAM INSTRUMENTATION S'eG.. R~ v, ca ra."?:. l. "I 'T.>1 ~c:.u~ <>> to'-(
L.5 CTS 3.17.4.7b requirement that alternate monitoring be initiated within 48 hours with two Reactor Vessel Water Level channels inoperable is deleted. ITS 5.6.6 requires the Accident Monitoring Report to outline the preplanned alternate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels of the Function to Operable status. These controls assure that the appropriate margin of safety is maintained. This change is consistent with
NUREG-1432.
L.6 CTS 3.17.4.7d requirement that the inoperable channels for functions 16 through 21 be restored to Operable status prior to startup from the next refueling is deleted. However, the NRC will be informed of the schedule for restoring the system, as required by ITS 5.6.6. This requirement will ensure that the inoperable channel(s) are restored to Operable status in a timely manner. Also. good operating practice and management oversight dictate that plant systems be restored as soon as possible. This change is consistent with NUREG-1432. L. 7 CTS 3 .17.4. 3 provides action requirements for inoperable position indication of containment isolation valves which allows the associated valves to be locked in the closed position. ISTS Table 3. 3. 11-1 Note (a) also provides this allowance but also allows other options such as isolation of the associated penetration by any a_utomatic valve which is closed and de-activated, a closed manual valve, a blind flange, or a check valve with flow through the valve secured. Since all of these options provide acceptable isolation of the penetration, ITS 3. 3. 7 will adopt the additional allowances as provided in the ISTS. This change is consistent with NUREG-1432 .
Palisades Nuclear Plant Page 4 of 5 55-e.-
01/20/98
ATTACHMENT 3
LESS RESTRICTIVE CHANGES (L)
L.1 DISCUSSION OF CHANGES SPECIFICATION 3.3. 7, PAM INSTRUMENTATION The ITS Actions contain completion times which are longer than their CTS counterparts; ITS Action A.1 allows 30 days for restoration of a single inoperable channel for one or more Functions; CTS 3.3.4 Actions la, 3a, and Sa allow 7 days for this restoration. The longer ITS completion times are taken from the STS. They are based on the additional operating experience available when the STS were written and take into account the passive nature of the instruments (no automatic action is assumed to occur from these instruments), and the low probability of an event requiring accident monitoring instrumentation during this interval. The 30 day Completion Time for restoration of a single inoperable channel (for one or more Functions) also considers the remaining Operable channel. This change is consistent with NUREG 1432. L.2 The ITS Actions contain completion times which are longer than their CTS counterparts; ITS Action D. l allows 3 days for restoration for one of two inoperable channels of Hydrogen Monitoring; CTS Action 2a allows 2 days for this restoration. The longer ITS completion times are taken from the STS. They are based on the additional operating experience available when the STS were written and take into account the passive nature of the instruments (no automatic action is assumed to occur
L.3 from these instruments), and the low probability of an event requiring accident monitoring instrumentation during this interval. These changes are consistent with NUREG 1432.
The ITS Actions contain completion times which are longer than their CTS counterparts; ITS Action C. l ailows 7 days for restoration for one of two inoperable channels (for Functions other than Hydrogen Monitoring); CTS Actions 2a and 6a allow 2 days for this restoration. The longer ITS completion times are taken from the STS. They are based on the additional operating experience available when the STS were written and take into account the passive nature of the instruments (no automatic action is assumed to occur from these instruments), and the low probability of an event requiring accident monitoring instrumentation during this interval. This change is consistent with NUREG 1432 .
Palisades Nuclear Plant
-d Page 3 of 5 05/30/99
ATTACHMENT 3
L.4 DISCUSSION OF CHANGES SPECIFICATION 3.3.7, PAM INSTRUMENTATION The ITS Action B.1 allows continued operation, provided a report is filed with the NRC, if a single inoperable channel for one or more functions cannot be restored to operable status within the specified completion time of Action A.1; CTS Actions 4a and 4b require a plant shutdown in this event (for Functions other than Reactor Vessel Water Level and Containment Area Radiation Monitoring).
The requirement to file a report with the NRC (initiate action in accordance with Specification 5.6.6), is appropriate given the likelihood of plant conditions that would require information provided by this instrumentation. Also, alternative Required Actions are identified before a loss of functional capability condition occurs. This change is consistent with NUREG-1432. L.5 CTS 3.17.4.7b requirement that alternate monitoring be initiated within 48 hours with two Reactor Vessel Water Level channels inoperable is deleted. The instrumentation which is used as an alternative to RVWL channels is also required by CTS 3.17.4 and ITS 3.3.7. No new alternative measurement methods are initiated to implement the requirements of CTS 3.17.4. 7b, rather, the Emergency Operating Procedures list alternative indications to be used in lieu of operable RVWL channels, which provide similar, although not identical, information. The PCS temperature channels and
Pressurizer Level instruments provide information to supplement any inoperable RVWL channels. ITS 5.6.6 requires the Accident Monitoring Report to outline the preplanned alternate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels of the Function to Operable status. These controls assure that the appropriate margin of safety is maintained. This change is consistent with NUREG-1432.
L.6 CTS 3.17.4.7d requirement that the inoperable channels for functions 16 through 21 be restored to Operable status prior to startup from the next refueling is deleted. However, the NRC will be informed of the schedule for restoring the system, as required by ITS 5.6.6. This requirement will ensure that the inoperable channel(s) are restored to Operable status in a timely manner. Also, good operating practice and management oversight dictate that plant systems be restored as soon as possible. This change is consistent with NUREG-1432. L.7 CTS 3.17.4.3 provides action requirements for inoperable position indication of containment isolation valves which allows the associated valves to be locked in the closed position. ISTS Table 3.3.11-1 Note (a) also provides this allowance but also allows other options such as isolation of the associated penetration by any automatic valve which is closed and de-activated, a closed manual valve, a blind flange, or a check valve with flow through the valve secured. Since all of these options provide
acceptable isolation of the penetration, ITS 3.3. 7 will adopt the additional allowances as provided in the ISTS. This change is consistent with NUREG-1432.
Palisades Nuclear Plant 5S-Page 4 of 5 05/30/99
LESS RESTRICTIVE CHANGE L.1 ATTACHMENT 4 NO SIGNIFICANT HAZARDS CONSIDERATION SPECIFICATION 3.3.7, PA.i\1 INSTRm-lENTATION i erable containment isolation valve (Function 15) to be restored to Operable status withi 7 da , or place the reactor in Hot Shutdown within the following 12 hours, and place t reactor a condition where the affected equipment is not required, within the follow* g 48 hours. S 3.3.7 Condition A is applicable to all accident monitoring function , and increases the mpletion Time from 7 days to 30 days. ITS 3.3.7 Condition B equires action be initiated imme *ately in accordance with Specification 5.6.6 if the inoper le channel is not~
restored to Operable tarus within 30 days. The 30 day Completion Tim 1s based on operating experience an takes into account the remaining Operable c nnel, the passive nature of the instrument (n critical automatic action is assumed to ccur from these instruments), and the low pro bility of an event requiring acci nt monitoring instrumentation during this inter . The requirement to init" te action in accordance with Specification 5.6.6 is appropriate in *eu of a shutdown re irement, given the likelihood of plant conditions that would require info ation provid by this instrumentation. Also, alternative Required Actions are identifie efore a ss of functional capability condition occurs. This change is consistent with NU ~32 .
1. Does the change involve a signifi an accident previously evalua a?
The proposed change reg ces the shutdown track stat ent with a requirement to send a report to the NRC. e report will be issued when one equired channel of Post Accident Monitori (PAM) instrument is inoperable and th ompletion Time cannot be met. The P instrumentation is not an initiator of any ana zed event. Post-accident mo
oring is still available from *the one remaining OPE BLE channel and the pre-p ed alternate method of monitoring. Also, the PAM inst entation does not pr ide an active safety function. The proposed change does not sig

1cantly affi initiators or mitigation of analyzed events, and therefore does not invo e a
'gnificant increase in the probability or consequences of an accident previous!
evaluated.
s. '3.1 -o :s'
Palisades Nuclear Plant Page 1of11
~5-1J 01/20/98
ATTACHMENT4
LESS RESTRICTIVE CHANGE L.1 NO SIGNIFICANT HAZARDS CONSIDERATION SPECIFICATION 3.3.7, PAM INSTRUMENTATION The ITS Actions contain compietion times which are longer than their CTS counterparts; ITS Action A.1 allows 30 days for restoration of a single inoperable channel for one or more Functions; CTS 3.3.4 Actions la, 3a, and Sa allow 7 days for this restoration. The longer ITS completion times are taken from the STS. They are based on the additional operating experience available when the STS were written and take into account the passive nature of the instruments (no automatic action is assumed to occur from these instruments), and the low probability of an event requiring accident monitoring instrumentation during this interval. The 30 day Completion Time for restoration of a single inoperable channel (for one or more Functions) also considers the remaining Operable channel. This change is consistent with NUREG 1432.
1. Does the change involve a significant increase in the probability or consequences of an accident previously evaluated?
The proposed change revises the time available to restore an inoperable PAM instrumentation channel from 7 days to 30 days before Action is required. PAM instruments are not initiators of any analyzed event. The other Operable PAM
instrumentation channel is available for indication, and the PAM instruments are only required to provide indication, no active function. The proposed change does not significantly affect initiators or mitigation of analyzed events, and therefore does not involve a significant increase in the probability or consequences of an accident previously evaluated.
2. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?
The proposed change revises the time available to restore an inoperable PAM instrumentation channel from 7 days to 30 days before Action is required. This change will not involve a significant change in the design or operation of the plant. No hardware is being added to the plant as part of the proposed change. The proposed change will not introduce any new accident initiators. Therefore, the change does not create the possibility of a new or different kind of accident from any accident previously evaluated.
3. Does the change involve a significant reduction in a margin of safety?
The proposed change revises the time available to restore an inoperable PAM instrumentation channel from 7 days to 3'0 days before Action is required. The other OPERABLE PAM instrument channel is available for indication, and the PAM instruments are only required to *provide indication, no active function. Therefore, the change does not involve a significant reduction in the margin of safety. Palisades Nuclear Plant Page 1of9 05/30/99 55-ccr-
ATTACHMENT 4
LESS RESTRICTIVE CHANGE L.2 NO SIGNIFICANT HAZARDS CONSIDERATION SPECIFICATION 3.3.7, PAM INSTRUMENTATION The ITS Actions contain completion times which are longer than their CTS counterparts; ITS Action D. l allows 3 days for restoration for one of two inoperable channels of Hydrogen Monitoring; CTS Action 2a allows 2 days for this restoration. The longer ITS completion times are taken from the STS. They are based on the additional operating experience avajlable when the STS were written and take into account the passive nature of the instruments (no automatic action is assumed to occur from these instruments), and the low probability of an event requiring accident monitoring instrumentation during this interval. These changes are consistent with NUREG 1432.
1. Does the change involve a significant increase in the probability or consequences of an accident previously evaluated?
The proposed change revises the allowed outage time from 48 hours to 72 hours when 2 channels of hydrogen monitor instrumentation are inoperable. The hydrogen monitor instrumentation is not an initiator of any analyzed event. The hydrogen monitor instrumentation does not provide any safety function in the mitigation of analyzed events. This instrument provides information to the operators, who are functioning as a backup to the automatic systems designed to mitigate accidents previously evaluated . The proposed change does not significantly affect initiators or mitigation of analyzed events, and therefore does not involve a significant increase in the probability or 2. consequences of an accident previously evaluated. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated? The proposed change revises the allowed outage time from 48 hours to 72 hours when 2 channels of hydrogen monitor instrumentation are inoperable. The change will not involve a significant change in design or operation of the plant. No hardware is being added to the plant as part of the proposed change. The proposed change will not introduce any new accident initiators. Therefore, the change does not create the possibility of a new or different kind of accident from any accident previously evaluated.
3. Does the change involve a significant reduction in a margin of safety?
The proposed change revises the allowed outage time from 48 hours to 72 hours when 2 channels of hydrogen monitor instrumentation are inoperable. The hydrogen monitor instrumentation does not provide any safety function in the mitigation of analyzed events. Also, the likelihood is remote that an event would occur which would require
\
the hydrogen monitor instrumentation. Therefore, the proposed change does not
involve a significant reduction in a margin of safety .
Palisades Nuclear Plant Page 2 of 9 05/30/99
ATTACHMENT 4
LESS RESTRICTIVE CHANGE L.3 NO SIGNIFICANT HAZARDS CONSIDERATION SPECIFICATION 3.3.7, PAM INSTRUMENTATION The ITS Actions contain completion times which are ionger than their CTS counterparts; ITS Action C. l allows 7 days for restoration for one of two inoperable channels (for Functions other than Hydrogen Monitoring); CTS Actions 2a and 6a allow 2 days for this restoration.
The longer ITS completion times are taken from the STS. They are based on the additional operating experience available when the STS were written and take into account the passive nature of the instruments (no automatic action is assumed to occur from these instruments), and the low probability of an event requiring accident monitoring instrumentation during this interval. This change*is consistent with NUREG 1432.
1. Does the change involve a significant increase in the probability or consequences of an accident previously evaluated?
The proposed change revises the allowed outage time from 48 hours to 7 days when 2 channels of PAM instrumentation are inoperable except for hydrogen monitor instrumentation. The PAM instrumentation is not an initiator of any analyzed event. The PAM instrumentation does not provide any safety function in the mitigation of analyzed events. These instruments provides information to the operators, who are functioning as a backup to the automatic systems designed to mitigate accidents previously evaluated. The proposed change does not significantly affect initiators or mitigation of analyzed events, and therefore does not involve a significant increase in the probability or consequences of an accident previously evaluated.
2. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?
The proposed change revises the allowed outage time from 48 hours to 7 days when 2 channels of PAM instrumentation are inoperable except for hydrogen monitor instrumentation. The change will not involve a significant change in design or operation of the plant. No hardware is being added to the plant as part of the proposed change. The proposed change will not.introduce any new accident initiators. Therefore, the change does not create the possibility of a new or different kind of accident from any accident previously evaluated.
3. Does the change involve a significant reduction in a margin of safety?
The proposed change revises the allowed outage time from 48 hours to 7 days when 2 channels of PAM instrumentation are inoperable except for hydrogen monitor instrumentation. The PAM instrumentation does not provide any safety function in the
mitigation of analyzed events. Also, the likelihood is remote that an event would occur which would require the PAM instrumentation. Therefore, the proposed change does not involve a significant reduction in a margin of safety.
Palisades Nuclear Plant Page 3 of 9 05/30/99 55_;_
ATTACHMENT 4
LESS RESTRICTIVE CHANGE L.4 NO SIGNIFICANT HAZARDS CONSIDERATION SPECIFICATION 3.3.7, PAM INSTRUMENTATION The ITS Action B.1 aiiows continued operation, provided a report is filed with the NRC, if a single inoperable channel for one or more functions cannot be restored to operable status within the specified completion time of Action A. I; CTS Actions 4a and 4b require a plant shutdown in this event (for Functions other than Reactor Vessel Water Level and Containment Area Radiation Monitoring).
The requirement to file a report with the NRC (initiate action in accordance with Specification 5.6.6), is appropriate given the likelihood of plant conditions that would require information provided by this instrumentation. Also, alternative Required Actions are identified before a loss of functional capability condition occurs. This change is consistent with NUREG-1432.
1. Does the change involve a significant increase in the probability or consequences of an accident previously evaluated?
The proposed change replaces the shutdown track statement with a requirement to send a report to the NRC. The report will be issued when one required channel of Post Accident Monitoring (PAM) instrument is inoperable and the Completion Time cannot be met. The PAM instrumentation is not an initiator of any analyzed event. Post-accident monitoring is still available from the one remaining OPERABLE channel and the pre-planned alternate method of monitoring. Also, the PAM instrumentation does not provide an active safety function. The proposed change does not significantly affect initiators or mitigation of analyzed events, and therefore does not involve a significant increase in the probability or consequences of an accident previously evaluated.
2. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?
The proposed change replaces the shutdown track statement with a requirement to send a report to the NRC. The change will not involve a significant change in design or operation of the plant. No hardware is being added to the plant as part of the proposed change. The proposed change will not introduce any new accident initiators. Therefore, the change does not create the possibility of a new or different kind of accident from any accident previously evaluated .
Palisades Nuclear Plant 55-Page 4 o~
05/30/99
ATTACHMENT 4
3.
NO SIGNIFICANT HAZARDS CONSIDERATION SPECIFICATION 3.3.7, PAM INSTRUMENTATION Does the change involve a significant reduction in a margin of safety? The proposed change replaces the shutdown track statement with a requirement to send a report to the NRC. The report will be issued when one required channel of PAM instrument is inoperable and the Completion Time cannot be met. The remaining required channel or the pre-planned alternate method of monitoring is available to provide the required indication for PAM. The PAM instruments provides no automatic actuation functions. Therefore, the proposed change does not involve a significant reduction in a margin of safety. LESS RESTRICTIVE CHANGE L.5 CTS 3.17.4. 7b requirement that alternate monitoring be initiated within 48 hours with two Reactor Vessel Water Level channels inoperable is deleted. The instrumentation which is used as an alternative to RVWL channels is also requited by CTS 3.17.4 and ITS 3.3.7. No new alternative measurement methods are initiated to implement the requirements of CTS 3.17.4. 7b, rather, the Emergency Operating Procedures list alternative indications to be used in lieu of operable RVWL channels, which provide similar, although not identical, information. The PCS temperature channels and Pressurizer Level instruments provide
information to supplement any inoperable RVWL channels. ITS 5.6.6 requires the Accident Monitoring Report to outline the preplanned alternate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels of the Function to Operable status. These controls assure that the appropriate margin of safety is maintained. This change is consistent with NUREG-1432.
~
1. Does the change involve a significant increase in the probability or consequences of an accident previously evaluated?
The proposed change deletes the requirement that alternate monitoring be initiated within 48 hours with two Reactor Vessel Water Level channels inoperable. The Reactor Vessel Water Level instrumentation is not an initiator of any analyzed events. The requirement to report the Reactor Vessel Water Level instrumentation preplanned alternate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the instrument channels of the Function to OPERABLE status to the NRC will ensure that alternate monitoring is initiated in a timely manner. Good operating practice and management oversight dictate that alternate monitoring be initiated as soon as possible. The Reactor Vessel Water Level instrumentation does not provide any active safety function in the mitigation of analyzed events. The proposed change does not significantly affect initi3ttors or mitigation of analyzed events, and therefore does not involve a significant increase in the probability or consequences of
an accident previously evaluated_.
Palisades Nuclear Plant 55-k Page 5 of 9 05/30/99
CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO JANUARY 6, 1999 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.3, INSTRUMENTATION NRC REQUEST: 3.3.7-04 SR 3.3.7.2 The STS SR 3.3.7.2 Note that indicates that neutron detectors are excluded from Channel Caiibration has been deleted in the ITS Markup. Attachment 6, Discussion 6 states that this exclusion is being incorporated within the definition of Channel Calibration. Comment: The staff has not approved Channel Calibration definition changes. The submittal should be revised to conform to the STS. Consumers Energy Response: See response to "Definitions RAI" within this submittal. Affected Submittal Pages: See response to "Definitions RAI" within this submittal.
56
I CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS " RESPONSE TO JANUARY 6, 1999 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.3, INSTRUMENTATION
NRC REQUEST:
3.3.7-05 NRC REQUEST: 3.3.7-06 Resolved during 10/27/98 meeting; Consumers Energy not required to respond. DOC L.5 This DOC is used to justify eliminating a 48 hour requirement to initiate the alternate preplanned monitoring method for RVLS instruments inoperable and incapable of monitoring reactor level for post-accident management. Comment: Provide additional discussion for why the proposed change assures that the appropriate safety margin is maintained. Consumers Energy Response: DOC L.5 has been revised to provide additional explanation, as requested. Affected Submittal Pages: Att 3, DOC 3.3.7, page 4 of 5 (former submittal) Att 3, DOC 3.3.7, page 4 of 5 Att 4, NSHC 3.3.7, page 6of11 (former submittal) Att 4, NSHC 3.3.7, page 6of11
57
i

L.4 SPECIFICATION 3.3.7, PAM INSTRUMENTATION CTS 3.17.4.5 and 3.17.4.7.c require the one inoperable channel for functions 16 ATTACHMENT 3 DISCUSSION OF CHANGES through 21 to be restored to Operable status within 7 days, or submit a report to the NRC. ITS 3.3.7 Condition A is applicable to all accident monitoring functions, and increases the Completion Time from 7 days to 30 days. ITS 3.3.7 Condition B requires action be initiated immediately in accordance with Specification 5.6.6 if the inoperable channel is not restored to Operable status within 30 days. The 30 day Completion Time is based on operating experience and takes into account the remaining Operable channel, the passive nature of the instrument (no critical automatic action is assumed to occur from these instruments), and the low probability of an event requiring accident monitoring instrumentation during this interval. The requirement to initiate>
action in accordance with Specification 5 .6.6 is appropriate, given the likelihood of plant conditions that would require information provided by this instrumentation. Also, alternative Required Actions are identified before a loss of functional capability condition occurs. This change is consistent with NUREG-1432. L.5 CTS 3 .17. 4.70 reciHireffieRt that alterHate ffiSRitsriHg be iAitiatea witfiiR 48 ASHFS with t'NS Reaetsr Vessel '1\'ater bwel ef1aflflels iRsf:>erable is cleletea. ITS 5.6.6 reqHires the AssiasRt Monitoring R:efJOFt to 012tline tR@ fJr@plam:i@8 alt0n;;iat@ R;;i@tl:loe of R;;ionitoring, tl:i@ sa12se of tl:i@ inoperaeilir,*, ane tl:ii plans ans ssl:ieswl@ tor nistoring tR8 iRstraffl:eHtatisH ehanRels sf tae fHHetioR ts Of)eraele staras. These eoRtrols ass12re tRat tl:i@ appropriate r:J:targin of sat'et;* is R;;iaintained Tbii cbar:ige is CQRliiiter:it mitb.
}>l'URJ!G 1432. }c:e. Re.cJ1~e."T:> L..S" 1;.HS~\.t~c:,10~ ~"' 1.1.'l-o' L.6 CTS 3 .17.4. 7d requirement that the inoperable channels for functions 16 through 21 be restored to Operable status prior to startup from the next refueling is deleted.
However, the NRC will be informed of the schedule for restoring the system, as required by ITS 5.6.6. This requirement will ensure that the inoperable channel(s) are restored to Operable status in a timely manner. Also, good operating practice and management oversight dictate that plant systems be restored as soon as possible. This change is consistent with NUREG-1432. L. 7 CTS 3 .17 .4.3 provides action requirements for inoperable position indication of containment isolation valves which allows the associated valves to be locked in the closed position. ISTS Table 3.3.11-1 Note (a) also provides this allowance but also allows other options such as isolation of the associated penetration by any ~tomatic valve which is closed and de-activated, a closed manual valve, a blind flange, or a check valve with flow through the valve secured. Since all of these options provide acceptable isolation of the penetration, ITS 3 .3. 7 will adopt the additional allowances as provided in the ISTS. This change is consistent with NUREG-1432 .
Palisades Nuclear Plant Page 4 of S 57-()._/
01/20/98
ATTACHMENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.3.7, PAM INSTRUMENTATION L.4 The ITS Action B. l allows continued operation, provided a report is filed with the NRC, if a single inoperable channel for one or more functions cannot be restored to I"' () operable status within the specified completion time of Action A. l; CTS Actions 4a ! and 4b require a plant shutdown in this event (for Functions other than Reactor Vessel Water Level and Containment Area Radiation Monitoring). . IV\ (V\ The requirement to file a report with the NRC (initiate action in accordance with Specification 5.6.6), is appropriate given the likelihood of plant conditions that would require information provided by this instrumentation. Also, alternative Required Actions are identified before a loss of functional capability condition occurs. This change is consistent with NUREG-1432. L.5 CTS 3.17 .4. 7b requirement that alternate monitoring be initiated within 48 hours with two Reactor Vessel Water Level channels inoperable is deleted. The instrumentation which is used as an alternative to RVWL channels is also required by CTS 3.17.4 and ITS 3.3.7. No new alternative measurement methods are initiated to implement the requirements of CTS 3.17.4.7b, rather, the Emergency Operating Procedures list alternative indications to be used in lieu of operable RVWL channels, which provide similar, although not identical, information. The PCS temperature channels and
Pressurizer Level instruments provide information to supplement any inoperable RVWL channels. ITS 5.6.6 requires the Accident Monitoring Report to outline the preplanned alternate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels of the Function to Operable status. These controls assure that the appropriate margin of safety is maintained. This change is consistent with NUREG-1432.
L.6 CTS 3.17.4.7d requirement that the inoperable channels for functions 16 through 21 be restored to Operable status prior to startup from the next refueling is deleted. However, the NRC will be informed of the schedule for restoring the system, as required by ITS 5.6.6. This requirement will ensure that the inoperable channel(s) are restored to Operable status in a timely manner. Also, good operating practice and management oversight dictate that plant systems be restored as soon as possible. This change is consistent with NUREG-1432. L.7 CTS 3.17.4.3 provides action requirements for inoperable position indication of containment isolation valves which allows the associated valves to be locked in the closed p0sition. ISTS Table 3.3.11-1 Note (a) also provides this allowance but also allows other options such as isolation of the associated penetration by any automatic valve which is closed and de-activated, a closed manual valve, a blind flange, or a check valve with flow through the valve secured. Since all of these options provide acceptable isolation of the penetration, ITS 3.3. 7 will adopt the additional allowances as provided in the ISTS. This change is consistent with NUREG-1432. 51-Palisades Nuclear Plant Page 4 of 5 05/30/99
--- -- -- _ _ _ _ _ J
ATTACIThlENT 4 NO SIGNIFICANT HAZARDS CONS ID ERA TION
2.
SPECIFICATION 3.3.7, PMI INSTRmfENTATION Does the change create the possibility of a new or different kind of accident from any accident previously evaluated? The proposed change revises the time available to restore an inoperable PAM instrumentation channel from 7 days to 30 days before Action is required. Instead of a special report being submitted within 7 days, Action will be taken in accordance with Technical Specification 5 .6. 7 within 30 days. This change will not involve a significant change in the design or operation of the plant. No hardware is being added to the plant as part of the proposed change. The proposed change will not introduce any new accident initiators. Therefore, the change does not create the possibility of' a new or different kind of accident from any accident previously evaluated.
3. Does the change involve a significant reduction in a margin of safety?
The proposed change revises the time available to restore an inoperable PAM instrumentation channel from 7 days to 30 days before Action is required. Instead of a special report being submitted within 7 days, Action will be taken in accordance with Technical Specification 5 .6. 7 within 30 days. The other OPERABLE PAM instrument channel is available for indication, and the PAM instruments are only required to
provide indication, no active function. Therefore, the change does not involve a significant reduction in the margin of safety.
LESS RESTRICTIVE CHANGE L.5 01120/98 Palisades Nuclear Plant Page 6of11
..5?-C-
ATTACHMENT 4 NO SIGNIFICANT HAZARDS CONSIDERATION SPECIFICATION 3.3.7, PAM INSTRUMENTATION
3. Does the change involve .a significant reduction in a margin of safety?
The proposed cha.i1ge replaces the shutdown track statement with a requirement to send . a report to the NRC. The report will be issued when one required channel of PAM M instrument is inoperable and the Completion Time cannot be met. The remaining required channel or the pre-planned alternate method of monitoring is available to "" provide the required indication for PAM. The PAM instruments provides no automatic actuation functions. Therefore, the proposed change does not involve a significant reduction in a margin of safety. LESS RESTRICTIVE CHANGE L.5 CTS 3.17.4.7b requirement that alternate monitoring be initiated within 48 hours with two Reactor Vessel Water Level channels inoperable is deleted. The instrumentation which is used as an alternative to RVWL channels is also requited by CTS 3.17.4 and ITS 3.3.7. No new alternative measurement methods are initiated to implement the requirements of CTS 3.17.4.7b, rather, the Emergency Operating Procedures list alternative indications to be M used in lieu of operable RVWL channels, which provide similar, although not identical, information. The PCS temperature channels and Pressurizer Level instruments provide (\rl
information to supplement any inoperable RVWL channels. ITS 5.6.6 requires the Accident Monitoring Report to outline the preplanned alternate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels of the Function to Operable status. These controls assure that the appropriate margin of safety is maintained. This change is consistent with NUREG-1432.
1. Does the change involve a significant increase in the probability or consequences of an accident previously evaluated?
The proposed change deletes the requirement that alternate monitoring be initiated within 48 hours with two Reactor Vessel Water Level channels inoperable. The Reactor Vessel Water Level instrumentation is not an initiator of any analyzed events. The requirement to report the Reactor Vessel Water Level instrumentation preplanned alternate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the instrument channels of the Function to OPERABLE status to the NRC will ensure that alternate monitoring is initiated in a timely manner. Good operating practice and management oversight dictate that alternate monitoring be initiated as soon as possible. The Reactor Vessel Water Level instrumentation does not provide any active safety function in the mitigation of analyzed events. The proposed change does not significantly affect initiators or mitigation of analyzed events, and therefore does not involve a significant i~crease in the probability or consequences of an accident previously evaluated .
!57-d Palisades Nuclear Plant Page 5 of 9 05/30/99
I I CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO JANUARY 6, 1999 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.3, INSTRUMENTATION
NRC REQUEST:
3.3.8-01 NRC REQUEST: 3.3.8-02 Resolved during 10/27/98 meeting; Consumers Energy not required to respond. ITS LCO 3.3.8 STS LCO 3.3.12 and STS 3.3.12 Condition A are reworded to specify "one channel of each Shutdown System Function" in the corresponding ITS 3.3.8 Specification format. This rewording seems to be a presentation preference only. Comment: Reformat ITS 3.3.8 LCO, Condition A, Required Action A.1 consistent with STS 3.3.12, including Table 3.3.12-1 listing of Functions and Required number of channels. Consumers Energy Response: LCO 3.3.8, Condition A, and Table 3.3.8-1 have been reworded to more closely emulate the STS, as requested. Affected Submittal Pages: Att 1, ITS 3.3.8, page 3.3.8-1 Att 1, ITS 3.3.8, page 3.3.8-3 Att 5, NUREG, page 3.3-44 Att 5, NUREG, page 3.3-46 insert (former submittal) Att 5, NUREG, page 3.3-46 insert
58
Alternate Shutdown System
3.3 INSTRUMENTATION
3.3.8 APPLICABILITY
MODES 1, 2, and 3. ACTIONS
-------------------------------------NOTES------------------------------------
1. LCO 3.0.4 is not applicable.

2. Separate Condition entry is allowed for each Function.
CONDITION REQUIRED ACTION COMPLETION TIME ON~ oe MOV:E' efQtJlteED 1?AI.
~l)tJC.T I Of.)* IAJOPE~A8Lf. ~.3.~-02 Restore required 3.3.S-D7 ~u.>>.l.&JLD-...1.1 to OPERABLE B. Required Action and B.1 Be in MODE 3. 6 hours associated Completion Time not met. AND B.2 Be in MODE 4. 30 hours
58-~
Amendment No. 01/20/98 Palisades Nuclear Plant 3.3.8-1
Alternate Shutdown System
Table 3.3.8-1 (page 1 of 1)
Alternate Shutdown System Instrumentation and Controls FUNCTION/INSTRUMENT OR CONTROL PARAMETER
-----t ~s~*:~.;
3.3.8 Ji?U 1:~.&*02~ I
1. Source Range Neutron Flux 1

2. Pressurizer Pressure 1

3. Pressurizer Level 1

4. Primary Coolant System (PCS) 11 Hot Leg Temperature - .. J.

5. PCS #2 Hot Leg Temperature --*- 1

6. PCS #1 Cold Leg Temperature 1

7. PCS 12 Cold Leg Temperature 1

8. Steam Generator (SG) A Pressure 1

9. SG B Pressure -1

10. SG A Wide Range Level 1

11. SG B Wide Range Level J.
12.
13. 14. 15. Safety Injection Refueling Water (SIRW) Tank Level Auxiliary Feedwater (AFW) Flow Indication to SG A AFW Flow Indication to SG B AFW Low Suction Pressure Alann (P-88) 1 1 1 1
16. AFW Pump P-88 Steam Supply Valve Control 1

17. AFW Flow Control to SG A 1.

18. AFW Flow Control to SG B L
d8-*b Palisades Nuclear Plant 3.3.8-3 Amendment No. 01/20/98

3.3 INSTRUMENTATION Shutdown System (Afii1bCi' Ll.11.5] The be OPERABLE.
Shutdown System Functions 1n Table 3.3.~ shall APPLICABILITY: HODES 1, 2, and 3. ACTIONS
---------------------~--------------NOTES------------------------------------
1. LCO 3.0.4 is not applicable.

2. Separate Condition entry is allowed for each Function.
CONDITION REQUIRED ACTION COMPLETION TIME
Q.11.~.1] A. One or more required A.I Restore required 30 days RA.!.
Functions inoperable. Functions to OPERABLE 3,3,a-oi status. 3,3.8-oi [3. n .~. z.J B. Required Action and associated Completion B. l Be in MODE 3. 6 hours Time not met.
~
AHl2 B.2 Be in MODE 4. ours
'RA'!.
3.'5.B-o"l.
'S.~.a-01
(
CEOG STS 3.3-44 Rev 1, 04/07/95 58-e-

SECTION 3.3 INSERT 1 Table 3.3.8-1 (page 1 of 1)
Alternate Shutdown System Instrumentation and Controls FUNCTION/INSTRUMENT OR CONTROL PARAMETER 1. 2.
3. Pressurizer Level

4. Primary Coolant System s.

6. PCS #1 Cold Leg Temperature

7. PCS #2 Cold Leg Temperature

8. Steam Generator (SG) A Pressure

9. SG B Pressure

10. SG A Wide Range Level

11. SG B Wide. Range Level

12. Safety Injection Refueling Water

13. Auxiliary Feedwacer (
14. 15.
16. -8B Steam Supply Valve Control 17.

18. Flow Control to SG B fl AI 1* J
f'- 0 2.

3.3-46 58-J
SECTION 3.3
FUNCTION INSERT 1 Table 3.3.8-1 (page 1 of 1)
Alternate Shutdown System Instrumentation and Controls REQUIRED INSTRUMENT OR CONTROL PARAMETER CHANNELS
1. Source Range Neutron Flux 1

2. Pressurizer Pressure 1

3. Pressurizer Level 1

4. Primary Coolant System (PCS) #1 Hot Leg Temperature 1

5. PCS #2 Hot Leg Temperature 1

6. PCS #1 Cold Leg Temperature 1

7. ~ PCS #2 Cold Leg Temperature 1

8. Steam Generator (SG) A Pressure 1

9. SG 8 Pressure 1
10.
11. 12. SG A Wide Range Level SG 8 Wide Range Level Safety Injection Refueling Water (SIRW) Tank Level 1 1 1
13. Auxiliary Feedwater (AFW) Fiow Indication to SG A 1

14. AFW Flow Indication to SG 8 1

15. AFW Low Suction Pressure Alarm (P-88) 1

16. AFW Pump P-88 Steam Supply Valve Control 1

17. AFW Flow Control to SG A 1 18 . AFW Flow Control to SG 8 1 3.3-46
CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO JANUARY 6, 1999 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.3, INSTRUMENTATION NRC REQUEST: 3.3.8-03 CTS 4.17.5 DOC L.1 Ail CTS Table 4.17.5 required quarteily CHANNEL CHECKS are deleted in the corresponding ITS 3.3.8. Deletion of the CHANNEL CHECKS is based on the fact that the channels have to be energized each quarter (transferring actuation capability to the Alternate Shutdown System) in order to perform the testing. The only other required surveillance on these channels is performed every 18 months as a CHANNEL CALIBRATION. Comment: Explain what is meant by "control is removed from the control room" in L.1 in order to perform a CHANNEL CHECK surveillance requirement as required by CTS. Discuss operational hardships and safety concerns that result from the test. Consumers Energy Response: The instrumentation provided in the Palisades Alternate Shutdown System uses the same transmitter as the normal channel plant instrumentation. When the transfer switches are thrown to place the Alternate Shutdown instruments in service, the transmitter is switched out of the normal (control room) circuit and into the Alternate Shutdown System circuit. The affected control room channels fail to the zero current condition. An attachment from our quarterly test procedure, which lists the specific effects of switching to the "Alternate Hot Shutdown Panel" (AHSP), is attached. Affected Submittal Pages: No page changes .
59
Proc No Q0-23 Attachment 1 Revision 7 Page 1of3 AHSP EFFECTS ON CONTROL ROOM INSTRUMENTATION The actions which occur upon switching control to C-1501150A are as follows:
1) Auxiliary Feedwater (AFW) Flow: (FT-0727A and FT-0749A) a) Lo.s.e: Indication of AFW flows to SIG A and 8 from P-8A and P-88 on Control Room panel C-11.
b) Lo.s.e: Flow signal inputs to AFW pump starting logic. (Successful auto operation of P-8A would no longer block auto start of P-8C and P-88.) c) Lo.s.e: Palisades Plant Computer (PPC) inputs from Fl-0727 A and Fl-0749A. d) Gain: Indication of AFW flows to SIG A and 8 from P-8A and P-88 at C-150 panel.
2) AFW Flow Control: (FT-0727 and fT,;0749) a) Lo.s.e: Control of AFW flow to SIG A and 8 from P-8A and P-88 (CV-0727 and CV-0749) at both Control Room panel C-01 and remote panel C-33.
b) Gain: Control of AFW flow to SIG A and 8 from P-8A and P-88 (CV-0727 and CV-0749) at panel C-150. _3) Wide Range Steam Generator Level (LT-0757A and LT-0758A) a) Lo.s.e: Indication of Wide Range SIG A and 8 level (ie, one of two channels per SIG) in the Control Room. b) Lo.s.e: PPC inputs from SIG wide range level. (These inputs come
~ from channel A).
c) Gain: Indication of SIG A and 8 level at C-150 panel.
4) Pressurizer Level (LT-0102) a) Lo.s.e: Indication of one pressl.Jrizer level channel LIA-0102A in the Control Room. ,
b) c)
Lo.s.e: Input to the Control Room "Pressurizer Level High" alarm. Lo.s.e: PPC input pressurizer narrow range level.
.59-o-
Proc No Q0-23 Attachment 1 Revision 7 Page 2 of 3 AHSP EFFECTS ON CONTROL ROOM INSTRUMENTATION d) .Gain: indication of pressurizer ievei at C-150 panei.
5) Pressurizer Pressure (PT-0110) a) This instrument serves C-150 only and is operable whenever power is available.

6) PCS Temperature (Tl-0112HA, Tl-0112CA, Tl-0122HA, Tl-0122CA, TIA-0111H, TR-0111, TDR-0111/0121, Tl-0115) a) L.o.s..e: Temperature indications - fail high (high temperature input to TM/LP Channel A); (For TDR-0111 /TDR-0121, will NOT lose Loop 2 of delta T, For TR-0111, will NOT lose T Re1).
b) Gain: Temperature indication, Tl-0112CAA, Tl-0112HAA, Tl-0122CAA, Tl-0122HAA, at C150 Panel.
c) d)
e) L.o.s..e: Channel "A" LTOP Operability. L.o.s..e: Delta T Loop 1, Feeds SPI for POil Monitoring. Lose: Channel "A" TM/LP Operability. f) Lose: Channel "A" Variable High Power Trip Operability. g) Lose: Channel "A" Axial Shape Index Alarm Operability.
7) SIG Pressure (Pl-0751 A, Pl-0752A) a) Lose: S/G #1 and #2 pressure indications in Control Room.
b) Lose: A Channel for SIG Lo Pressure MSIV Closure Logic. c) .Gain: SIG #1 and #2 pressure indications in AHSP.
8) TAvlf Ref CONTROLLER (TYT-0100) a) Prevent Loss: This instrument provides TAve parameter to the pressurizer le'vel controllers LIC-0101 A and LIC-0101 B, and steam dump controller HIC-0780A when *selector switch SS-TAve is in the Unit 1 position.
Inputs to TYT-0100 will fail high. Switch SS-TAve should be placed in the Unit 2 position. S9-b
Proc No Q0-23 Attachment 1 Revision 7 Page 3 of 3 AHSP EFFECTS ON CONTROL ROOM INSTRUMENTATION b} I ""'"""
~* T1 ave 1rrave"') De"i ...+'1on A'"'rm 1 *IQ* IQ 1.
c) .Las.a: Tave 1 rrret Deviation Alarm . 59-e-
CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO JANUARY 6, 1999 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.3, INSTRUMENTATION
NRC REQUEST:
3.3.8-04 ITS 3.3.8 The STS SR 3.3.12.3 Note excluding the neutron detectors from the CHANNEL CALIBRATION ls deleted in ITS SR 3.3.8.4. Comment: The staff has not approved Channel Calibration definition changes. The submittal should be revised to conform to the STS. Consumers Energy Response: See response to "Definitions RAI" within this submittal. Affected Submittal Pages: See response to "Definitions RAI" within this submittal.
60
CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO JANUARY 6, 1999 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.3, INSTRUMENTATION
NRC REQUEST:
3.3.8-05 NRC REQUEST: 3.3.8-06 Resolved during 10/27/98 meeting; Consumers Energy not required to respond. Resolved during 10/27/98 meeting; Consumers Energy not required to respond. NRC REQUEST: 3.3.8-07 ITS Required Action A.1, SR 3.3.8.1 Proposed ITS Action A.1 requires "Provide equivalent shutdown capability." The language specifying the action is vague. Upon completing the actions to provide an equivalent capability it appears that the LCO is met. Comment: Provide discussion explaining the deviation from the STS action based on Palisades specific design. Consumers Energv Response: Consumers Energy agrees with this comment. The subject action is a CTS requirement, but does not fit modern Technical Specificat!on usage. The subject action has been deleted from the ITS, bringing the proposed requirements closer to those in STS. This change is addressed as "L-2."
Affected Submittal Pages:
Att Att Att Att 1, 3, 3, 3, ITS 3.3.8, page 3.3.8-1 CTS 3.3.8, page 3-72 DOC 3.3.8, page 4 of 4 (former submittal) DOC 3.3.8, page 4 of 5 Att 3, DOC 3.3.8, page 5 of 5 Att 4, NSHC 3.3.8, page 2 of 2 Att 4, NSHC 3.3.8, page 2 of 3 Att 4, NSHC 3.3.8, page 3 of 3 Att 5, NUREG page 3.3-44
61
Alternate Shutdown System 3.3.8 3.3 INSTRUMENTATION APPLICABILITY: MODES 1, 2, and 3. ACTIONS
-------------------------------------NOTES------------------------------------
1. LCO 3.0.4 is not applicable.

2. Separate Condition entry is allowed for each Function.
CONDITION REQUIRED ACTION COMPLETION TIME
A.1~21 OAJr=- oe Mot:e- ~G011eEb ru
~l)k!C..TlcW5" flJOPE~AgLf. 3'.3.<a-02 Restore required 3.3.B-01 ""'""""u.u.a......, to OPERABLE B. Required Action and B.l Be in MODE 3. 6 hours associated Completion Time not met. AND B.2 Be in MODE 4. 30 hours
0 / Amendment No. 01/20/98 Palisades Nuclear Plant 3.3.8-1
3.3.8
@1
[3.~.sJ 3 .17*
3.17.5 INSTRUMENTATION SYSTEMS Spec1f1 cat1 on The Alternate* Shutdown System 1nstrumentat1on and controls l 1sted in Table 3.17.5 shall be OPERABLE~ .....@ Note: Specificat1ons~ ~.4, ~ d-;- not apply. Aoo]lcab!llty Spec1f1cat1on 3.17.5
*~
appl1es~eiCthRCS 'Ump~ u?i:is\i: 30'§'.i i1 MODES IJ 21 a1<e..<.~- .JY
Action
~oA] 3.17.5.1 With one or more Alternate Shutdown System channels 1noperable:
a I [~A A.1.] b) Restore the inoperable channels to OPERABLE status within [t.ONb 8] 3.17.5.2 If any action required by-J.17.5.1 is not met ANO the associated completion time has expired: ,,.. """,) ,
~cD~ ~ ~" a, V a) The reactor shall be placed in eOi:;sHU'fQOW~ within
[RA e:?.J b)
Amendment No. -146, ~. 162
~.
October 26, 1994 3-72 (pl - b
SPECIFICATION 3.3.8, ALTERNATE SHUTDOWN SYSTEM TECHNICAL CHANGES - MOVEMENT OF INFORMATION TO LICENSEE-CONTROLLED DOCUMENTS (LA)
ATTACHMENT 3 DISCUSSION OF CHANGES LA.l CTS Tables 3.17.5 and 4.17.5 contain details related to Alternate Shutdown System component identification. These details are not retained in the ITS and are relocated to the Bases. These details are not necessary to ensure OPERABILITY of the Alternate Shutdown System. ITS 3.3.8 establishes the necessary requirements to ensure Alternate Shutdown System OPERABILITY, and therefore these details are not required to be in the Technical Specifications to provide adequate protection of the public health and safety. Any changes to these requirements in the Bases will require compliance with the Bases Change Control Program, as described in ITS Section 5. 0. This change is a less restrictive movement of details change with no impact on safety. This change is consistent with NUREG-1432. TECHNICAL CHANGES - LESS RESTRICTIVE (L) L.1 CTS Table 4.17 .5 provides CHANNEL CHECK Surveillance Requirements for the Alternate Shutdown System instrumentation. ISTS SR 3.3.12.1 also provides CHANNEL CHECK Surveillance Requirements for Remote Shutdown System instrumentation Functions, but requires the CHANNEL CHECK to be performed only if the required instrumentation channel is normally energized. CTS requires the check regardless of the normal energization state. However, none of the Alternate Shutdown System instrumentation is normally energized. Therefore, the ISTS bracketed SR is not included in the ITS. Since this change eliminates the quarterly CHANNEL CHECKS, the change is less restrictive than CTS. This change is appropriate and acceptable since the CTS CHANNEL CHECK performance requires placing the system in service, i.e., actuating the transfer to the Alternate Shutdown System instrumentation. This performance also removes the controls from the control room during normal operation. Obviously this is not a preferred condition. The 18 month CHANNEL CALIBRATION provides the necessary confirmation of channel OPERABILITY and is considered sufficient based on industry experience that demonstrates channel failure is rare, and on the use of less formal but more frequent checks of portions of the instrumentation channels during normal operational use of displays associated with the required channels. This change is consistent with NUREG-1432 .
Palisades Nuclear Plant Page 4 of 4
~1-c 01/20/98
ATTACHMENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.3.8, ALTERNATE SHUTDOWN SYSTEM TECHNICAL CHANGES - LESS RESTRICTIVE (L) Ll CTS Table 4.17.5 provides CHANNEL CHECK Surveillance Requirements for the Alternate Shutdown System instrumentation. ISTS SR 3.3.12.1 also provides CHANNEL CHECK Surveillance Requirements for Remote Shutdown System instrumentation Functions, but requires the CHANNEL CHECK to be performed only if the required instrumentation channel is normally energized. CTS requires the check regardless of the normal energization state. However, none of the Alternate Shutdown System instrumentation is normally energized. Therefore, the ISTS bracketed SR is not included in the ITS. Since this change eliminates the quarterly CHANNEL CHECKS, the change is less restrictive than CTS. This change is appropriate and acceptable since the CTS CHANNEL CHECK performance requires placing the system in service, i.e., actuating the transfer to the Alternate Shutdown System instrumentation. This . performance also removes the controls from the control room during normal operation. Obviously this is not a preferred condition. The 18 month CHANNEL CALIBRATION provides the necessary confirmation of channel OPERABILITY and is considered sufficient based on industry experience that demonstrates channel failure is rare, and on the use of less formal but more frequent checks of portions of the instrumentation channels during normal operational use of displays associated with the
L.2 required channels. This change is consistent with NUREG-1432 .
The two CTS 3.17 .5 Actions which are required when _one or more Alternate Shutdown System Function is inoperable have been replaced by the corresponding STS Action. The CTS Actions require: a) Provide *equivalent shutdown capability within 7 days, and b) Restore the inoperable channels to OPERABLE status within 60 days. The STS and ITS require: A.1 Restore required Functions to Operable status [within] 30 days. With CTS Actions, equivalent shutdown capability (i.e., functionality) would be restored within 7 days, but re~toration to full Operable status would not be required for 60 days. When Action 3.17.5a) was initially added to the CTS, each Function in the table carried an explicit instrument channel designator. Therefore, CTS required a) restoration of the ability to monitor the specified parameters or to provide equivalent control functions quickly, and b) restoration of the designated instrument channel to Operable status within 60 days .
Palisades Nuclear Plant LPl-d Page 4 of 5 05/30/99
ATTACHMENT 3
L.2 (continued)
DISCUSSION OF CHANGES SPECIFICATION 3.3.8, ALTERNATE SHUTDOWN SYSTEM ITS a.'ld STS each require restoration of Operability \Vithin 30 days, but make no t-
<J requirement to restore any level of functionality before that. The omission of the I requirement to restore functionality with 7 days is a Less Restrictive feature of this cf) change. That omission is considered to be acceptable based on the requirement to more quickly restore full operability, the fact that the subject equipment is not required to perform any immediate, or automatic, safety function. There are no particular design features at Palisades which would make the Completion Times approved for inclusion in STS inappropriate for use in ITS. This change is consistent with NUREG-1432 .
Palisades Nuclear Plant Page 5 of 5 05/30/99

2.
SPECIFICATION 3.3.8, ALTERNATE SHUTDOWN SYSTEM ATTACHMENT 4 NO SIGNIFICANT HAZARDS CONSIDERATION Does the change create the possibility of a new or different kind of accident from any accident previousiy evaiuated? The proposed change does not necessitate a physical alteration of the plant (no new or different type of equipment will be installed) or changes in parameters governing normal plant operation. The proposed change will still ensure proper surveillances are required for the equipment considered in the safety analysis. Thus, this change does not create the possibility of a new or different kind of accident from any accident previously evaluated.
3. Does the change involve a significant reduction in a margin of safety?
The margin of safety for the Alternate Shutdown System instrumentation is based on availability and capability of the instrumentation to provide the required information to the operator. The acceptability of the elimination of the CHANNEL CHECK Surveillance Requirement is based on unit operating experience that demonstrates channel failure is rare, and on the use of less formal but more frequent checks of portions of the channels during normal operational use of the displays associated with the required channels. Therefore, the availability and capability of the Alternate Shutdown System instrumentation continues to be assured by the remaining Surveillance Requirements and this change does not involve a significant reduction in a margin of safety .
Palisades Nuclear Plant Page 2 of 2 lo l - _f-01/20/98
ATTACHMENT 4
3.
NO SIGNIFICANT HAZARDS CONSIDERATION SPECIFICATION 3.3.8, ALTERNATE SHUTDOWN SYSTEM Does the change involve a significant reduction in a margin of safety? The margin of safety for the Alternate Shutdown System instrumentation is based on availability and capability of the instrumentation to provide the required information to the operator. The acceptability of the elimination of the CHANNEL CHECK Surveillance Requirement is based on unit operating experience that demonstrates channel failure is rare, and on the use of less formal but more frequent checks of portions of the channels during normal operational use of the displays associated with the required channels. Therefore, the availability and capability of the Alternate Shutdown System instrumentation continues to be assured by the remaining Surveillance Requirements and this change does not involve a significant reduction in a margin of safety. LESS RESTRICTIVE CHANGE L.2 The two CTS 3.17.5 Actions which are required when one or more Alternate Shutdown I System Function is inoperable have been replaced by the corresponding STS Action. The CTS I Actions require: I I
a) b)
Provide equivalent shutdown capability within 7 days, and Restore the inoperable channels to OPERABLE status within 60 days. The STS and ITS require: I I I I I A.1 Restore required Functions to Operable status [within] 30 days. I With CTS Actions, equivalent shutdown capability (i.e., functionality) would be restored within 7 days, but restoration to full Operable status would not be required for 60 days. When Action 3.17.Sa) was initially added to the CTS, each Function in the table carried an explicit instrument channel designator. Therefore, CTS required a) restoration of the ability to monitor the specified parameters or to provide equivalent control functions quickly, and b) restoration of the designated instrument channel to Operable status within 60 days. ITS and STS each require restoration of Operability within 30 days, but make no requirement to restore any level of functionality before that. The omission of the requirement to restore functionality with 7 days is a Less Restrictive feature of this change. That omission is considered to be acceptable based on the requirement to more quickly restore full operability, the fact that the subject equipment is not required to perform any immediate, or automatic, safety function. There are no particular design features at Palisades which would make the Completion Times approved for inclusion in STS inappropriate for use in ITS. This change is consistent with NUREG-1432 .. Palisades Nuclear Plant 05/30/99
ATTACHMENT 4 NO SIGNIFICANT HAZARDS CONSIDERATION SPECIFICATION 3.3.8, ALTERNATE SHUTDOWN SYSTEM
1. Does the change involve a significant increase in the probability or consequences of an accident previously evaluated?
The proposed change replaces the CTS 3.17.5 Action statements requiring restoration of functionality within 7 days and restoration of Operability within 60 days with a single ITS Action statement requirement requiring restoration of Operability within 30 days. The Alternate Shutdown System is not an initiator of any analyzed event. No analyzed accident relies upon functioning of the Alternate Sh4tdown System. Automatic and manual shutdown capability is still available from the control room. Also, the Alternate Shutdown System does not provide any immediate or automatic safety function. The proposed change does not significantly affect initiators or mitigation of analyzed accidents and therefore does not involve a significant increase in the probability or consequences of an accident previously evaluated.
2. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?
The proposed change replaces the CTS 3.17 .5 Action statements requiring restoration of functionality within 7 days and restoration of Operability within 60 days with a
single ITS Action statement requirement requiring restoration of Operability within 30 days. The change will not involve a change in design or operation of the plant. No hardware is being added to the plant as part of the proposed change. The proposed change will not introduce any new accident initiators. Therefore, the change does not create the possibility of a new O! different kind of accident from any accident previously evaluated.
3. Does the change involve a significant reduction in a margin of safety?
The proposed change replaces the CTS 3 .17. 5 Action statements requiring restoration of functionality within 7 days and restoration of Operability within 60 days with a -... single ITS Action statement requirement requiring restoration of Operability within ~ 30 days. The Alternate Shutdown System is not an initiator of any analyzed event. No analyzed accident relies upon functioning of the Alternate Shutdown System. Automatic and manual shutdown capability is still available from the control room. Also, the Alternate Shutdown System does not provide any immediate or automatic safety function. Therefore, the proposed change does not involve a significant reduction in a margin of safety .
Palisades Nuclear Plant Loi-~
Page 3 of 3 05/30/99
~fi'Ln-d) ~ ~
3.3 INSTRUMENTATION 3.3.Q Gliilid\i) Shutdown System <Afii!ID ihutdown System
~ ~ .JI"'&">
LCD ~ ~Shutdown be OPERABLE. System Functions 1n Table 3.3.~ shall APPLICABILITY: MODES 1, 2, and 3. ACTIONS
---------------------~--------------NOTES------------------------------------
1. LCO 3.0.4 is not applicable.

2. Separate Condition entry is allowed for each Function.
CONDITION REQUIRED ACTION COMPLETION TIME A. One or more required A.l Restore required 30 days RAI. Functions inoperable. Functions to OPERABLE 3,3,S*O'l. status. 3.!.8*07 [5.n.~.-z.J 8. Required Action and 8.1 Be in MODE 3. 6 hours associated Completion Time not met.
~ours AW2 8.2 Be in MODE 4.
li!lto.l:. 3,'3.&-o"l.
!.1.s-01
( CEOG STS 3.3-44 Rev 1, 04/07/95
CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS
RESPONSE TO JANUARY 6, 1999 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.3, INSTRUMENTATION

NRC REQUEST:
3.3.9-05 TSTF-136 is incorporated in ITS 3.3.9 Required Action A.2 to reflect the combination of STS 3.1.1 and 3.1.2 in the ITS (See JFD 12). Comment: TSTF-136 does not show that both SR 3.1.1.1 and SR 3.1.2.1 are retained. Instead the traveler shows these SRs were combined into a single test requirement. Consumers Energy Response: The former submittal utilized TSTF-136 to combine LCOs 3.1.1 and 3.1.2, however utilized two SOM SRs (3.1.1.1 and 3.1.1.2) to address differing SOM requirements during different MODES. A subsequent revision to Section 3.1, submitted on March 1, 1999, combines those two SRs as was done in TSTF-136. In this response, ITS LCO 3.3.9, Action A.2 has been revised to reflect the use of a single SOM SR. Affected Submittal Pages: Att 1, ITS 3.3.9, page 3.3.9-1 Att 5, NUREG, page 3.3-47 Att 6, JFD 3.3.13, page 3 of 3 (former submittal) Att 6, JFD 3.3.13, page 2 of 2
66
Neutron Flux Monf torfng Channels
3.3 INSTRUMENTATION 3.3.9 Neutron Flux Monitoring Channels 3.3.9 LCO 3.3.9 Two channels of neutron flux monitoring instrumentation shall be OPERABLE.
APP LI CAB IL ITV: ACTIONS CONDITION REQUIRED ACTION COMP LET ION TIME A. One or more required A.l Suspend all Immediately channel(s) inoperable. operations involving positive reactivity additions. Mill A.2 4 hours Once per 12 hours thereafter
Palisades Nuclear Plant
(.p(o - 3.3.9-1 Amendment No. 01/20/98
IJ e..v.+Yo"' \=luil
~:£:5...U:!!!!l>lu....i--CM.~ Hon i tori ng Channels C.T~
~:~INS;~.::~:l ~~~~C~~:.?i, &,g r-Cil-\J('._IJ_+Yl_o_~~~~lu...,,()
[!./?. 'J LCO 3.3. Two channels of [loglt1thltiit] pbWir li¥i'l>monitoring instrumentation shall be OPERABLE. 03.11.1. (1)) (}. 11.1.] APPLICABILITY: 11 ~ *11 *(. (I)] ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME [3. 11. lo' l] A. One or more required A.1 Suspend all I11111ediately channel(s) inoperable. operations involving ~ positive reactivity additions .
rum A.2 Perform SOM verification in accordance with SR 3.1.1.1, 4 hours Once per 12 hours R.Al:
3:S.9-05 thereafter SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.3.~ Perform CHANNEL CHECK. 12 hours (continued)
CEOG STS 3.3-47 Rev 1, 04/07/95

Chan&e ATTACHMENT 6 JUSTIFICATION FOR DEVIATIONS SPECIFICATION 3.3.13, POWER MONITORING CHANNELS Discussion 12 .

Palisades Nuclear Plant Page 3 of 3 01120/98
ATTACHMENT 6 JUSTIFICATION FOR DEVIATIONS SPECIFICATION 3.3.13, POWER MONITORING CHANNELS Change Discussion
9. ISTS SR 3.3. 13.2 for performance of a CHANNEL FUNCTIONAL TEST is not being proposed since the CHANNEL CHECK and CHANNEL CALIBRATION Frequencies are considered adequate to ensure the OPERABILITY of the equipment. The proposed Frequencies are consistent with the Surveillance Frequencies of the Post Accident Monitoring Instrumentation Frequencies in ITS 3.3.7, which provides indication-only Functions. Therefore, this change is considered to be consistent with NUREG-1432 for similar type instrumentation functions. This change is also supported by the current licensing basis since no periodic CHANNEL FUNCTIONAL TEST Surveillance is provided in the CTS. A CHANNEL FUNCTIONAL TEST is required just prior to each startup but its purpose is to verify monitoring capability for the startup, not for monitoring during the shutdown conditions. Therefore, the CTS CHANNEL
. FUNCTIONAL TEST is appropriately addressed in ITS 3.3.1.
10. Not used.

11. Text referring to other plant designs has been deleted to make the ITS specific to Palisades .
12. TSTF-136 is incorporated to reflect the combination of LCO 3 .1.1, Shutdown Margin
- Tavg > 200 F and LCO 3.1.2, Shutdown Margin - Tavg < = 200 F (see Section 3.1). ~
Palisades Nuclear Plant Page 2 of 2 05/30/99
~) CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO JANUARY 6, 1999 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.3, INSTRUMENTATION
NRC REQUEST:
3.3.8-08 NRC REQUEST: 3.3.8-09 Resolved during 10/27/98 meeting; Consumers Energy not required to respond. Resolved during 10/27/98 meeting; Consumers Energy not required to respond. NRC REQUEST: 3.3.8-10 Resolved during 10/27/98 meeting; Consumers Energy not required to respond. NRC REQUEST: 3.3.9-01 CTS 3.17.6 DOC R.3 provides discussion and justification for relocating CTS 3.17.6, Table 3.17.6, Functions 8, 9, 10, and 11 and associated note (a), and Table 4.17.6 Functions 8, 9, 10, and 11 and associated Note (c), containing requirements for primary safety valve position indicators, PORV position indicators, PORV block valve position indicator, and for the service water break detector respectively. Comment: The CTS 3.17 .6 Mark-up does not include Functions 10 and 11. Provide the CTS Mark-up pages that include these Functions. DOC R.3 provides adequate justification for the relocated requirements. Consumers Energy Response: The requested mark-up pages are to be found in the "3.3 Relocated" section at the back of the Section 3.3 mark-ups. See Pages 4 of 9 and 7 of 9. Affected Submittal Pages: No page changes. 62
CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO JANUARY 6, 1999 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.3, INSTRUMENTATION
NRC REQUEST:
3.3.9-02 CTS 3.17.6 DOC R.4 provides discussion and justification for relocating CTS 3.17.6, Table 3.17.6 Function 19 and associated Note (b) requiring two fuel pool area radiation monitors OPERABLE in HOT STANDBY and above. Comment: The CTS 3.17.6 Mark-up does not include the pages containing these requirements. Provide the CTS Mark-up pages that include this Function. DOC R.4 provides adequate justification for the relocated requirements. Consumers Energy Response: The requested mark-up pages are to be found in the "3.3 Relocated" section at the back of the Section 3.3 mark-ups. See Pages 5 of 9 and 7 of 9. Affected Submittal Pages: No page changes .
63
CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO JANUARY 6, 1999 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.3, INSTRUMENTATION
NRC REQUEST:
3.3.9-03 The STS SR 3.3.13.3 Note excluding the neutron detectors from the CHANNEL CALIBRATION is deleted in !TS SR 3.3.9.2. Comment: The staff has not approved Channel Calibration definition changes. The submittal should be revised to conform to the STS. Consumers Energy Response: See response to "Definitions RAI" within this submittal. Affected Submittal Pages: See response to "Definitions RAI" within this submittal.
64
CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO JANUARY 6, 1999 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.3, INSTRUMENTATION
NRC REQUEST:
3.3.9-04 CTS Table 3.17.6 DOCA.3 The CTS Source Range Monitor (SRM) applicability ("less than 1E-4% Rated Power, with fuel in the reactor") becomes applicabilities in Section 3.3 and 3.9 of the ITS. A.3 states that in the ITS format, this CTS applicability is not directly translated into the ITS MODES; however, the SRM applicability requirements would place the plant in one of following four conditions: MODE 6; MODE 2; or MODES 3, 4, or 5. MODE 6 requires two SRMs. A.3 states that MODE 2 requirements are stated in ITS 3.3.1, and that MODE 3, 4, 5 requirements are stated in 3. 9 except for those given in LCO 3.3.1. Comment: Provide additional discussion to support the A.3 .discussion that the SRM channel requirements of the CTS are unchanged (< 1E-4 %RTP and fuel in the reactor) in the proposed ITS, particularly for applicability in "MODE 2" (LCO 3.3.1) and in "MODES 3, 4, 5 with more than one rod capable of withdrawal and the PCS boron concentration at the REFUELING BORON CONCl;:NTRATION" (LCO 3.3.1 ). Comment: The ITS proposes to delete STS LCO 3.3.2 (RPS-Shutdown) yet RPS shutdown requirements are included in LCO 3.3.1. Additionally, JFD 8 does not fully justify changing the CTS and STS applicabilities to accommodate the Palisades individual control rod circuit breakers. This presentation must provide a design basis for not adopting STS format. Consumers Energy Response:
The formerly proposed Applicability for ITS LCO 3.3.9 was intended to complement the ITS LCO 3.3.1 applicability. LCO 3.3.1 requires two channels of Wide Range Nuclear Instrumentation as support for the High Startup Rate Trip. The former submittal intended to require these instruments in LCO 3.3.9 when they were not required by LCO 3.3.1 or by the Refueling LCO 3.9.2.
After further review, the Applicability of LCO 3.3.9 has been revised to simply "Modes 3, 4, and 5." That revision will result in these instrument channels being required by more than one LCO at the same time, but that condition exists for many other instrument channels which provide support for more than one function. The Bases and supporting documentation has been revised accordingly. Affected Submittal Pages: Att 1, ITS 3.3.9, page 3.3.9-1 Att 3, CTS 3.3.9, page 3-77 Att 3, DOC 3.3.9, page 2 of 4 (former submittal) Att 3, DOC 3.3.9, page 2 of 4 Att 5, NUREG, page 3.3-47 Att 6, JFD 3.3.13, page 2 of 3 (former submittal) Att 6, JFD 3.3.13, page 1 of 2
65
Neutron Flux Monitoring Channels
3.3 INSTRUMENTATION 3.3.9 Neutron Flux Mon1tor1ng Channels 3.3.9 LCO 3.3.9 Two channels of neutron flux monitoring instrumentation shall be OPERABLE.
APPL ICAB IL ITV: ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more required A.1 Suspend all Immediately channel(s) inoperable. operations involving positive reactivity additions. Mill A.2 4 hours Once per 12 hours thereafter
Palisades Nuclear Plant eo5 -
3.3.9-1 Amendment No. 01/20/98
3. 3'. '}

3. 17 INSTRUMENTATION SYSTEMS ~
Table 3.17.6 Instrumentation Ooerat1ng Reauirements for

Other Safety Functions M1n1mum Required OPERABLE App 11cab1 e Instryment Channels Channels Cond1t1ons Q.1.iJ 1. Neutron Flux Monitoring 2
[.>tppli0 2.
3. 2'.. 1

4. l/L 0

5. l/L i ne 6.
7. a.
c.
d. Humidity Monitor 1 1
8. 2/valve1* 1 alve Above 3oo*F T.,*.

9. bove 21o*F T.,.. wh Indication ~ ~lve is open or 1 position indica 'on system s inoperable.
(a) me yovisions of/Specificattefis 3.0.4 ..atld 4.0.4 ar.e"'not appl i°'ble. (b) The requir channels sha~e one channel acb of 7a~, 7c, d. (c) The nimum channels~ be any o channel of ~7b, 7c, (continued) Amendment No. 3, 67, 96, 98, 11§, ll8, 121, 124, 129, 136, 162 October 26, 1994
3-77

ATTACHi\'lENT 3 DISCUSSION OF CHAi"l"GES SPECIFICATION 3.3.9, NEUTRON FLUX MONITORING CHAi"\TNELS A.3

he CTS Applicability in CTS Table 3.17.6 for item 1 (Neutron Flux itoring) is "below 10-i3 RATED POWER with fuel in the reactor."
ITS 3.9 Applicability for this Function is MODES 3, 4 and 5, with no than on control rod capable of being withdrawn and the Primary Cool t System ( S) boron concentration less than the REFUELING BORO CONCENT TION, and also MODES 3, 4 and 5, with the Prim y Coolant System (PCS) oron concentration at the REFUELING BORON CONCENTRA N. This Applicability complements ITS 3 .1 such that either ITS 3.3.1 or TS 3.3.9 is Applicable in all of MOD 1, 2, 3, 4, and 5. The CTS Applicable onditions of "below 10-4 3 RAT POWER with fuel in ' the reactor" is not direc related to control rod with awal capability, nor to PCS boron concentration. owever, with the react "below 10*4 3 RA TED POWER with fuel in the rea r," the plant is in ne of four conditions as identified in the Applicability ~ ITS. The pl t must be in one of the following conditions: a) MODE* , b) MOD 2, or c) MODES 3, 4, or 5. MODE 6 neutron flux monitoring is ddr sed in ITS 3.9.2. MODE 2 neutron flux monitoring is addressed in ITS 3. . If the plant is in MODES 3, 4, or 5, it is also in one of the following fou on 'tions: I) more than one control rod is
capable of withdrawal and the PC oron c centration is less than the REFUELING BORON CONC TRATION; ")more than one control rod is capable of withdrawal and th CS boron conce ration is at the REFUELING BORON CONCENTRAT N; iii) no more than o control rod is capable of withdrawal and the PCS oron concentration is less t n the REFUELING BORON CONCENT TION; iv) no more than one co rol rod is capable of withdrawal and th CS boron concentration is at the RE ELING BORON CONCENTRA1: N. The first of these conditions is add re ed by ITS 3. 3 .1.
The remainin three conditions are addressed in this Specificat n, ITS 3. 3. 9. Since the C S Applicable Conditions encompass the ITS Applica "lity and also include c nditions where the power level is low but more than one ntrol rod is cap e of being withdrawn, this change is less restrictive than CT . ver, since neutron flux monitoring in the other conditions is addre ed by o er Specifications within the ITS, this change is considered administrativ
')e12.... ~ ~ I) \ ~ E. "'::> ""':>O C-jl A- I 1. ~. ~ - 0 L/
Palisades Nuclear Plant Page 2 of 4 6?'5.- C-01120/98
ATTACHMENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.3.9, NEUTRON FLUX MONITORING CHANNELS A.3 The CTS Applicability in.CTS Table 3.17.6 for item 1 (Neutron Flux Monitoring) is "below 104 % RATED POWER with fuel in the reactor." The ITS 3.3.9 Applicability for this Function is MODES 3, 4 and 5. The CTS Applicable Conditions of "below 104 % RATED POWER with fuel in the reactor" is not directly related to ITS Modes, however, with the reactor "below 104 % RATED POWER with fuel in the reactor", the plant must be in either MODE 2, 3, 4, 5, or
6. The Wide Range Neutron Flux Monitors, which provide flux monitoring when below 104 % power, are required to be operable in MODE 2 by ITS LCO 3.3.1 as the instrumentation associated with the High Startup Rate RPS trip function and in MODE 6 by ITS 3.9.2. This LCO, combined with ITS LCOs 3.3.1 and 3.9.2, require the Neutron Flux Monitoring channels to be operable whenever reactor power is below 10-4 % RTP.
Since neutron flux monitoring in the other Modes is addressed by other Specifications within the ITS, this change is considered to only address the MODES 3, 4, and 5 part of the CTS "less than 10-4 %" applicability, this change is considered to be Administrative. A.4 CTS Table 4.17.6 item 1 (Neutron Flux Monitoring) requires a CHANNEL FUNCTIONAL TEST to be performed in accordance with Footnote (a) which is once within 7 days prior to each reactor startup. This requirement is not included in the ITS 3.3.9 for this Function since it is identified as necessary for "reactor startup" which is MODE 2 in ITS and addressed, i.e., retained, for the High Startup Rate Function in ITS 3.3.1, "Reactor Protective System (RPS) Instrumentation," which utilizes these wide range instrumentation channels. The Neutron Flux Monitoring Function is required in MODES 3, 4, and 5, with the appropriate accompanying conditions, as discussed in DOC A.3, above. The plant is subcritical in these MODES, therefore performing a CHANNEL FUNCTIONAL TEST within 7 days prior to a reactor startup does not address the surveillance Frequency for these MODES of operation. Since this Function provides an indication of neutron flux levels and is not required to provide an automatic actuation Function, the current 18 month Frequency for CHANNEL CALIBRATION and 12 hour Frequency for CHANNEL CHECK are considered to be sufficient to ensure the Function remains OPERABLE to provide adequate information to the operators to indicate a change of neutron flux levels. Since the CTS CHANNEL FUNCTIONAL TEST Frequency does not address the requirements in MODES 3, 4 and 5, with no more than one control rod capable of being withdrawn, where the Neutron Flux Monitoring instrumentation is required to be OPERABLE, this change is considered to be an administrative.change. The proposed Frequencies are consistent with the Surveillance Frequencies of the Post Accident Monitoring Instrumentation Frequencies in ITS 3.3.7, which also only provide indication only Functions. Therefore this change is considered to be
consistent with NUREG~1432 for similar type instrumentation functions .
Palisades Nuclear Plant U?5-d Page 2 of 3 05/30/99
fJ e.v+v-or-. ~lui1
~~:U!iill!bLU~BJ:J Hon 1tor 1ng Channels C.T'S 3.3 INSTRUMENTATION ~eu+Yo~ ~lu~
[.!./?. 'J Two channels of [loq\t1thlttit] pbWir instrumentation shall be OPERABLE.
~nitor1ng (13.11.t. (1))
t}. 11. £.] APPLICABILITY: fr 3. 11. c. (1)] ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME [3.11. lo. l] A. One or more required A.1 Suspend al 1 I1t111ed1ately channel(s) inoperable. operations involving positive reactivity additions .
Arm A.2 Perform SOM verification in accordance with SR 3.1.1.1, 4 hours Once per 12 hours thereafter SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.3.~ Perform CHANNEL CHECK. 12 hours (continued)

CEOG STS 3.3-47 Rev 1, 04/07/95

Change SPECIFICATION 3.3.13, POWER MONITORJNG CHANNELS Discussion ATTACHMENT 6 JUSTIFICATION FOR DEVIATIONS 8.
could occur in these conditions and, therefore, neutron fl
dication is necessary so ttia e plant operator can take the appropriate acti . The Palisades design does not include r or trip circuit breakers, ho er it does have individual control rod circuit breakers. er these
. circu ces, the capability to withdraw one control rod provides flexibility i ucting control rod Surveillances and tests.
9. ISTS SR 3.3.13.2 for performance of a CHANNEL FUNCTIONAL TEST is not being proposed since the CHANNEL CHECK and CHANNEL CALIBRATION Frequencies are considered adequate to ensure the OPERABILITY of the equipment. The proposed Frequencies are consistent with the Surveillance Frequencies of the Post Accident Monitoring Instrumentation Frequencies in ITS 3. 3. 7, which provides indication-only Functions. Therefore, this change is considered to be consistent with NUREG-1432 for similar type instrumentation functions. This change is also supported by the current licensing basis since no periodic CHANNEL FUNCTIONAL TEST Surveillance is provided in the CTS. A CHANNEL FUNCTIONAL TEST is required just prior to each startup but its purpose is to verify monitoring capability for the startup, not for monitoring during the shutdown conditions. Therefore, the CTS CHANNEL FUNCTIONAL TEST is appropriately addressed in ITS 3.3.1.

10. The Note excluding neutron detectors from CHANNEL CALIBRATION is deleted.
This exclusion has been incorporated into the definition of CHANNEL CALIBRATION.
11. Text referring to other plant designs has been deleted to make the ITS specific to Palisades.
J.1.13 #"f
Palisades Nuclear Plant Page 2 of 3 (oCS-J-01/20/98
ATTACHMENT 6
Change Note:
Discussion JUSTIFICATION FOR DEVIATIONS SPECIFICATION 3.3.13, POWER MONITORING CHANNELS rnis 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.
5. 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. This change reflects the current licensing basis/technical specification.
6. The word Analog was removed from the title of each specification in Section 3.3.
Palisades is an Analog plant and specifically listing this in the title is unnecessary. The terms, "Analog" and "digital," were placed in NUREG-1432 to distinguish between the Specifications for the two types of instrumentation.
7. ISTS 3.3.2, 3.3.8, 3.3.9 and 3.3.10 are not included in the ITS, therefore ISTS 3.3.13 has been renumbered as 3.3.9. The Specification, LCO and Surveillances have been renumbered, where applicable, to reflect these deletions.

8. The Applicability of the Specification has been modified to be consistent with the current Technical Specifications to ensure the neutron flux indicators are available whenever the plant is shutdown but not in MODE 6, MODES 3, 4, and 5 (MODE 6 is covered by ITS LCO 3.9.2). The Palisades design does not include reactor trip circuit breakers. While the requirements of ITS LCO 3.3.9 will somewhat overlap those of ITS LCO 3.3.1, many instrument channels fall under the operability requirements of several LCOs. Specifying the LCO 3.3.9 applicability as MODES 3, 4, and 5 assures that these instruments will be operable when required .
Palisades Nuclear Plant 05/30/99
'II CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO JANUARY 6, 1999 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.3, INSTRUMENTATION
NRC REQUEST:
3.3.9-06 DOC A.2 CTS 3.17.6.1 This discussion justifies eliminating the CTS Action to place the plant in Hot Shutdown within 15 minutes if neutron monitoring channels are inoperable. The justification relates that CTS requirements are moved to ITS LCO 3.1.1 because LCO 3.1.1 requirements for Shutdown Margin not met applies the same time frame (15 minutes) to restore shutdown margin. Comment: The DOC is unclear. Elaborate on the administrative nature of the change. Consumers Energy Response: DOC 3.3.9 A.2 has been revised as requested. Affected Submittal Pages: Att 3, DOC 3.3.9, page 1 of 4 (former submittal) Att 3, DOC 3.3.9, page 1 of 3
67

ATTACHMENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.3.9, NEUTRON FLUX MONITORING CHANNELS ADMINISTRATIVE CHANGES (A)
A. l 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 CTS 3.7.16.1 requires the plant to be in HOT SHUTDOWN or belo within inutes with one or two neutron flux monitoring channe
operable .
ITS 3.3.9 does include this requirement. The CTS defi *
n of HOT SHUTDOWN is that eactor is subcritical by an nt greater than or equal to the margin in Techn Specification and Tave is greater than 525°F. The requirements of CTS . e been incorporated in the Reactivity Control System Specification ITS .. 1, OWN MARGIN." In this Specification, if SDM is n et the plant is reqm limits in 15 minute erefore since the requirement to ore SDM is within the same tim me in both the CTS and the ITS, even though urrent requir nts are included in a different Specification, this change is co inistrative. This change is consistent with NUREG-1432 .

Palisades Nuclear Plant Page 1of4 u7 01/20/98
ATTACHMENT 3
ADMINISTRATIVE CHANGES (A)
A.l DISCUSSION OF CHANGES SPECIFICATION 3.3.9, NEUTRON FLUX MONITORING CHANNELS AH reformatting and renumbering are in accordance with t-.TTuREG-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 If there are less than two operable channels of nuclear instrumentation, CTS 3.17.6.lb requires that the plant must be placed in HOT SHUTDOWN or '° 0 below within 15 minutes; ITS 3.3.9 does not include this requirement. Since ' v-- the applicability of LCO 3.3.9 is "MODES 3, 4, and 5" the requirements of VI CTS Action 3.17.6.lb would already be met. Requirements for nuclear r instrument channels in Modes 1 and 2 are addressed by LCO 3.3.1. This change is consistent with NUREG-1432 .
Palisades Nuclear Plant Page 1of3 05/30/99
CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO JANUARY 6, 1999 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.3, INSTRUMENTATION
NRC REQUEST:
3.3.10-1 ITS 3.3.10 CTS Table 3.17.3 . DOCA.2 The ITS proposes a new LCO for Engineered Safeguards Room Ventilation instrumentation. This instrumentation isolates a pump room on a high radiation signal. In the CTS this Function is listed with other ESF isolation instrumentation. Comment: The justification for this proposed change states "an additional Specification is included ....... to ensure an assumption of the radiological consequences analysis of the LOCA is maintained. The staff notes that ITS specifications 3.3.3 and 3.3.4 contain the other ESF Functions from CTS Table 3.17.3, including radiation monitors which actuate valve closure signal for example on the MSIVs and the MFIVs. The staff questions the basis for the additional specification, and recommends instead including the ESRV Instrumentation requirements in ITS LCOs 3.3.3 and 3.3.4. As a note to the construction of the proposed LCO 3.3.10, separate condition entry is not needed because the same action is required for one or both required channels of ESRV inoperable. Consumers Energy Response: ITS 3.3.10 is, in effect, the Palisades equivalent of STS LCO 3.3.9. That isolation signal LCO (STS 3.3.9) was not included in STS LCOs for ESFAS instrumentation.
The separate condition entry allowance is appropriate since each channel serves to isolate a different Engineered Safeguards Room. Without the separate condition entry allowance, separate completion times would not be assigned to each failure. LCO 3.6.3, Containment Isolation Valves, where each valve serves to isolate a separate penetration, provides a similar example. This usage is explained in Section 1.3 of the ITS.
The 3.3.10 Bases discussion of the subject note has been revised. Affected Submittal Pages: ATT 2, page B 3.3.10-3
68
ESRV Instrumentation B 3.3.10
BASES ACTIONS (continued)
In the event a channel's trip setpoint is found nonconservative with respect to the Allowable Value, or the transmitter, instrument loop, signal processing electronics~ or RPS bistable trip unit is found inoperable, then all affected Functions provided by that channel must be declared inoperable, and the plant must enter the Condition for the particular protection Functions affected. A Note has been added to the ACTIONS to clarify the application of the Completion Time rules. The Conditions of S1 ~ 1 ,IL. "'11-c.t< c"'"""'"'° this Specification may be entered independently for each c Ht41JtJ6'- Seeu*":. iD t!l.o-c.,...Tl2- ~1-ffte~*~. The Completion Times. of each inoperable .f1::1net; &n cH~HMtt.L..
,.. ~it&~*~".,.. wi 11 be tracked separately1 for ead1 f'1.1Rsti ~ starting from f.,J&.. S,.Feda"~,,~ i?oet1. the ti me the Condition was entered.
Condition A addresses the failure of one or both ESRV Instrumentation high radiation monitoring channels. Operation may continue as long as action is immediately initiated to isolate the ESRV System. With the inlet and exhaust dampers closed, the ESRV Instrumentation is no longer required since the potential pathway for radioactivity to escape to the environment has been removed. The Completion Time for this Required Action is commensurate with the importance of maintaining the ES pump room atmosphere isolated from the outside environment when the ES pumps are circulating primary coolant. SURVEILLANCE SR 3.3.10.1 REQUIREMENTS Performance of the CHANNEL CHECK once every 12 hours ensures that a gross failure of instrumentation has not occurred. A CHANNEL CHECK is normally a comparison of the parameter indicated on one channel to a similar parameter on other channels. It is based on the assumption that instrument channels monitoring the same parameter should read approximately the same value. Significant deviations between the two instrument channels could be an indication of excessive instrument drift in one of the channels or of something even more serious.* CHANNEL CHECK will detect gross channel failure; thus, it is key to verifying the instrumentation continues to operate properly between each CHANNEL CALIBRATION. Palisades Nuclear Plant B 3.3.10-3 01/20/98 lo 8 -
ENCLOSURE 3 CONSUMERS ENERGY COMPANY PALISADES PLANT DOCKET 50-255
CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO JANUARY 6, 1999 NRC COMMENTS CONCERNING ITS SECTION 3.3 SECTION 1.0, REPLACEMENT PAGES R_e_ct2-l veJ w -~ Le...~
cla:k-e& o& / 11 /Ci.Ci * ,, ~ qq6bl<i?Oll5
CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO JANUARY 6, 1998 REQUEST FOR ADDITIONAL INFORMATION REVISED PAGES FOR SECTION 1.0

Page Change Instructions Revise the Palisades submittal for conversion to Improved T~chnical 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 1.1-2 ITS 1.1-2 05/30/99 Definitions RAI ATTACHMENT 2 TO ITS CONVERSION SUBMITTAL No page changes. ATTACHMENT 3 TO ITS CONVERSION SUBMITTAL CTS 1.0, pg 1-1 CTS 1.0, pg 1-1 N.A. Definitions RAI CTS 1.0, pg 1-2 CTS 1.0, pg 1-2 N.A. Definitions RAI DOC 1.0, pg 2of12 DOC 1.0, pg 2of12 05/30/99 Definitions RAI DOC 1.0, pg 8of12. DOC 1.0, pg 8of12 05/30/99 Definitions RAI ATTACHMENT 4 TO ITS CONVERSION SUBMITTAL No page changes. ATTACHMENT 5 TO ITS CONVERSION SUBMITTAL NUREG 1.0, pg 1.1-2 NUREG 1.0, pg 1.1-2 N.A. Definitions RAI ATTACHMENT 6 TO ITS CONVERSION SUBMITTAL JFD 1.0, pg 3 of 3 JFD 1.0, pg 3 of 3 05/30/99 Definitions RAI
Defi niti ans 1.1
1.1 Definitions
- CHANNEL CALIBRATION A CHANNEL CALIBRATION shall be the adjustment, as necessary, of the channel output such that it responds within the necessary range and accuracy to known values of the parameter that the channel monitors. The CHANNEL CALIBRATION shall encompass all devices in the channel required for channel OPERABILITY and the CHANNEL FUNCTIONAL TEST. I Calibration of instrument channels with Resistance I Temperature Detector (RTD) or thennocouple sensors may consist of an inplace qualitative assessment of sensor behavior and nonnal calibration of the remaining adjustable devices in the channel.
Whenever an RTD or thennocouple sensing element is I replaced, the next required CHANNEL CALIBRATION I shall include an inplace cross calibration that I* compares the other sensing elements with the I recently installed sensing ~lement. I 1 The CHANNEL CALIBRATION may be perfonned by means of any series of sequential, overlapping, or total channel steps. CHANNEL CHECK A CHANNEL CHECK shall be the qualitative assessment, by observation, of channel behavior during operation. This detennination shall include, where possible, comparison of the channel indication and status to other indications or status derived from independent instrument channels measuring the same parameter. CHANNEL FUNCTIONAL TEST A CHANNEL FUNCTIONAL TEST shall be:
a. Analog and bistable channels-the injection of a simulated or actual signal into the channel as close to the sensor as practicable to verify OPERABILITY, of all devices in the channel required for channel OPERABILITY;

b. Digital channels-the use of diagnostic programs to test digital hardware and the injection of simulated process data into the channel 'to verify OPERABILITY, of all devices in the channel required for channel
Palisades Nuclear Plant OPERABILITY .
1.1-2 Amendment No. 05/30/99.
Spec1f1cat1on 1.1
---------------NOTE----------------
The defined terms of this section
...------1 appear in capitalized type and ore oppl1coble throughout these Techmcol Spec1f1cotions ond Bases ,-----------~ TECHNICAL SPECIFICATIONS It.0 Use ond Applicot1on l .@1) DEFINITIONS AVERAGE DISINTEGRATION ENERGY - E AVERAGE DISINTEGRATION ENERGY shall be the average (weighted in proportion to the concentration of each radionuclfde in the reactor coolant at the time of sa11pling) of the sum of the average beta and gilAllla energies per disintegration (in l'EV) for isotopes, other than iodines, with half lives greater than 15 11inutes, making up at- least 95% of the total noniodine activity in the coolant *
I Split into seporate definitions AXIAL OFFSET ~APE INPEX - AO 12!J ASI '-
AXIAL OFFSET or AXIAL SHAPE INDEX shall be e r o the power generated in the lower half of the core nus t e power generated in the upper ha 1f of the core, o the sum o ose powe "'g__e_n_e_r_o..,...te--r-1-n--:-r-...------r------ divided by hol ves of the CHAlfNEL CALIBRATION ASI, all devices in the channel required for chonnel operob1hty ond
'DEF' RAI
Amendment No.
1-1
** .a, 54, 5-1-, 68, tt8, tr4, m, T3T, ~. 174 I
I I L
Spec1f1cat1on 1.1
b. D191tol chonnels - the use of d1ognostic progroms to test d1g1tol hordwore ond the in1ect1on of simulated process doto into the chonnel to ven f y OPERABILITY of oll devices in the chonnel required for channel OPERABILITY.
'DEF' RA!
1~ QEEINITIONS (continued) CHANNEL EUNCTIQNAL TEST a.Analog ond b1stoble channels - A CHANNEL FUNCTIONAL TEST shall b he 1nject1on of a simulated ~inal or i:llt~ fl' channe 1 to verif7 that it is OPERABLE, f1 ~ainrany;a I a aMj _1 __ iti;J'tinq,'func)1on._ Z 7 Z Z LL L LJ os close to th_e sensor os proct1coble of oll devices in the channel COLP SHUTDOWN required for chonnel OPERABILITY: 5 MODE The COLD SHUTDOWN condition shall be when the primary coolant SHUTDOWN BORON CONCENTRATION and T1 ft is less than 210°E
CORE OPERATING LIMITS REPORT (COLR)
The COLR 1s the document that provides cycle specific parameter limits for the current reload cycle. These cycle specific parameter limits
~hall be deten1ined for each reload cycle in accordance with . Spec1ficat1on~6.5. Plant operation within these limits is addressed r;:';\ in individual pecif1cations. ~ 5 POSE EQUIVALENT I-131 DOSE EQUIVALENT I-131 shall be that concentration of I-131 (µCi/gm) which alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131, I-132, I-133,. I-134 and I-135 actually present. The thyroid dose conversion factors used for this calculation shall be those listed 1n Table Ill of TID-14844,t"!i~lat1on of)
Distance factors for Power and Test Reactor Sites.* AEC, 1962 @ 1-2 Amendment No. 3-1, ~. 5'4, 9, 68, He, *4, *8, ffi, ~.--rr-4, 184
A.4 ATTACHMENT 3 DISCUSSION OF CHANGES CHAPTER 1.0, USE AND APPLICATION The CTS definition of CHANNEL CHECK states "A CHANNEL CHECK shall include verification that the monitored parameter is within limits imposed by the Technical Specifications." This sentence was originally added to the CTS to address the problem wherein the TS contained requirements that various parameters be within a particular limit but there was not a corresponding surveillance requirement specified to verify the limit was being met. By adding these words to the definition of CHANNEL CHECK in the CTS, the CHANNEL CHECK would not only verify channel operability but also that the parameter was within* limits. By adopting TS which are modeled after NUREG-1432, there is no need to have this "surveillance requirement" specified as part of the CHANNEL CHECK requirement since there will be a separate surveillance requirement specified which requires that the parameter be verified within limit. Therefore, there is no change in requirements, only in presentation of requirements and this is considered to be an administrative change. This change is consistent with NUREG-1432.

A.5 The CTS definition of CHANNEL FUNCTIONAL TEST is expanded in the ITS to provide further descriptive information for Analog and bistable channels, and to add a

discussion for digital channels. To address digital channels, the following wording is added to the defin~tion for CHANNEL FUNCTIONAL TEST: "the use of diagnostic programs to test digital hardware and the injection of simulated process data into the channel to verify OPERABILITY, of all devices in the channel required for channel OPERABILITY." This section is added to specify the appropriate requirements for digital equipment which has been added to the original plant design.
The existing CTS definitfon relates to Analog and bistable channels and has been expanded in the proposed ITS to further describe components of a channel by using the wording "of all devices in the channel required for channel OPERABILITY." The phrase "as close to the sensor as practicable" is added following the phrase "into the channel" to make it clear where the simulated signal is to be injected. The phrase " The CHANNEL FUNCTIONAL TEST may be performed by means of any series of sequential, overlapping, or total channel steps so that the entire channel is
tested." is also added to the CTS to provide clarification that as long as the entire channel is tested, the testing can be broken up into different tests.
These changes are administrative in that they provide descriptive or explanatory information in addition to that which is contained in the CTS to clarify their application. These changes are consistent with NUREG-1432 .
Palisades Nuclear Plant Page 2of12 05/30/99

ATTACHMENT 3 DISCUSSION OF CHANGES CHAPTER 1.0, USE AND APPLICATION A.17 CTS definition of "OPERABLE-OPERABILITY" includes "electrical power" but does not specify whether it is normal or emergency power. In the proposed ITS definition for "OPERABLE" the words "normal or emergency" are added to clarify that either is acceptable for determining OPERABILITY. Section 3.8 of the proposed ITS addresses actions to take on loss of an off-site circuit or emergency diesel generator and any other actions which must be taken for the supported systems. This structure clarifies that there is no need to declare all supported equipment from electrical power sources inoperable upon loss of either the normal or emergency power source.
The addition of the words "normal or emergency" is considered an administrative change since it simply clarifies the existing application of the CTS definition of "OPERABLE-OPERABILITY." This change is consistent with NUREG-1432. A.18 The CTS definition of Quadrant Power Tilt - T q states in part " ... shall be the algebraic ratio .... " The word algebraic is replaced in the proposed ITS with "maximum positive." This is an administrative change to more correctly reflect that the Quadrant Power Tilt Ratio will be expressed in terms of a positive value.
A.19 The CTS does not include explanatory material related to logical connectors, completion times, or frequencies. The proposed ITS adds a discussion of each of these topics to standardize the use and application of the TS. The proposed sections to be included in the ITS are 1.2, Logical Connectors, 1.3 Completion Times, and 1.4 Frequencies. The addition of this information is considered to be an administrative change since it is simply explaining the rules which are used to develop and use the ITS. This change is consistent with NUREG-1432.
A.20 The CTS definition for "Channel Calibration" states in part" ... The CHANNEL CALIBRATION shall encompass the entire channel including the sensor, alarm, interlock, .... " The proposed ITS uses the words "all devices in the channel required for channel OPERABILITY." This change is made to clarify that only the "required" components in the channel (meaning those required by the ITS) must have the channel calibration. This is an administrative change since the wording is changed for clarification of the requirements and does not change the requirements themselves. This change is consistent with NUREG-1432, as modified by TSTF-205 .
Palisades Nuclear Plant Page 8of12 05/30/99
Definitions
/Sn=-- ~ 1.1 20$ L1.!v ALL 'DE-VIG~ \~ ~E:- C:Ata.tJ~E:-L ~UIK"E'D
1.1 Definitions CHANNEL CALIBRATION (continued)
~oi<. C.~.A.t.ll-l~L 1')PE-'f?.~E \ **-! T'r ~t>
arm, is l el , A'nc l u ng a tri func ans and incl e the CHANNEL FUNCTIONAL TEST. Calibration
.DEF t<~:r of instrument channels with resistance temperature detector (RTO) or thermocouple sensors may consist of an inplace qualitative assessm~nt of sensor @l behavior and normal calibration of the remaining ~
adjustable devices in the channel. Whenever a ,-~~~F\.E sensing element is replaced, the next required------ CHANNEL CALIBRATION shall include an inplace cross calibration that compares the other sensing elements with the recently instilled sensing element. The CHANNEL CALIBRATION may be performed . by means of any series of sequential, overlapping, or total channel steps so that the entire channel is calibrated.
CHANNEL CHECK A CHANNEL CHECK shall be the qualitative 1 assessment, by observation, of channel behavior during operation. This determination shall include, where possible, comparison of the channel indication and status to other indications or status derived from independent instrument channels measuring the same parameter.
CHANNEL FUNCTIONAL TEST A CHANNEL FUNCTIONAL TEST shall be:
a. Analog and bistable channels-the
-rsn=-
2D'5
© b.
The CHANNEL FUNCTIONAL TEST may be performed by means of any series of sequential, overlapping, or total channel steps so that the entire channel is tested .
CEOG STS 1.1-2 (continued)
Rev 1, 04/07/95
14.
ATTACl'ENT 6 JUSTIFICATIONS FOR DEVIATIONS CHAPTER 1.0, USE AND APPLICATION The Palisades CTS contains the term "Quadrant Power Tilt (Tq)" and this term is also included in the proposed ITS. The Quadrant Power Tilt is defined as "Tq shall be the maximum positive ratio of the power generated in any quadrant minus the average quadrant power, to the average quadrant power."
15. The wording of the Identified Leakage definition has been altered to clarify that leakage which might affect the operation of leakage detection systems must be classified as unidentified leakage. It is believed that this is the intent of the STS definition.

16. The Channel Functional Test and Channel Calibration definitions have been revised to reflect the changes made by TSTF 205.

17. The words "RTD or thermocouple" have been added to the forth sentence of the Channel Calibration definition to assure it is understood that the requirements of that sentence only apply to those types of sensor. This is an editorial change made for clarification only .
*Palisades Nuclear Plant Page 3of3 05/30/99
ENCLOSURE 4 CONSUMERS ENERGY COMPANY PALISADES PLANT DOCKET 50-255
CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO JANUARY 6, 1999 NRC COMMENTS CONCERNING ITS SECTION 3.3 SECTION 3.3, REPLACEMENT PAGES
CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO JANUARY 6, 1998 REQUEST FOR ADDITIONAL INFORMATION REVISED PAGES FOR SECTION 3.3 Page Change Instructions Due to the number of ITS Section.3.3 pages affected by our response to the NRC staff comments, replacement pages have been provided for all of Section 3.3(with the exception of the marked up STS bases which have nor been remarked). PAGES TO BE REPLACED: REV DATE ATTACHMENT 1 TO ITS SECTION 3.3 CONVERSION SUBMITTAL All pages of Attachment 1 are replaced 05/30/99 ATTACHMENT 2 TO ITS SECTION 3.3 CONVERSION SUBMITTAL All pages of Attachment 2 are replaced 05/30/99 ATTACHMENT 3 TO ITS SECTION 3.3 CONVERSION SUBMITTAL All pages of Attachment 3 are replaced 05/30/99 ATTACHMENT 4 TO ITS SECTION 3.3 CONVERSION SUBMITTAL All pages of Attachment 4 are replaced 05/30/99 ATTACHMENT 5 TO ITS SECTION 3.3 CONVERSION SUBMITTAL AllJ.&.Q Pages of Attachment 5 for (3.3-1 through 3.3-48) are replaced N.A.
ATTACHMENT 6 TO ITS SECTION 3.3 CONVERSION SUBMITTAL All pages of Attachment 6 are replaced 05/30/99
RPS Instrumentation 3.3.1
3~3 INSTRUMENTATION 3.3.1 Reactor Protective System (RPS) Instrumentation LCO 3.3.1 Four RPS trip units, associated instrument channels, and associated Zero Power Mode (ZPM) Bypass removal channels for each Function in Table 3.3.1-1 shall be OPERABLE.
APPLICABILITY: According to Table 3.3.1-1. ACTIONS
-------------------------------------NOTE-------------------------------------
Separate Condition entry is allowed for each Function. CONDITION REQUIRED ACT ION COMPLETION TIME A. ----*----NOTE---------- A.1 Place affected trip 7 days Not applicable to High unit in trip. Startup Rate, Loss of I Load, or ZPM Bypass ' I Removal Functions: I One or more Functions with one RPS trip unit or associated instrument channel inoperable. B. One High Startup Rate B.1 Restore trip unit and Prior to trip unit or associated instrument entering MODE 2 associated instrument channel to OPERABLE from MODE 3 channel inoperable. status. C. One Loss of Load trip C.l Restore trip unit and Prior to unit or1associated associated instrument increasing instrument channel channel to OPERABLE THERMAL POWER to inoperable. status. ~ 17% RTP following entry into MODE 3
Palisades Nuclear Plant 3.3.1-1 Amendment No. 05/30/99
RPS Instrumentation 3.3.1
ACTIONS D.
CONDITION One or more ZPM Bypass D.1 REQUIRED ACT ION Remove the affected COMPLETION TIME Immediately Removal channels ZPM Bypasses. inoperable. D.2 Declare affected trip Immediately units inoperable. E. --------NOTE---------- ------------NOTE------------- Not applicable to LCO 3.0.4 is not applicable. ZPM Bypass Removal Function.
--------------~------- E.1 Place one trip unit 1 hour in trip.
One or more Functions with two RPS trip AND units or associated instrument channels ------------NOTE-------------
inoperable. Not applicable to High Startup Rate or Loss of Load Functions.
E.2 Restore one trip unit 7 days and associated instrument channel to OPERABLE status. F. Two power range F.1 Restrict THERMAL 2 hours channels inoperable. POWER to s 70% RTP .
Palisades Nuclear Plant 3.3.1-2 Amendment No. 05/30/99
RPS Instrumentation
ACTIONS CONDITION REQUIRED ACTION 3.3.1 COMP LET ION TIME G. Required Action and G.1 Be in MODE 3. 6 hours associated Completion Time not met.
G.2.1 Verify no more than 6 hours one full-length Control room ambient control rod is air temperature capable of being
> 90°F. withdrawn.
OR G.2.2 Verify PCS boron 6 hours concentration is at REFUELING BORON CONCENTRATION .
-------------------------------------NOTE-------------------------------------
Refer to Table 3.3.1-1 to determine which SR shall be performed for each Function. SURVEILLANCE FREQUENCY SR 3.3.1.1 Perform a CHANNEL CHECK. 12 hours SR 3.3.1.2 Verify control room temperature is ~ 90°F. 12 hours
Palisades Nuclear Plant 3.3.1-3 Amendment No. 05/30/99
RPS Instrumentation 3.3.1
SURVEILLANCE REQUIREMENTS SR 3.3.1.3 SURVEILLANCE
-------------------NOTE--------------------
FREQUENCY Not required to be performed until 12 hours after THERMAL POWER is ~ 15% RTP. Perform calibration (heat balance only) and 24 hours adjust the power range excore and ~T power channels to agree with calorimetric calculation if the absolute difference is
~ 1.5%.
SR 3.3.1.4 -------------~-----NOTE-------------------- Not required to be performed until 12 hours after THERMAL POWER is ~ 25% RTP. Calibrate the power range excore channels 31 days using the incore detectors .
SR 3.3.1.5 Perform a CHANNEL FUNCTIONAL TEST and verify the Thermal Margin Monitor Constants.
92 days SR 3.3.1.6 Perform a calibration check of the power 92 days range excore channels with a test signal. SR 3.3.1.7 Perform a CHANNEL FUNCTIONAL TEST of High Once within Startup Rate and Loss of Load Functions. 7 days prior to each reactor startup
Palisades Nuclear Plant 3.3.1-4 Amendment No. 05/30/99
RPS Instrumentation 3.3.1
SURVEILLANCE REQUIREMENTS SR 3.3.1.8 SURVEILLANCE
-------------------NOTE--------------------
FREQUENCY Neutron detectors are excluded from the CHANNEL CALIBRATION. Perfonn a CHANNEL CALIBRATION. 18 months
Palisades Nuclear Plant 3.3.1-5 Amendment No. 05/30/99
Palisades Nuclear Plant 3.3.1-6 Amendment No. 05/30/99 RPS Instrumentation
Table 3.3.1-1 (page 2 of 2)
Reactor Protective System Instrumentation APPLICABLE SURVEILLANCE 3.3.1 FUNCTION MODES REQUIREMENTS ALLOWABLE VALUE
9. Thermal Margin/
Low Pressure Trip~l 1, 2. 3(*). 4(*). 5(*) SR 3.3.1.1 Table 3.3.1-2 SR 3.3.1.2 SR 3.3.1.3 SR 3.3.1.4 SR 3.3.1.5 SR 3.3.1.6 SR 3.3.1.8
10. Loss of Load Trip 1(d) SR 3.3.1.7 NA SR 3.3.1.8

11. Containment High Pressure Trip 1,2,3(*) 1 4(*) 1 5C*l SR 3.3.1.5 ::: 3.70 psi g SR 3.3.1.8

12. Zero Power Mode Bypass Automatic Removal 1, 2. 3(*) *4(*). 5(*) SR 3.3.1.8 NA
(a) With more than one full-length control rod capable of being withdrawn and PCS boron concentration less than REFUELING BORON CONCENTRATION.
(c) Trips may be bypassed when Wide Range Power is automatically removed when Wide Range Power is
~
lE-4% RTP. lE-4% RTP. Bypass shal 1 be (d) When THERMAL POWER is ~ 17% RTP .
Palisades Nuclear Plant 3.3.1-7 Amendment No. 05/30/99
RPS Instrumentation
Table 3.3.1-2 (page 1 of 1)
Thennal Margin/Low Pressure Trip Function Allowable Value 3.3.1 The Allowable Value for the Thermal Margin/Low Pressure Trip, Ptrip* is the higher of two values, Pmin and Pvar* both in psia: Pmin = 1750 Pvar = 2012(QA) (QR 1 ) + 17 .O(T 10 ) - 9493 Where: QA = - 0.720(ASI) + 1.028; when - 0.628 ~ ASI < - 0.100 QA = - 0.333(ASI) + 1.067; when - 0.100 ~ ASI < + 0.200 QA = + 0.375(ASI} + 0.925; when + 0.200 s ASI s + 0.565 ASI = Measured ASI when Q :<: 0.0625 ASI = 0.0 when Q < 0.0625 QR 1 = 0.412(Q} + 0.588; when Q ~ 1.0 QR1 = Q;, when Q > 1.0 Q = THERMAL POWER/RATED THERMAL POWER T10 = Maximum primary coolant inlet temperature, in °F ASI, T10 and Q are the existing values as measured by the associated instrument channel *
Palisades Nuclear Plant 3.3.1-8 Amendment No. 05/30/99
RPS Logic and Trip Initiation
3.3 INSTRUMENTATION 3.3.2 Reactor Protective System (RPS) Logic and Trip Initiation 3.3.2 LCO 3.3.2 Six channels of RPS Matrix Logic, four channels of RPS Trip Initiation Logic, and two channels of RPS Manual Trip shall be OPERABLE.
APPLICABILITY: MODES 1 and 2, MODES 3, 4, and 5, with more than one full-length control rod capable of being withdrawn and Primary Coolant System (PCS) boron concentration less than REFUELING BORON CONCENTRATION. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A* One Matrix Logic A.1 Restore channel to 48 hours channel inoperable. OPERABLE status. B. One channel of Trip B.1 De-energize the 1 hour Initiation Logic affected clutch power - inoperable.
\
supplies.
c. f One channel of Manual C.1 Restore channel to Prior to Trip inoperable. OPERABLE status. entering MODE 2 from MODE 3
Palisades Nuclear Plant 3.3.2-1 Amendment No. 05/30/99
RPS Logic and Trip Initiation 3.3.2
ACTIONS CONDITION D. Two channels of Trip D.1 REQUIRED ACT ION De-energize the COMPLETION TIME Immediately Initiation Logic affected clutch power affecting the same supplies.
trip leg inoperable. E. Required Action and E.1 Be in MODE 3. 6 hours associated Completion Time not met. AND QR E.2.1 Verify no more than 6 hours one full-length One or more Functions control rod is with two or more capable of being Manual Trip, Matrix withdrawn. Logic or Trip Initiation Logic QR channels inoperable for reasons other than E.2.2 Verify PCS boron 6 hours
Ccinditi on D. concentration is at REFUELING BORON CONCENTRATION.
SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.3.2.1 Perfonn a CHANNEL FUNCTIONAL TEST on each 92 days RPS Matrix Logic channel and each RPS Trip Initiation Logic channel. SR 3.3.2.2 Perfonn a CHANNEL FUNCTIONAL TEST on each Once within RPS Manual Trip channel. 7 days prior to each reactor startup
Palisades Nuclear Plant 3.3.2-2 Amendment No.

05/30/99
ESF Instrumentation 3.3.3
*- 3.3 INSTRUMENTATION 3.3.3 Engineered Safety Features (ESF) Instrumentation LCO
3.3.3 Four ESF bistab*les and associated instrument channels for each Function in Table 3.3.3-1 shall be OPERABLE.
APPLICABILITY: As specified in Table 3.3.3-1. ACTIONS
-------------------------------------NOTE---~---------------------------------
Separate Condition entry is allowed for each Function. CONDITION REQUIRED ACT ION COMPLETION TIME A. --------NOTE---------- Not applicable to RAS *
One or more Functions with one ESF bistable or associated instrument channel inoperable.
A.I Place affected bistable in trip.* 7 days B. ---------NOTE-------~ ------------NOTE----------- Not applicable to RAS. LCO 3.0.4 is not applicable. One or more Functions B.1
Place one bistable 8 hours w-ith two ESF bistables in trip.
or associated instrument
-channels inoperable. AND B.2 Restore one 7 days bistable and associated instrument to OPERABLE status. \
P~lisades Nuclear Plant 3.3.3-1 Amendment No. 05/30/99
ESF Instrumentation
ACTIONS CONDITION REQUIRED ACTION 3.3.3 COMPLETION TIME
c. One RAS bistable or c.1 Bypass affected 8 hours associated instrument bistable.
channel inoperable. AND C.2 Restore bistable 7 days and associated instrument channel to OPERABLE status. D. Required Action and D.l Be in MODE 3. 6 hours associated Completion Time not met for AND Functions 1. 2 3, 4, 1 or 7. D.2 Be in MODE 4. 30 hours
E. Required Action and associated Completion Time not met for Functions 5 or 6.
E.1 AND Be in MODE 3. 6 hours E.2 Be in MODE 5. 36 hours
Palisades Nuclear Plant 3.3.3-2 Amendment No. 05/30/99
ESF Instrumentation
SURVEILLANCE REQUIREMENTS 3.3.3
-------------------------------------NOTE--------------~----------------------
Refer to Table.3.3.3-1 to determine which SR shall be performed for each Function. SURVEILLANCE FREQUENCY SR 3.3.3.1 Perform a CHANNEL CHECK. 12 hours SR 3.3.3.2 Perform a CHANNEL FUNCTIONAL TEST. 92 days SR 3.3.3.3 Perform a CHANNEL CALIBRATION . 18 months Palisades Nuclear Plant 3.3.3-3 Amendment No. 05/30/99
1 ESF Instrumentation -* Table 3.3.3-1 (page 1 of 2) Engineered Safety Features Instrumentation 3.3.3 APPLICABLE SURVEILLANCE ALLOWABLE FUNCTION MODES REQUIREMENTS VALUE
1. Safety Injection Signal (SIS)

a. Pressurizer Low Pressure 1,2,3 SR 3.3.3.1 ::<: 1593 psia SR 3.3.3.2 SR 3.3.3.3

2. Steam Generator Low Pressure Signal (SGLP)

a. Steam Generator A Low SR 3.3.3.1 ::<: 500 psia Pressure SR 3.3.3.2 SR 3.3.3.3

b. Steam 'Generator B Low SR 3.3.3.1 ::<: 500 psia Pressure SR 3.3.3.2 SR 3.3.3.3
3. Recirculation Actuation Signal (RAS)
a. SIRWT Low Level 1,2,3 SR 3.3.3.3 ::<: 21 inches and
~ 27 inches above tank-.
bottom
4. Auxiliary Feedwater Actuation Signal (AFAS)

a. Steam Generator A Low 1,2,3 SR 3.3.3.1 ::<: 25. 9%
Level SR 3.3.3.2 narrow range SR 3.3.3.3
b. Steam Generator B Low 1,2,3 SR 3.3.3.1 ::<: 25.9%
Level SR 3.3.3.2 narrow range SR 3.3.3.3 (a) Not required to be OPERABLE when a1'1 Main Steam Isolation Valves (MSIVs) are closed and deactivated, and all Main Feedwater Regulating Valves (MFRVs) and MFRV bypass valves are either closed and deactivated, or isolated by closed manual valves. Palisades Nuclear Plant 3.3.3-4 Amendment No. 05/30/99
ESF Instrumentation 3.3.3
FUNCTION Table 3.3.3-1 (page 2 of 2)
Engineered Safety Features Instrumentation APPLICABLE SURVEILLANCE MODES REQUIREMENTS ALLOWABLE VALUE
5. Containment High Pressure (CHP)

a. Containment High Pressure 1,2,3,4 SR 3.3.3.2 z 3.7 psig
- Left Train SR 3.3.3.3 and ~ 4.3 psig
b. Containment High Pressure 1,2,3,4 SR 3.3.3.2 z 3.7 psig
- Right Train SR 3.3.3.3 and ~ 4.3 psig
6. Containment High Radiation Signal (CHR)

a. Containment High Radiation 1,2,3,4 SR 3.3.3.l ~ 20 R/hour SR 3.3.3.2 SR 3.3.3.3
7

Automatic Bypass Removals
a. Pressurizer Low Pressure Bypass 1,2,3 1,2(a) ,3(a)
SR 3.3.3.3 ~ 1700 psi a
b. Steam Generator A Low SR 3.3.3.3 ~ 565 psi a Pressure Bypass

c. Steam Generator B Low l,2(a) ,3(a) SR 3.3.3.3 ~ 565 psi a Pressure Bypass (a) Not required to be OPERABLE when all Main Steam Isolation Valves (MSIVs) are closed and deactivated, and all Main Feedwater Regulating Valves (MFRVs) and MFRV bypass valves are either closed and deactivated, or isolated by closed manual valves .
Palisades Nuclear Plant 3.3.3-5 Amendment No. 05/30/99

ESF.Logic and Manual Initiation 3.3.4 3.3 INSTRUMENTATION 3.3.4 Engineered Safety Features (ESF) Logic and Manual Initiation LCO 3.3.4 Two ESF Manual Initiation and two ESF Actuation Logic channels and associated bypass removal channels shall be OPERABLE for each ESF Function specified in Table 3.3.4-1.
APPLICABILITY: According to Table 3.3.4-1. ACTIONS
-------------------------------------NOTE-------------------------------------
Separate Condition entry is allowed for each Function .
A.
CONDITION One or more Functions A.1 REQUIRED ACTION Restore channel to COMPLETION TIME 48 hours with one Manual OPERABLE status. Initiation, Bypass Removal, or Actuation Logic channel inoperable .
Palisades Nuclear Plant 3.3.4-1 Amendment No. 05/30/99

ACTIONS CONDITION ESF Logic and Manual Initiation REQUIRED ACT ION 3.3.4 COMPLETION TIME B. One or more Functions B.1 Be in MODE 3. 6 hours with two Manual Initiation, Bypass AND Removal, or Actuation Logic channels B.2 Be in MODE 4. 30 hours inoperable for Functions 1, 2, 3, or 4.
QR I Required Action and associated Completion Time of Condition A not met for Functions 1; 2, 3, or 4*
c. One or more Functions with two Manual Initiation, or Actuation Logic channels inoperable for Functions 5 or 6.
c.1 AND C.2 Be in MODE 3. Be in MODE 5. 6 hours 36 hours OR Required Action and associated Completion Time of Condition A not met for Functions 5 or 6.
Palisades Nuclear Plant 3.3.4-2 Amendment No. 05/30/99

ESF Logic and Manual Initiation 3.3.4" SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.3.4.1 Perfonn functional test of each SIS 92 days actuation channel nonnal and standby power functions.
SR 3.3.4.2 Perfonn a CHANNEL FUNCTIONAL TEST of each 92 days AFAS actuation logic channel. SR 3.3.4.3 Perfonn a CHANNEL FUNCTIONAL TEST . 18 months
Palisades Nuclear Plant 3.3.4-3 Amendment No. 05/30/99

ESF Logic and Manual Initiation 3.3.4 Table 3.3.4-1 (page 1 of 1)
Engineered Safety Features Actuation Logic and Manual Initiation APPLICABLE FUNCTION MODES
1. Safety Injection Signal {SIS)C*l 1,2,3

2. Steam Generator Low Pressure Signal (SGLP) CbJCcJ

3. Recirculation Actuation Signal (RAS) 1,2,3

4. Auxiliary Feedwater Actuation Signal 1,2,3 (AFAS)

5. Containment High Pressure Signal 1,2,3,4 (CHP)CeJ

6. Containment H~gh Radiation Signal (CHR) 1,2,3,4 (a) SIS actuation by Pressurizer Low Pressure may be manually bypassed when pressurizer pressure is ~ 1700 psia. The bypass shall be automatically removed whenever pressurizer pressure is > 1700 psia.
(b) SGLP actuation may be manually bypassed when SG pressure is ~ 565 psia. The bypass shall be automatically removed whenever steam generator pressure is
> 565 psia.
(c) Manual Initiation may be achieved by individual component controls. (d) Not required to be OPERABLE when all Main Steam Isolation Valves (MSIVs) are closed and deactivated, and all Main Feedwater Regulating Valves (MFRVs) and MFRV bypass valves are either closed and deactivated, or isolated by closed manual valves. (e) Manual Initiation channels not required .
Palisades Nuclear Plant 3.3.4-4 Amendment No. 05/30/99

DG - UV Start 3.3.5 3.3 INSTRUMENTATION 3.3.5 Diesel Generator (DG) - Undervoltage Start (UV Start)
LCO 3.3.5 Three channels of Loss of Voltage Function and three channels of Degraded Voltage Function auto-initiation instrumentation and associated logic channels for each DG shall be OPERABLE. APPLICABILITY: When associated DG is required to be OPERABLE. ACTIONS CONDIT.ION REQUIRED ACT ION COMPLETION TIME A. One or more Functi ans A.1 Enter applicable Immediately with one channel per
Conditi ans and DG inoperable

Requ1red Actions for the associated DG made inoperable by DG - UV Start instrumentation.
SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.3.5.1 Perfonn a CHANNEL FUNCTIONAL TEST on each 18 months DG-UV start logic channel .
Palisades Nuclear Plant 3.3.5-1 Amendment No. 05/30/99

SURVEILLANCE REQUIREMENTS DG - UV Start 3.3.5 SURVEILLANCE FREQUENCY SR 3.3.5.2 Perfonn CHANNEL CALIBRATION on each Loss of 18 months Voltage and Degraded Voltage channel with setpoint~ as follows:
a. Degraded Voltage Function ~ 2187 V and
2264 v Time delay

~ 0.5 seconds and

0.8 seconds; and

b. Loss of Voltage Function ~ 1780 v and
1940 v Time delay

~ 5.45 seconds and

8.15 seconds at 1400 V.
Palisades Nuclear Plant 3.3.5-2 Amendment No. 05/30/99

Refueling CHR Instrumentation 3.3.6 3.3 INSTRUMENTATION 3.3.6 Refueling Containment High Radiation (CHR) Instrumentation LCO 3.3.6 Two Refueling CHR Automatic Actuation Function channels and two CHR Manual Actuation Function channels shall be OPERABLE.
APPLICABILITY: During CORE ALTERATIONS, During movement of irradiated fuel assemblies within containment. ACTIONS CONDITION REQUIRED ACTION COMP LET ION TIME A. One or more Functions A.1 Place the affected 4 hours with one channel channel in trip. inoperable . OR A.2.1 Suspend CORE 4 hours AL TERA TI ONS . AND A.2.2 Suspend movement of 4 hours irradiated fuel assemblies within containment. B. One or more Functions 8.1 Suspend CORE Immediately with two channels AL TERATI ONS . inoperable. AND 8.2 Suspend movement of Immediately irradiated fuel assemblies within containment.
\
Palisades Nuclear Plant 3.3.6-1 Amendment No. 05/30/99

SURVEILLANCE REQUIREMENTS Refueling CHR Instrumentation 3.3.6 SURVEILLANCE FREQUENCY SR 3.3.6.1 Perfonn a CHANNEL CHECK of each refueling 12 hours CHR monitor channel.
SR 3.3.~.2 Perfonn a CHANNEL FUNCTIONAL TEST of each 31 days refueling CHR monitor channel. SR 3.3.6.3 Perfonn a CHANNEL FUNCTIONAL TEST of each 18 months CHR Manual Initiation channel. SR 3.3.6.4 Perfonn a CHANNEL CALIBRATION of each 18 months refueling CHR monitor channel *
Palisades Nuclear Plant 3.3.6-2 Amendment No. 05/30/99

3.3 INSTRUMENTATION PAM Instrumentation 3.3.7 3.3.7 Post Accident Monitoring (PAM) Instrumentation LCO 3.3.7 The PAM instrumentation for each Function in Table 3.3.7-1 shall be OPERABLE.
APPLICABILITY: MODES 1, 2, and 3. ACTIONS
-------------------------------------NOTES------------------------------------
1. LCO 3.0.4 is not applicable.
\
2. Separate Condition entry is allowed for each Function.
CONDITION REQUIRED ACTION COMPLETION TIME A. One or more Functions A.1 Restore required 30 days with one required channel to OPERABLE channel inoperable. status.
8. Required Action and 8.1 Initiate action in Immediately associated Completion accordance with Time of Condition A Specification 5.6.6.
not met.
c. ---------NOTE---------
Not applicable to hydrogen monitor channels. One or more Functions c.1 Restore one channel 7 days with two required to OPERABLE status. channels inoperable. \
Palisades Nuclear Plant 3.3.7-1 Amendment No. 05/30/99

ACTIONS PAM Instrumentation 3.3.7 CONDITION REQUIRED ACT ION COMP LET ION TIME D. Two hydrogen monitor D.1 Restore one hydrogen 72 hours channels inoperable. monitor channel to OPERABLE status.
E. Required Action and E.1 Enter the Condition I.mmedi ately associated Completion referenced in Time of Condition C Table 3.3.7-1 for the or D not met. channel. F. As required by F.1 Be in MODE 3. 6 hours Required Action E.1
and referenced in Table 3 . 3 *7-1.
AND F.2 Be in MODE 4. 30 hours G. As required by G.l Initiate action in Immediately Required Action E.1 accordance with and referenced in Specification 5.6.6. Tab l e 3 *3 . 7-1.
Palisades Nuclear Plant 3.3.7-2 Amendment No. 05/30/99

SURVEILLANCE REQUIREMENTS PAM Instrumentation 3.3.7
-------------------------------------NOTE-------------------------------------
These SRs apply to each PAM instrumentation Function in Table 3.3.7-1. SURVEILLANCE FREQUENCY SR 3.3.7.1 Perfonn CHANNEL CHECK. 31 days SR 3.3.7.2 ---------~---------NOTE-------------------- Neutron detectors are excluded from the CHANNEL CALIBRATION. Perfonn CHANNEL CALIBRATION. 18 months
Palisades Nuclear Plant 3.3.7-3 Amendment No. 05/30/99

Table 3.3.7-1 (page 1 of 1)
PAM Instrumentation 3.3.7 Post Accident Monitoring Instrumentation CONDITIONS REQUIRED REFERENCED FROM FUNCTION CHANNELS REQUIRED ACTION E.1
1. Primary Coolant System Hot Leg Temperature (wide range) 2 F

2. Primary Coolant System Cold Leg 2 F Temperature (wide range)

3. Wide Range Neutron Flux 2 F

4. Containment Floor Water Level F (wide range)

5. Subcooled Margin Monitor 2 F

6. Pressurizer Level (wide range) 2 F

7. Containment Hydrogen Monitors 2 F

8. Condensate Storage Tank Level 2 F

9. Primary Coolant System Pressure 2 F (wide range)

10. Containment Pressure (wide range) 2 F

11. Steam Generator A Water Level 2 F (wide range)

12. Steam Generator 8 Water Level 2 F (wide range)

13. Steam Generator A Pressure 2 F

14. Steam Generator B Pressure 2 F

15. Containment Isolation Valve 1 per valveCaJ F Position

16. Core Exit Temperature - Quadrant 1 4 F

17. Core Exit Temperature - Quadrant 2 4 F

18. Core Exit Temperature - Quadrant 3 4 F

19. Core Exit Temperature - Quadrant 4 4 F

20. Reactor Vessel Water Level 2 G

21. Containment Area Radiation 2 G (high range)
(a) Not required for isolation valves whose associated penetration is isolated by at least one closed and de-activated automatic valve, closed manual valve, blind flange, or check valve with flow through the valve secured. Palisades Nuclear Plant 3.3.7-4 Amendment No. 05/30/99
3.3 INSTRUMENTATION Alternate Shutdown System 3.3.8 3.3.8 Alternate Shutdown System LCO 3.3.8 The Alternate Shutdown System Functions in Table 3.3.8-1 shall be OPERABLE.
APPLICABILITY: MODES 1, 2, and 3. ACTIONS
-------------------------------------NOTES------------------------------------
1. LCO 3.0.4 is not applicable.

2. Separate Condition entry is allowed for each Function.
A.
CONDITION One or more required Functions inoperable. A.1 REQUIRED ACT ION Restore required Functions to OPERABLE status. COMPLETION TIME 30 days B. Required Action and B.1 Be in MODE 3. 6 hours associated Completion Time not met. AND 8.2 Be in MODE 4 . 30 hours
Palisades Nuclear Plant 3.3.8-1 Amendment No. 05/30/99

SURVEILLANCE REQUIREMENTS SURVEILLANCE Alternate Shutdown System FREQUENCY 3.3.8 SR 3.3.8.1 Perfonn CHANNEL FUNCTIONAL TEST of the Once within Source Range Neutron Flux Function. 7 days prior to each reactor startup SR 3.3.8.2 Verify each required control circuit and 18 months transfer switch is capable of perfonning the intended function.
SR 3.3.8.3 -------------------NOTES-------------------
1. Not required for Functions 16, 17, and 18.

2. Neutron detectors are excluded from
the CHANNEL CALIBRATION .
Perfonn CHANNEL CALIBRATION for each required instrumentation channel
18 months

Palisades Nuclear Plant 3.3.8-2 Amendment No. 05/30/99

Table 3.3.8-1 (page 1 of 1)
Alternate Shutdown System 3.3.8 Alternate Shutdown System Instrumentation and Controls FUNCTION REQUIRED INSTRUMENT OR CONTROL PARAMETER CHANNELS
1. Source Range Neutron Flux 1

2. Pressurizer Pressure 1 I

3. Pressurizer.Level 1 I

4. Primary Coolant System (PCS) #1 Hot Leg Temperature 1

5. PCS #2 Hot Leg Temperature 1

6. PCS #1 Cold Leg Temperature 1

7. PCS #2 Cold Leg Temperature 1

8. Steam Generator (SG) A Pressure 1

9. SG 8 Pressure 1
10. SG A Wide. Range Level 1
11. SG 8 Wide Range Leve 1 1

12. Safety Injection Refueling Water (SIRW)* Tank Level 1

13. Auxi l'i ary Feedwater (AFW) Fl ow In di ca ti on to SG A 1

14. AFW Flow Indication to SG 8 1

15. AFW Low Suction Pressure Alarm (P-88) 1

16. AFW_ Pump P-88 Steam Supply Valve Control 1

17. AFW Fl ow Contra 1 to SG A 1

18. AFW Flow Control to SG 8 1
Palisades Nuclear Plant 3.3.8-3 Amendment No. 05/30/99
Neutron Flux Monitorin~
Channels 3.3.9 3.3 INSTRUMENTATION 3.3.9 Neutron Flux Monitoring Channels LCO 3.3.9 Two channels of neutron flux monitoring instrumentation shall be OPERABLE. APPLICABILITY: MODES 3 4 and 5. 1 1 ACTIONS CONDITION REQUIRED ACT ION COMPLETION TIME
Palisades Nuclear Plant 3.3.9-1 Amendment No. 05/30/99

Neutron Flux Monitoring Channels 3.3.9 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.3.9.l Perfonn CHANNEL CHECK. 12 hours SR 3*.3.9.2 -------------------NOTE--------------------
Neutron detectors are excluded from the CHANNEL CALIBRATION. Perfonn CHANNEL .CALIBRATION. 18 months
Palisades Nuclear Plant 3.3.9-2 Amendment No. 05/30/99
-*- ESRV Instrumentation-3.3.10 3.3 INSTRUMENTATION 3.3.10 Engineered Safeguards Room Ventilation (ESRV) Instrumentation LCO 3.3.10 Two channels of ESRV Instrumentation shall be OPERABLE. APPLICABILITY: MODES 1, 2, 3, and 4. ACTIONS
-------------------------------------NOTE-------------------------------------
Separate Condition entry is allowed for each channel. CONDITION REQUIRED ACT ION COMPLETION TIME A. One or more channels. A.1 Initiate action to Immediately inoperable. isolate the associated ESRV System. SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.3.10.1 Perfonn a CHANNEL CHECK. 12 hours SR 3.3.10.2 Perform a CHANNEL FUNCTIONAL-TEST. 31 days SR 3.3.10.3 Perform a CHANNEL CALIBRATION. 18 months Verify high radiation s~tpoint on each ESRV Instrumentation radiation monitoring channel is ~ 2.2E+5 cpm . Palisades Nuclear Plant 3 .3 .10-1 Amendment No. 05/30/99
RPS Instrumentation B 3.3.1
B 3_. 3 INSTRUMENTATION B 3.3.1 Reactor Protective System (RPS) Instrumentation BASES BACKGROUND The RPS initiates a reactor trip to protect against violating the acceptable fuel design limits and breaching the reactor coolant pressure boundary during Anticipated Operational Occurrences (AOOs). By tripping the reactor, the RPS also assists the Engineered Safety Features (ESF) systems in mitigating accidents.
The protection and monitoring systems have been designed to ensure safe operation of the reactor. This is achieved by specifying Limiting Safety System Settings (LSSS) in terms of parameters directly monitored by the RPS, as well as LCOs on other reactor system parameters and equipment performance. The LSSS, defined in this Specification as the Allowable_ Values, in conjunction with the LCOs, establish the threshold for protective system action to prevent exceeding acceptable Jimits during Design Basis Accidents (DBAs). During AOOs, which are those events expected to occur one or more times during the plant life, the acceptable limits are:
The Departure from Nucleate Boiling Ratio (DNBR) shall be maintained above the Safety Limit (SL) value to prevent departure from nucleate boiling;

Fuel centerlin~ melting shall not occu~; and

The Primary Coolant System (PCS) pressure SL of 2750 psia shall not be exceeded.
Maintaining the parameters within the above values ensures that the offsite dose will be within the 10 CFR 50 (Ref. 1) and 10 CFR 100 (Ref. 2) c~iteria during AOOs. Accidents are events that are analyzed even though they are not expected to occur during the plant life. The acceptable limit during accidents is that the offsite dose shall be maintained within an acceptable fraction of 10 CFR 100 (Ref. 2) limits. Different accident categories allow a different fraction of these limits based on probability of occurrence. Meeting the acceptable dose limit for an
accident category is considered having acceptable consequences fo~ that event.
Palisades Nuclear Plant B 3.3.1-1 05/30/99
RPS Instrumentation B 3.3.1
_BA_S_E_S~~~~~~~~~~~~~~~~~~~~~~~---~~~--=-'-"
BACKGROUND The RPS is segmented into four interconnected modules. (continued) ~hese modules are:
Measurement channels;

RPS trip units; I

Matrix Logic; and

Trip Initiation Logic.
This LCO addresses measurement channels and RPS trip units. It also addresses the automatic bypass removal feature for those trips with Zero Power Mode bypasses. The RPS Logic and Trip Initiation Logic are addressed in LCO .3.3.2, "Reactor Protective System (RPS) Logic and Trip Initiation." The role of the measurement channels, RPS trip units, and RPS Bypasses is discussed below. - Measuremerit Channels. Measurement channels, consisting of pressure switches, field
transmitters, or process sensors and associated instrumentation, provide a measurable electronic s{gnal based upon the phy$ical characteristics of the parameter being measured.

With the.exception of Hi Startup Rate, which employs two instrument channels, and Loss of Load, which employs a single pressure sensor, four identical measurement channels with ~lectrical and physical separation are provided for each parameter used in the direct generation of trip signals. These are designated channels A through D. Some measurement channels provide input to more than one RPS trip unit within the same RPS channel. In addition, some measurement channels may also be used as inputs to Engineered.Safety Features (ESF) bistables, and most provide indication in the control room.

In the case of Hi Startup Rate and Loss of Load, where fewer than four sensor channels are employed, the reactor trips provided are not relied upon by the plant safety analyses.
The sensor channels do however, provide trip input signals to all four RPS channels. Palisades Nuclear Plant B 3.3.1-2 05/30/99
RPS Instrumentation B 3.3.1
_BA_S_E_S~~~~~~~~~~~~--------"'--~--'------"---'-----'-~~----'-;.___,;;,.._;,,;.;_
I BACKGROUND Measurement Channels (continued) When a channel monitoring a parameter exceeds a predetermined setpoint, indicating an abnormal condition, the bistable monitoring the parameter in that channel will trip. Tripping two or more channels of bistable trip units monitoring the same parameter de-energizes Matrix Logic, (addressed by LCO 3.3.2) which in turn de-energizes the Trip Initiation Logic. This causes all four DC clutch power supplies to de-energize, interrupting power to the control
~ rod drive mechanism clutches, allowing the full length*
control rods to insert into the core. For those trips relied upon in the safety analyses, three of the four measurement and trip unit channels are necessary to meet the redundancy and testability of GDC 21 in 10 CFR 50, Appendix A (Ref. 1). The fourth channel provides additional flexibility by allowing one channel to be removed from
~ervice (trip channel bypass) for maintenance or testing while still maintaining a minimum two-out-of-three logic.
Since no single failure will prevent a protective system actuation, this arrangement meets the requirements of IEEE
Standard 279-1971 (Ref. 3) .
Most of the RPS trips are generated by comparing a single measurement to a fixed bistable setpoint. Two trip Functions, Variable High Power Trip and-Thermal Margin Low Pressure Tripi make.use of more than one measurement to provide a trip .. The required RPS Trip Functions utilize the following input instrumentation:
Variable High Power Trip (VHPT)
The VHPT uses Q Power as its input. Q Power is the higher of NI power from the power range NI drawer and primary calorimetric power (~T power) based on PCS hot leg and cold leg temperatures. The measurement channels associated with the VHPT are the power range excore channels, and the PCS hot and cold leg temperature channels .
Palisades Nuclear Plant B 3.3.1-3 05/30/99
RPS Instrumentation B 3.3.1
_BA_S_E_S~...--~~~~~~~~~~~~~~~~~~~~~~~~~~
BAC KGRO UN D Measurement Channels
Variable High Power Trip (VHPT) (continued)
The Thermal Margin Monitors provide the complex signal processing necessary to calculate the TM/LP trip setpoint, VHPT trip setpoint and trip comparison, and Q Power calculation. On power decreases the VHPT setpoint tracks power levels downward so that it is always within a fixed increment above current power, subject to a minimum value. On power increases, the trip setpoint remains fixed unless manually reset, at which point it increases to the new setpoint, a fixed increment above Q Power at the time of reset, subject to a maximum value. Thus, during power escalation, the trip setpoint must be repeatedly reset to avoid a reactor trip.
High Startup Rate Trip .
The High Startup Rate trip uses the wide range Nuciear Instruments (Nis) to provide an input signal. There are only two wide range NI channels. The wide ,range channel signal processing electronics are physically mounted in RPS cabinet channels C (NI-1/3) and D (NI-2/4). Separate bistable trip units mounted within the NI-1/3 wide range channel drawer supply High Startup Rate trip signals to RPS channels A and C. Separate bistable trip units mounted within the NI-2/4 wide range channel drawer provide High Startup Rate trip signals to RPS channels B and D.
Low Primary Coolant Flow Trip The Low Primary Coolant Flow Trip utilizes 16 flow measurement channels which monitor the differential pressure across the primary side of the steam generators. Each RPS channel, A, B, C, and D, receives a signal which is the sum of four differential pressure signals. This totalized signal is compared with a setpoint in the RPS Low Flow bistable trip unit for that RPS channel *

Palisades Nuclear Plant B 3.3.1-4 05/30/99
RPS Instrumentation B 3.3.1
_BA_S_E_S~~~~~~~~~~~~~~~~~~~~~~~~~~~
BACKGROUND Measurement Channels (continued)
Low Steam Generator Level Trips There are two separate Low Steam Generator Level trips, one for each steam generator. Each Low Steam Generator Pressure trip monitors four level measurement channels for the associated steam generator, one for each RPS channel.

Low Steam Generator Pressure Trips There are also two separate Low Steam Generator Pressure trips, one for each steam generator. Each Low Steam Generator Pressure trip monitors four pressure measurement channels for the associated steam generator, one for each RPS channel.

High Pressurizer Pressur~ Trip The High Pressurizer Pressure Trip monitors four pressurizer pressure channels, one for each RPS channel.

Thermal Margin Low Pressure (TM/LP) Trip I
The TM/LP Trip utilizes bistable trip units. Each of these bistable trip units receives a calculated trip setpoint from the Thermal Margin Monitor (TMM) and compares it to the measured pressurizer pressure signal. The TM/LP setpoint is based on Q power (the higher of NI power from the power range NI drawer, or
~T power, based on PCS hot leg and cold leg temperatures) pressurizer pressure, PCS cold leg temperature, and Axial Shape Index. The TMM provide the complex signal processing necessary to calculate the TM/LP trip setpoint, TM/LP trip comparison signal, and Q Power .
Palisades Nuclear Plant B 3.3.1-5 05/30/99
RPS Instrumentation B 3.3.1
_BA_S_E~S~~"'--~~"--~-'-~-'--~~~'--~~...-~~~~~~~~~~~
BA CK GROUND Measurement Channels (continued)
Loss of Load Trip The Loss of Load Trip is actuated by turbine auxiliary relays 305L and 305R. Relay 305L provides input to RPS channels A and C; 305R to channels B and D.
Relays 305L and 305R are energized on a turbine trip. Their inputs are the same as the inputs to the turbine solenoid trip valve, 20ET. If a turbine trip is generated by loss of auto stop oil pressure, auto stop oil pressure switch 63/AST-2 will actuate relays 305L and 305R and generate a reactor trip . . If a turbine trip is generated by an input to the solenoid trip valve, relays 305L and 305R, which are wired in parallel, will also be actuated and will generate a reactor trip~
Containment High Pressure Trip The Containment High Pressure Trip i~ actuated by four pressure switches, one for each RPS channel.

Zero Power Mode Bypass Automatic Removal The Zero Power Bypass allows manually bypassing (i.e. disabling) four reactor trip functions, Low PCS Flow, Low SG A Pressure, Low SG B Pressure, and TM/LP (low PCS pressure), when reactor power (as indicated by the wide range nuclear instrument channels) is below 10" 4%. This bypassing is necessary to allow RPS testing and control rod drive mechanism testing when the reactor is shutdown and plant conditions would

cause a reactor trip to be present.
The Zero Power Mode Bypass removal interlock uses the wide range nuclear instruments (Nis) as measurement channels. There are only two wide range NI channels. Separate bistables are provided to actuate the bypass 0 removal .for each RPS channel. Bistables in the NI-1/3 channel provide the bypass removal function for RPS channels A and C; bistables in the NI~2/4 channel for RPS channels B and D*
Palisades Nuclear Plant B 3.3.1-6 05/30/99
RPS Instrumentation B 3.3.1
_BA_S__E_S~~~~~~~~~~~~~~~~~~~~~~~~~~~~
BACKGROUND Measurement Channels (continued) Several measurement instrument channels provide more than one required function. Those sensors shared for RPS and ESF functions are identified in Table B 3.3.1-1. That table provides a listing of those shared channels and the Specifications which they affect. RPS Trip Units Two types of RPS trip units are used iri the RPS cabinets; bistable trip units and auxiliary trip units: A bistable trip unit receives a measured process signal from its instrument channel and compares it to a setpoint; the trip unit actuates three relays, with contacts in the Matrix Logic channels, wh~n the measured signal is less conservative than. the setpoint~ They also.provide local trip indication and remote annunciation. An auxiliary trip unit receives a digital input (contacts open or closed);the trip unit actuates three relays, with contacts in the Matrix Logic channels, when the digital input is received. They also provide local trip'indication and remote annunciation. Each. RPS channe 1 has four auxiliary trip uni ts and seven bistable trip units~ The contacts from these trip unit relays are arranged into six coincidence matrices, ~omprising the Matrix Logic. If bistable trip units monitoring the same parameter in at least two channels trip, the Matrix Logic will generate a* reactor trip (two-out-of-four logic). Four of the RPS measurement channels provide contact outputs to the RPS, so the comparison of an analo~ input to a trip
setpoint is not necessary. In these cases, the bistable trip unit is replaced with an auxiliary trip unit. The auxiliary trip uni ts provide contact multi pl i ca ti on so the single input contact opening can provide multiple contact outputs to the coincidence logic as well as trip indication and annunciation .

Palisades Nuclear Plant B 3.3.1-7 05/30/99
RPS Instrumentation B 3.3.1
_BA_S_E_S~~~~~~~~~~~~~~~~~~_;.._~~~~~~~~~
BACKG ROUND RPS Trip Units (continued) Trips employing auxiliary trip units include the VHPT, which receives contact inputs from the Thermal Margin Monitors; the High Startup Rate trip which employs contact inputs from bistables mounted in the two wide range drawers; the Loss of Load Trip which receives contact inputs from one of two auxiliary relays which are operated by a single switch sensing turbine auto stop oil pressure; and the Containment High Pressure (CHP) trip, which employs containment pressure switch contacts. There are four RPS trip units, designated as channels A through D, each channel having eleven trip units, one for each RPS Function. Trip unit output relays de-energize when a trip occurs. All RPS Trip Functions, with the exception of the Loss of Load and CHP trips, generate a pretrip alarm as the trip setpoint is approached. The Allowable Values are specified for each safety related RPS trip Function which is credited in the safety analysis. Nominal trip setpoints are specified in the plant
procedures. The nominal setpoints are selected to ensure plant parameters do not exceed the Allowable Value if the instrument loop is performing as required. The methodology used to determine the nominal trip setpoints is also provided in plant documents. Operation with a trip setpoint less conservative than the nominal trip setpoint, but within its Allowable Value, is acceptable. Each Allowable Value specified is more conservative than the analytical limit determined in the safety analysis in order to account for uncertainties appropriate to the trip Function. These uncertainties are addressed as described in pl~nt documents.
A channel is inoperable if its actual setpoint is not within its Allowable Value. Setpoints in accordance with the Allowable Value will ensure that SLs of Chapter 2.0 are not violated during AOOs and the consequences of DBAs will be acceptable, providing the plant is operated from within the LCOs at the onset of the AOO or OBA and the equipment functions as designed. Note that in the accompanying LCO 3.3.1, the Allowable Values of Table 3.3.1-1 are the LSSS . Palisades Nuclear Plant B 3.3.1-8 05/30/99
~--- RPS Instrumentation B 3.3.1
_BA_S_E_S~~~~~~~~~~~~~~~~~~~~~~~~~~~~
BACKG ROUND Reactor Protective System Bypasses (continued) Three different types of trip bypass are utilized in the RPS, Operating Bypass, Zero Power Mode Bypass, and Trip Channel Bypass. The Operating Bypass or Zero Power Mode Bypass prevent the actuation of a trip unit or auxiliary trip unit; the Trip Channel Bypass prevents the trip unit output from affecting the Logic Matrix. A channel which is bypassed, other than as allowed by the Table 3.3.1-1 footnotes, cannot perform its specified safety function and must be considered to be inoperable. Operating Bypasses The Operating Bypasses are initiated and removed automatically during startup and shutdown as power level changes. An Operating Bypass prevents the associated RPS auxiliary trip unit from receiving a trip signal from the associated measurement channel. With the bypass in place, neither the pre-trip alarm nor the trip will actuate if the measured parameter exceeds the set point. An annunciator is provided for each Operating Bypass. The RPS trips with Operating Bypasses are:
a. High Startup Rate Trip bypass. The High Startup Rate trip is automatically bypassed when the associated wide range channel indicates below lE-4% RTP, and when the associated power range excore channel indicates above 13% RTP. These bypasses are automatically removed between lE-4% RTP and 13% RTP.
b. Loss of Load bypass. The Loss of Load trip is automatically bypassed when the associated power range excore channel indicates below 17% RTP. The bypass is automatically removed when the channel indicates above the set point. The same power range excore channel bistable is used to bypass the High Startup Rate trip and the Loss of Load trip for that RPS channel.
Each wide range channel contains two bistables set at lE-4% RTP, one bistable unit for each associated RPS channel. Each of the two wide range channel affects the Operating Bypasses for two RPS channels; wide range channel NI-1/3 for RPS channels A and C, wide range channel NI-2/4 for RPS channels B and D. Each of the four power range excore channel affects the Operating Bypasses for the associated RPS channel.. The power range excore channel bistables associated with the Operating Bypasses are set at a nominal 15%, and are.required to actuate between 13% RTP and 17% RTP. . Palisades Nuclear Plant B 3.3.1-9 05/30/99
RPS Instrumentation B 3.3.1
_BA_S_E_S~~~~~~~~~~~~~~~~'--'-~~~~~~~~~~
BACKGROUND Zero Power Mode (ZPM) Bypass (continued) The ZPM Bypass is used when the plant is shut down and it is desired to raise the control rods for control rod drop testing with PCS flow, pressure or temperature too low for the RPS trips to be reset. ZPM bypasses may be manually initiated and removed when wide range power is below lE-4% RTP, and are automatically removed if the associated wide range NI indicated power exceeds lE-4% RTP. A ZPM bypass prevents the RPS trip unit from actuating if the measured parameter exceeds the set point. Operation of the pretrip alarm is unaffected by the zero power mode bypass. An annunciator indicates the presence of any ZPM bypass. The RPS trips with ZPM bypasses are:
a. Low Primary Coolant System Flow.

b. Low Steam Generator Pressure.

c. Thermal Margin/Low Pressure.
The wide range NI channels provide contact closure permissive signals when indicated power is below lE-4% RTP . The ZPM bypasses may then be manually initiated or removed by actuation of key-lock switches. One key-lock switch located on each RPS cabinet co~trols the ZPM Bypass for the associated RPS trip channels. The bypass is automatically removed if the associated wide range NI indicated power exceeds lE-4% RTP. The same wide range NI channel bistables that provide the ZPM Bypass permissive and removal signals also provide the high startup rate trip Operating Bypass actuation and removal; Trip Channel Bypass A Trip Channel Bypass is used when it is desired to physically remove an individual trip unit from the system, or when calibration or servicing of a trip channel could cause an inadvertent trip. A trip Channel Bypass may be manually initiated or removed at any time by actuation of a key-lock switch. A Trip Channel Bypass prevents the trip unit output from effecting the RPS logic matrix. A light above the bypass switch indicates that the trip channel has been bypassed. Each RPS trip unit has an associated trip channel bypass:
Palisades Nuclear Plant B 3.3.1-10 05/30/99
RPS Instrumentation B 3.3.1
_BA_S_E_S~~~~~~~~~~~~~~~~~~~~~~~~~~~~
BACKGROUND Trip Channel Bypass {continued) The key-lock trip channel bypass switch is located above each trip unit. The key cannot be removed when in the bypass position. Only one key for each trip parameter is provided, therefore the operator can bypass only one channel of a given parameter at a time. During the bypass
condition, system logic changes from two-out-of~four to two-out-of-three channels required for trip.
APPLICABLE Each of the analyzed ~ccidents and transients can be SAFETY ANALYSES detected by one or more RPS Functions. The accident analysis contained in Reference 4 takes credit for most RPS trip Functions. The High Startup Rate and Loss of Load Functions, which are not specifically credited in the accident analysis are part of the NRC approved licensing basis for the plant. The High Startup Rate and Loss of Load trips are purely equipment protective, and their use minimizes the potential for equipment damage. The specific safety analyses applicable to .each protective Function are identified below.
1. Variable High Power Trip (VHPT)
The VHPT provides reactor core protection against positive reactivity excursions.* The safety analysis assumes that this trip is OPERABLE to terminate excessive positive reactivity insertions during power operation and while shut down. *
2. High Startup Rate Trip There are no safety analyses which take credit for functioning of the High Startup Rate Trip. The High Startup.Rate trip is used to trip the reactor when excore wide range power indicates an excessive rate of change. The High Startup Rate trip minimizes transients for events such as a continuous control rod withdrawal or a boron dilution event from low power levels. The trip may be operationally bypassed when THERMAL POWER is < lE-4% RTP, when poor counting
. statistics may lead to erroneous indication. It may also be operationally bypassed at > 13% RTP, where moderatpr temperature coefficient and fuel temperature
Palisades Nuclear Plant coefficient make high rate of change of power unlikely.
B 3.3.1-11 05/30/99
RPS Instrumentation B 3.3.1 BASES APPLICABLE 2. High Startup Rate Trip (continued) SAFETY ANALYSES There are only two wide range drawers, with each supplying contact input to auxiliary trip units in two RPS channels.
3. Low Primary Coolant System Flow Trip The Low PCS Flow trip provides DNB protection during events which suddenly reduce the PCS flow rate during power operation, such as loss of power to, or seizure of, a primary coolant. pump.
Flow in each of the four PCS loops is determined from pressure drop from inlet to outlet of the SGs. The total PCS flow is determined, for the RPS flow channels, by summing the loop pressure drops across the SGs and correlating this pressure sum with the sum of SG differential pressures which exist at 100% flow (four pump operation at full power Tave). Full PCS flow is that flow which exists at RTP, at full power Ta~* with four pumps operating .
4, 5. Low Steam Generator Level Trip The Low Steam Generator Level trips are provided to trip the reactor in the event of excessive steam demand (to prevent overcooling the PCS) and loss of feedwater events (to prevent overpressurization of the PCS).
The Allowable Value assures that there will be* sufficient water inventory in the SG at the time of trip to allow a safe and orderly plant shutdown and to prevent SG dryout assuming minimum AFW capacity. Each SG level is sensed by measuring the differential pressure in the upper portion of the downcomer annulus in the SG. These trips share four level sensing channels on each SG with the AFW actuation signal .
Palisades Nuclear Plant B 3.3.1-12 05/30/99
RPS Instrumentation B 3.3.1
_BA_S_E~S~~~~.:..;__~~~~~~~~~~~---'-~~~~~~~~_;_:~
APPLICABLE 6, 7. Low Steam Generator Pressure Trip SAFETY ANALYSES (continued) The Low Steam Generator Pressure trip provides protection against an excessive rate of heat extraction from the steam generators, which would result in a rapid uncontrolled cooldown of the PCS. This trip provides a mitigation function in the event of an MSLB. The Low SG Pressure channels are shared with the Low SG Pressure signals which isolate the steam and feedwater lines.
8. High Pressurizer Pressure Trip The High Pressurizer Pressure trip, in*conjunction with pressurizer safety valves and Main Steam Safety Valves (MSSVs), provides protection against overpressure conditions in the. PCS when at operating temperature. The safety analyses assume the High Pressurizer Pressure trip is OPERABLE during accidents and transients which suddenly reduce PCS cooling
(e.g., Loss of Load, Main Steam Isolation Valve (MSIV) closure, etc.) or which suddenly increase reactor power (e.g., rod ejection accident).
The High Pressurizer Pressure trip shares four safety grade instrument channels with the TM/LP trip, Anticipated Transient Without Scram (ATWS) and PORV circuits, and the Pressurizer Low Pressure Safety Injection Signal.
9. Thermal Margin/Low Pressure (TM/LP) Trip I
The TM/LP trip is provided to prevent reactor operation when the DNBR is insufficient. The TM/LP trip protects against slow reactivity or temperature increases, and against pressure decreases. The trip is initiated whenever the PCS pressure signal drops below a minimum value (P~") or a computed value (P~rl as described below, whichever is higher. The TM/LP trip uses Q Power, AS!, pressurizer pressure, and c~ld leg temperature (Tc) as inputs. Palisades Nuclear Plant B 3.3.1-13 05/30/99
RPS Instrumentation B 3.3.1
_BA_S_E_S~~---~~~----~---~~----~--'--'-~~'-'-~.;___;_~~~~~~~
APPLICABLE 9. Thermal Margin/Low Pressure (TM/LP) Trip (continued) SAFETY ANALYSES Q Power is the higher of core THERMAL POWER (~T Power) or nuclear power. The ~T power uses hot leg and cold leg RTDs as inputs. Nuclear power uses the power range excore channels as inputs. Both the ~T and excore power signals have provisions for calibration by calorimetric calculations. The AS! is calculated from the upper and lower power range excore detector signals, as explained in Section 1.1, 11 Definitions. 11 The signal is corrected for the difference between the flux at the core periphery and the flux at the detectors. The Tc value is the higher of the two cold leg signals. The Low Pressurizer Pressure trip 1i mit (P var) is calculated using the equations given in Table 3.3.1-2. The calculated limit (Pvar) is then compared to a fixed* Low Pressurizer Pressure trip 1i mit (P min). The auctioneered highest of these signals becomes the trip 1imi t (Ptrip). Ptri~ is compared to the measured PCS pressure and a trip signal is generated when the measured pressure for that channel is less than or equal to PtMp" A pre-trip alarm is also generated _ when P is less than or equal to the pre-trip setting,* Ptrip + t..P
The TM/LP trip setpoint is a complex function of these inputs and represents a minimum acceptable PCS pressure for the existing temperature and power conditions. It is compared to actual PCS pressure in the TM/LP trip unit .

Palisades Nuclear Plant B 3.3.1-14 05/30/99
RPS Instrumentation B 3.3.1
- BASES APPLICABLE SAFETY ANALYSES (continued)
10. Loss of Load Trip There are no safety analyses which take credit for functioning of the Loss of Load Trip.
The Loss of Load trip is provided to prevent lifting the pressurizer and main steam safety valves in the event of a turbine generator trip while at power. The trip is equipment protective. The safety analyses do not assume that this trip functions during any accident or transient. The Loss of Load trip uses a single pressure switch in the turbine auto stop oil circuit to sense a turbine trip for input to all four RPS auxiliary trip units.
11. Containment High Pressure Trip The Containment High Pressure trip provides a reactor trip in the event of a Loss of Coolant Accident (LOCA) or Main Steam Line Break (MSLB). The Containment High Pressure trip shares sensors with the Containment High Pressure sensing logic for Safety Injection,
12.
Containment Isolation, and Containment Spray. Each of these sensors has a single bellows which actuates two microswitches. One microswitch on each of four sensors provides an input to the RPS: Zero Power Mode Bypass Removal The only RPS bypass considered in the safety analyses is the Zero Power Mode 1ZPM) Bypass. The ZPM Bypass is used when the plant is shut down and it is desired to raise the control rods for control rod drop testing with PCS flow or temperature too low for the RPS Low PCS Flow, Low SG Pressure, or Thermal Margin/Low
, Pressure trips to be reset. ZPM bypasses are automatically removed if the wide range NI indicated power exceeds lE-4% RTP .
Palisades Nuclear Plant B 3.3;1-15 05/30/99
RPS Instrumentation B 3.3.1
*_BA_S_E_S~~~~~~~~~~~~~~~~...,....-~~~~~~~~~~~-
APPLICABLE 12. Zero Power Mode Bypass Removal (continued) SAFETY ANALYSES The safety analyses take credit for automatic removal of the ZPM Bypass if reactor criticality due to a Continuous Control Rod Bank Withdrawal should occur with the effected trips bypassed and PCS flow, pressure, or temperature below the values at which the RPS could be r~set. The ZPM Bypass would effectively be removed when the first wide range NI channel indication reached lE-4% RTP. With the ZPM Bypass for two RPS channels removed, the RPS would trip on one of the un-bypassed trips. This would prevent the reactor reaching an excessive power level. If a reactor criticality due to a Continuous Control Rod Bank Withdrawal should occur when PCS flow, steam generator pressure, and PCS pressure (TM/LP) were above their trip set points, a trip would terminate the event when power increased to the minimum setting (nominally 30%) of the Variable High Power Trip. In
'this case, the monitored parameters are at or nea~.
their normal operational values, and a trip initiated at 30% RTP provides adequate protection .
. The RPS design also includes automatic removal of the Operating Bypasses for the High Startup Rate and Loss of Load trips. The safety analyses do not assume functioning of either these trips or the automatic removal of their bypasses.
The RPS instrumentation satisfies Criterion 3 of 10 CFR 50.36(c){2). LCO The LCO requires all instrumentation performing an RPS Function to be OPERABLE. Failure of the trip unit (including its output relays), any required portion of the associated instrument channel, or both, renders the affected channel(s) inoperable and reduces the reliability of the affected Functions. Failure of an automatic ZPM bypass removal channel may also impact the associated instrument channel(s) and reduce the reliability of the affected Functions .
Palisades Nuclear Plant B 3.3.1-16 05/30/99
RPS Instrumentation BASES LCD (continued) B 3.3.1 Actions allow Trip Channel Bypass of individual channels, but the bypassed channel must be considered to be inoperable. The bypass key used to bypass a single channel cannot be simultaneously used to bypass that same parameter in other channels. This interlock prevents operation with more than one channel of the same Function bypassed. The plant is normally restricted to 7 days in a trip channel bypass, or otherwise inoperable condition before either restoring the Funct1on to four thannel operat~on (two-out-of~four logic) or placing the channel in trip (one-out-of-three logic). The Allowable Values are specified for each safety related RPS trip Function which is credited in the safety analysis. Nominal trip setpoints are specified in the plant procedures .. The nominal setpoints are selected to ensure plant parameters do not exceed the Allowable Value if the instrument loop is performing as required. Operation with a trip setpoint less conservative than the nominal trip setpoint, but within its Allowable Value, is acceptable. Each Allowable Value specified is more conservative than the analytical limit determined in the safety analysis in order to account for uncertainties appropriate to the trip
F~nction. These uncertainties are addressed as described in plant documents. Neither Allowable Values nor setpoints are specified for the non-saf~ty related RPS Trip Functions, since* no safety analysis assumptions would be violated if they are not.set at a particular value~
The following Bases for each trip Function identify the above RPS trip Function criteria items that are applicable* to establish the trip Function OPERABILITY .
. 1. Variable High Power Trip (VHPT)
This LCD requires all four channels of the VH~T Function to be OPERABLE. The Allowable Value is high enough to provide an operating envelope that prevents unnecessary VHPT trips during normal plant operations. The Allowable Value is low enough for the system to function adequately during reactivity addition events.
\'
Palisades Nuclear Plant B 3.3.1-17 05/30/99
RPS Instrumentation B 3.3.1
-- BASES LCO 1. Variable High Power Trip CVHPT) (continued)
The VHPT is designed to limit maximum reactor power to its maximum design and to terminate power excursions initiating at lower powers without power reaching this full power limit. During plant startup, the VHPT trip setpoint is initially at its minimum value, ~ 30%. Below 30% RTP, the VHPT setpoint is not required to "track" with Q Power, i.e., be adjusted to within 15% RTP. It remains fixed until manually reset, at which point it increases to ~ 15% above existing Q Power. The maximum allowable setting of the VHPT is 106.5% RTP~ Adding to this the,possible variation in trip setpoint due to calibration and instrument error, the maximum actual steady state power at which a trip would be actuated is 115%, which is the value assumed in the safety an~lysis.
2. High Startup Rate Trip
-This LCO requires four channels of High Startup Rate Trip Function to be OPERABLE in MODES 1 and 2. The High Startup Rate trip may be bypassed when the wide range NI indicates below lOE-4% or when THERMAL POWER .
is above 13% RTP. If a High Startup Rate trip is bypassed when power is between these limits, it must be considered to be inoperable. The High Startup Rate trip serves as a backup to the administratively enforced startup rate limit. The Function is not credited in the accident analyses; therefore, no Allowable Value for the trip or operating bypass Functions is derived from analytical limits and none is specified. The four channels of the High Startup Rate trip are derived from two wide range NI signal processing drawers. Thus, a failure in one wide range channel could render two RPS channels inoperable. It is acceptable to continue operation in this condition because the High Startup Rate trip is not credited in any safety analyses .
Palisades Nuclear Plant B 3.3.1-18 05/30/99
RPS Instrumentation B 3.3.1
_BA_S_E_S~~~~~~~~~~~~~~~~~~~~~~~~~~~
LCO 3. Low Primary Coolant System Flow Trip (continued) This LCO requires four channels of Low PCS Flow Trip Function to be OPERABLE.
\
This trip is set high enough to maintain fuel integrity during a loss of flow condition. The setting is low enough to allow for normal operating fluctuations from offsite power. The Low PCS Flow trip setpoint of 95% of full PCS flow insures that the reactor cannot operate when the flow rate is less than 93% of the nominal value considering instrument error,s. Full PCS flow is that flow which exists at RTP. at full power Tave' with four pumps operating. 4, 5. Low Steam Generator Level Trip This LCO requires four channels of Low Steam Generator Level Trip Function per steam generator to be OPERABLE . The 25.9% Allowable Value assures that there is an adequate water inventory in the steam generators when the reactor is critical and is based upon narrow range instrumentation. The 25.9% indicated level corresponds to the location of the feed ring. 6, 7. Low Steam Generator Pressure Trip This LCO requires four channels of Low Steam Generator Pressure Trip Function per steam generator to be OPERABLE. The Allowable Value of 500 psia is sufficiently below the full load operating value for steam pressure so as not to interfere with normal plant operation. but still high enough to provide the required protection in the event of excessive steam demand. Since excessive steam demand causes the PCS to cool' down, resulting in positive reactivity addition to the core, a reactor trip is required to offset that effect .
Palisades Nuclear Plant B 3.3.1-19 05/30/99
RPS Instrumentation B 3.3.1
_BA_S_E_S~~~~~~~~--~~~~~~~~~~~~--'-'-~~~~~~
LCO 8. High Pressurizer Pressure Trip (continued) This LCO requires four channels of High Pressurizer Pressure Trip Function to be OPERABLE. The Allowable Value is set high enough to allow for pressure increases in the PCS during normal operation (i.e., plant transients) not indicative of an abnormal condition. The setting is below the lift setpoint of the pressurizer safety valves and low enough to initiate a reactor trip when an abnormal condition is indicated.
9. Thermal Margin/Low Pressure (TM/LP) Trip This LCO requires four channels of TM/LP Trip Function to be OPERABLE.
The TM/LP.trip setpoints are derived from the core thermal limits through application of appropriate allowances for measurement uncertainties and processing errors. The allowances specifically . account for instrument drift in both power and inlet temperatures, calorimetric power measurement, inlet temperature measurement, and primary system pressure measurement. Other uncertainties including allowances for assembly power tilt, fuel pellet manufacturing tolerances, core flow measurement uncertainty and core bypass flow, inlet temperature measurement time delays, and ASI measurement, are included in the development of the TM/LP trip setpoint used in the accident analysis.
10. Loss of Load Trip The LCO requires four Loss of Load Trip Function channels to be OPERABLE in MODE 1 with THERMAL POWER z 17% RTP.
The Loss of Load trip may be bypassed or be inoperable with THERMAL POWER< 17% RTP, since it is no longer needed to prevent lifting of the pressurizer safe~y valves or steam generator safety valves in the event of a Loss of Load. Loss of Load Trip unit must be considered inoperable if it is bypassed when THERMAL POWER is above 17% RTP .
Palisades Nuclear Plant B 3.3.1-20 05/30/99
RPS Instrumentation B 3.3.1
_BA_S_E_S~~~~~~~~~~~~~~~-,-~~~~~~~~~~~~
LCO 10. Loss of Load Trip (continued) This LCO requires four RPS Loss of Load auxiliary trip units, relays 305L and 305R, and pressure switch 63/AST-2 to be OPERABLE. With those components OPERABLE, a turbine trip will generate a reactor trip. The LCO does not require the various turbine trips, themselves, to b~ OPERABLE. The Nuclear Steam Supply System and Steam Dump System are capable of accommodating the Loss of Load without requiring the use of ,the above equipment. The Loss of Load Trip Function is not credited in the accident analysis; therefore, an Allowable Value for the trip cannot be derived from analytical limits, and is not specified.
11. Containment High Pressure Trip This LCO requires four channels of Containment High Pressure Trip Function to be OPERABLE .
The Allowable Value is high enough to allow for small pressure increases in containment expected during normal operation (i.e., plant heatup) that are not indicative of an abnormal condition. The setting is low enough to initiate a reactor trip to prevent containment pressure from exceeding design pressure following a OBA and ensures the reactor is shutdown before initiation of safety injection and containment spray. 12 ZPM Bypass The LCO requires that four channels of automatic Zero Power Mode (ZPM) Bypass removal instrumentation be OPERABLE. Each channel of automatic ZPM Bypass removal includes a shared wide range NI channel, an actuating bistable in the wide range drawer, and a relay in the associated RPS cabinet. Wide Range NI channel 1/3 is shared between ZPM Bypass removal channels A and C; Wide Range NI channel 2/4, between ZPM Bypass removal channels B and D. An operable bypass removal cllannel must be capable of automatically removing the capability to bypass the
Palisades Nuclear Plant affected RPS trip channels with the ZPM Bypass key switch at the proper setpoint.
8 3.3.1-21 05/30/99
RPS Instrumentation B 3.3.1
_BA_S_E_S~~~~~~~~~~~~~~~~~~~~~~~~~~~
APPLICABILITY This LCO requires all safety related trip functions to be
OPERABLE in accordance with Table 3.3.1-1. While in MODES 3, 4, or 5, if PCS boron concentration is at REFUELING BORON CONCENTRATION, or if no more than one full-length control rod is capable of being withdrawn, the RPS Function is already fulfilled (the safety analyses and the SHUTDOWN MARGIN definition both use the assumption that the highest worth withdrawn full-length control rod will .fail to insert on a trip) and the safety analyses assumptions and SHUTDOWN MARGIN requirements will be met without the RPS trip Function.
The High Startup Rate Trip Function is required to be OPERABLE in MODES 1 and 2, but may be bypassed when the associated wide range NI channel indicates below lE-4% power, when poor counting statistics may lead to erroneous indication. It may also be bypassed when THERMAL POWER is above 13% RTP, where moderator temperature coefficient and fuel temperature coefficient make high rate of change of power unlikely. In MODES 3, 4, 5, and 6, the High Startup Rate trip is not required to be OPERABLE. Wide range channels are required to be OPERABLE in MODES 3, 4, and 5, by LCO 3.3.9, Neutron Flux Monitoring Channels," and in 11 MODE 6, by LCO 3.9.2, Nuclear Instrumentation." 11 The Loss of Load trip is required to be OPERABLE with THERMAL POWER at or above 17% RTP. Below 17% RTP, the ADVs are capable of relieving the pressure due to a Loss of Load event without challenging other overpressure protection. The trips are designed to take the reactor subcritical, maintaining the SLs during AOOs and assisting the ESF in providing acceptable consequences during accidents .
Palisades Nuclear Plant B 3.3.1-22 05/30/99
RPS Instrumentation B 3.3.1
- _BA_S_E_S~~~~~~~~~~~~~~~~~~~~~~~~~~--'---"-
ACTIONS The most common causes of channel inoperability are outright failure of loop components or drift of those loop components which is sufficient to exceed the tolerance provided in the plant setpoint analysis. Loop component failures are typically identified by the actuation of alarms due to the channel failing to the "safe" condition, during CHANNEL CHECKS (when the instrument is compared to the redundant channels), or during the CHANNEL FUNCTIONAL TEST (when an automatic component might not respond properly). Typically, the drift of the loop components is found to be small and results in a.delay of actuation rather than a total loss of function. Excessive loop component drift would, most likely, be identified during a CHANNEL CHECK (when the instrument is compared to the redundant channels) or during a CHANNEL CALIBRATION (when instrument loop components are checked against reference standards). In .the event a channel's trip setpoint is found nonconservative with respect to the Allowable Value~ or the transmitter, instrument loop, signal processing electronics, or RPS bistable trip unit is found inoperable, all affected
Functions provided by that channel must be declared inoperable, and the plant must enter the Condition for the

particular protection Functions affected .
When the number of inoperable channels in a trip Function exceeds that specified in any related Condition associated with the same trip Function, then the plant is outside the safety analysi.s. Therefore, LCO 3~0.3 is immediately entered if applicable in the current MODE of operation. A' Note has been added to the ACTIONS to clarify the application of the Completion Time rules. The Conditions of this Specification may be entered independently for each Function. The Completion Times of each inoperable Function will be tracked separately for each function, starting from the time the Condition was entered. Condition A applies to the failure of a single channel in any required RPS Function~ except High Startup Rate, Loss of Load, or ZPM Bypass Removal. (Condition A is modified by a Note stating that this Condition does not apply to the High
Startup Rate, Loss of Load, or ZPM Bypass Removal Functions.
The failure of one channel of those Functions is addressed by Conditions B, C, or D.) '
Palisades Nuclear Plant B 3.3.1-23 05/30/99
RPS Instrumentation B 3.3.1
_BA_S_E_S~~~~-"-~~~~~~~~~~~~~~~~~~~~~~
AGTIONS A.1 (continued) If one RPS bistable trip unit or associated instrument channel is inoperable, operation is allowed to continue. Since the trip unit and associated instrument channel combine to perform the trip function, this Condition is also appropriate if both the trip unit and the associated instrument channel are inoperable. Though not required, the inoperable channel may be bypassed or tripped. The provision of four trip channels al.lows one channel to be bypassed (removed from service) during operations, placing the RPS in two-out-of-three coincidence logic. The failed channel must be restored to OPERABLE status or placed in trip within 7 days. Required Action A.1 places the Function in a one-out-of-three configuration. In this configuration, common cause failure of dependent channels cannot prevent trip. The Completion Time of 7 days is based on operating experience, which has demonstrated that a random failure of a second channel occurring during the 7 day period is a low
*probability event.
I .* Condition B applies to the failure of a single High Startup Rate trip unit or associated instrument channel. If one trip unit or associated instrument channel fails, it must be restored to OPERABLE status prior to entering MODE 2 from MODE 3. A shutdown provides the appropriate opportunity to repair the trip function and conduct the necessary testing. The Completion Time is based on the fact that the safety analyses take no credit for the functioning of this trip .
Palisades Nuclear Plant B 3.3.1-24 05/30/99
RPS Instrumentation B 3.3.1
_BA_S_E_S~~~~~~~~~~~~~~~~~~~~~~~~~~~
ACTIONS C.1 (continued) Condition C applies to the failure of a single Loss of Load or associated instrument channel. If one trip unit or associated instrument channel fails, it must be restored to OPERABLE status prior to THERMAL POWER
~ 17% RTP following a shutdown. If the plant is shutdown at the time the channel becomes inoperable, then the failed channel must be restored to OPERABLE status prior to THERMAL POWER~ 17% RTP. For this Completion Time, 11 following a shutdown means this Required Action does not have to be 11 completed until prior to THERMAL POWER ~ 17% RTP for the first time after the plant has been in MODE 3 following entry into the Condition. The Completion Ti.me is based on the fact that the safety analyses take no credit for the functioning of this trip.
D.1 and D.2 Condition D applies when one or more automatic ZPM Bypass* removal channels are inoperable. If the ZPM Bypass removal channel cannot be restored to OPERABLE status, the affected ZPM Bypasses must be immediately .removed, or the bypassed RPS trip Function channels must be immediately declared to be inoperable. Unless additional circuit failures exist, the ZPM Bypass may be removed by placing the associated 11 Zero Power Mo.de Bypass key operated switch in *the normal 11 position. A trip channel which is actually bypassed, other than as. allowed by the Table 3.3.1~1 footnotes, cannot perform its specified safety function and must immediately be declared to be inoperable.
E.1 and E.2 Condition E applies to the failure *of two channels in any RPS Function, except ZPM Bypass Removal Function. (The failure of ZPM Bypass Removal Functions is addressed by Condition D.).
Condition E is modified by a Note stating that this Condition does not apply to the ZPM Bypass Removal Function . I
Palisades Nuclear Plant B 3.3.1-25 05/30/99
RPS Instrumentation B 3.3.1
BASES ACTIONS E.1 and E.2 (continued)
The Required Actions are modified by a Note stating that LCO 3.0.4 is not applicable. The Note was added td allow the changing of MODES even though two channels are inoperable, with one channel tripped. _MODE changes in this configuration are allowed because two trip channels for the affected function remain OPERABLE. A trip occurring in either or both of those channels would cause a reactor trip. While it is conceptually possible that, if the two operable channels were those that do not have total channel separation in their cable routings, a single failure could disable both from tripping, in reality, such failures are extremely unlikely. Most failures involving a common cable fault would cause the affected channel(s) to fail in the de-energized condition, thereby initiating a reactor trip not preventing one. In this configuration, the protection system is in a one-out-of-two logic, and the probability of a common cause failure affecting both of the OPERABLE channels during the 7 days permitted is remote. Required Action E.1 provides for placing one inoperable channel in trip within the Completion Time of 1 hour. Though not required, the other inoperable channel may be (trip channel) bypassed. This Comple~ion Time is sufficient to allow the operaior to take all appropriate actions for the failed channels while ensuring that the r~sk involved in operating with the failed channels is acceptable. With one channel of protective instrumentation bypassed or inoperable in an unripped condition, the RPS is in a two-out-of-three logic for that function; but with another channel failed, the RPS may be operating in a two-out-of-two logic. This is outside the assumptions made in the analyses and should be corrected. To correct the problem, one of the inoperable channels is placed in trip. This places the RPS in a one-out-of-two for that function logic. If any of the other unbypassed - channels for that function receives a trip signal, the reactor will trip. Action E.2 is modified by a Note stating that this Action does not apply to (is not required for) the High Startup Rate and Loss of Load Functions *
Palisades Nuclear Plant B 3.3.1-26 05/30/99
RPS Instrumentation* B 3.3.1
_BA_S_E_S______..;;____:______.;.;......;._..;..___;:_________:.. . . .;______:_____::..__:__.:....__________
ACTIONS E.1 and E~2 (~ontinued) One channel is required to be restored to OPERABLE status within 7 days for reasons similar to those stated under Condition A. After one channel is restored to OPERABLE status, the provisions of Condition A still apply to the remaining inoperable channel. Therefore, the channel that is still inoperable after completion of Required Action E.2 must be placed in trip if more than 7 days have elapsed since the initial channel failure. F.1 The power range excore channels are used to generate the internal AS! signal used as an input to the TM/LP trip. They also provide input to the Thermal Margin Monitors for determination of the Q Power input for the TM/LP trip and the VHPT~ If two power range excore channels cannot be restored to OPERABLE status, power is restricted or reduced during subsequent operations because of increased uncertainty associated with inoperable power range excore
*channels which provide input to those trips .
The Completion Time of 2 hours is adequate to reduce power in an orderly manner without challenging plant systems. G.1. G.2.1. -and G.2.2 Conditiorr G is entered when the Required* Action and associated Completion Time of Condition A, B, C, D, E, or F are not met, or if the control room ambient air temperature exceeds 90°F.
If the control room ambient air temperature exceeds 90°F, all Thermal Margin Monitor channels are rendered inoperable because their environmental qualification temperature limit is exceeded. In this condition, or if the Required Actions and associated Completion Times are not met, the reactor must be pl aced in a condition in which *the LCO does not apply. To accomplish this, the plant must be placed in MODE 3, with no more than one full-length control rod capable of being withdrawn or with the PCS boron concentration at REFUELING BORON CONCENTRATION in 6 hours *
Palisades Nuclear Plant B 3.3.1-27 05/30/99
RPS Instrumentation B 3.3.1
BASES ACTIONS G.l. G.2.1. and G.2.2 (continued)
The Completion Time is reasonable, based on operating experience, for placing the plant in MODE 3 from full power conditions in an orderly manner and without challenging plant systems. The Completion Time is also reasonable to ensure that no more than one full-length control rod is capable of betng withdrawn or that the PCS boron concentration is at REFUELING BORD~ CONCENTRATION. SURVEILLANCE The SRs for any particular RPS Function are found in the SR REQUIREMENTS column of Table 3.3.1-1 for that Function. Most Functions are subject to CHANNEL CHECK, CHANNEL FUNCTIONAL TEST, and CHANNEL CALIBRATION. While Palisades is not committed to performing all testing discussed in ANSI/IEEE Standard 338-1977, CHANNEL CHECKS, CHANNEL FUNCTIONAL TESTS, AND CHANNEL CALIBRATIONS are performed in accordance with the guidance of ANSI/IEEE Standard 338-1977, which is.endorsed by Regulatory Guide 1.118 .
SR 3. 3.1.1 Performance of the CHANNEL CHECK once every 12 hours ensures that gross failure of instr~~entation has not occurr~d. A CHANNEL CHECK is normally a comparison of the parameter indicated on one channel to a ~imilar parameter on other channels. It is based on the assumption that instrument channels monitoring the s~me parameter should read approximately the same value. Significant deviations between the two instrument channels could be an indication of excessive instrument drift in one of the channels or of something even more serious. Under most conditions, a CHANNEL CHECK will detect gross channel failure; thus, it is key to verifying that the instrumentation continues to operate properly between each CHANNEL CALIBRATION.
Agreement criteria are determined by the plant staff based on a combination of the channel instrument uncertainties, including indication and readability. If a channel is outside the criteria, it may be an indication that the transmitter or the signal processing equipment has drifted outside its limits .
Palisades Nuclear Plant B 3.3.1-28 05/30/99
RPS Instrumentation B 3.3.1
_BA_S_E_S~~~~~~~~~~~~~~~~~---..__~-'-~~~~~--'
SURVEILLANCE SR 3.3.1.1 (continued) REQUIREMENTS The Containment High Pressure and Loss of Load channels are pressure switch actuated. As such, they have no associated control room indicator and do not require a CHANNEL CHECK. The Frequency, about once every shift, is based on operating experience that demonstrates the rarity of channel failure. Since the probability of two random failures in redundant channels in any 12 hour period is extremely low, the CHANNEL CHECK minimizes the chance of loss of protective function due to failure of redundant channels. The CHANNEL CHECK supplements less formal, but more frequent, checks of channel OPERABILITY during normal operational use of the displays associated with the LCO required channels. SR 3.3.1.2 This SR verifies that the control room ambient air temperature is within the environmental qualification temperature limits for the most restrictive RPS components, I which are the Therma 1 Margin Mani to.rs. These monitors
provide input to both the VHPT Function and the TM/LP Trip Function. The 12 hour.Frequency is reasonable based on engineering SR 3.3.1.3 judgement. and operating experience.
A daily calibration (heat balance) is performed when THERMAL POWER is ~ 15%. The daily calibration consists of adjusting the 11 nuclear power calibrate 11 potentiometers to agree with the calorimetric calculation if the absolute difference is ~ 1.5%. Nuclear power is adjusted via a potentiometer, or THERMAL POWER is adjusted via a Thermal Margin Monitor bias number, as necessary, in accordance with the daily calibration (heat balance) procedure. Performance of the daily calibration ensures that the two inputs to the Q power measurement are indicating accurately with respect to the much more accurate secondary calorimetric ca lcul ati on. The Frequency of 24 hours is based on plant operating experience and takes into account indications and alarms located in the control room to detect deviations in channel outputs .
Palisades Nuclear Plant B 3.3.1-29 05/30/99
RPS Instrumentation B 3.3.1
_BA_S_E_S~~~~~~~~~~~--'"-'--'--~~~~~~~~~~~~~
SURVEILLANCE SR 3.3.1.3 (continued) REQUIREMENTS The Frequency is modified by a Note indicating this Surveillance must be performed within 12 hours after THERMAL POWER is ~ 15% RTP. The secondary calorimetric is inaccurate at lower power levels. The 12 hours allows time requirements for plant stabilization, data taking, and instrument calibration. SR 3 .3.1.4 It is necessary to calibrate the power range excore channel upper and lower subchannel amplifiers such that the measured AS! reflects the true core power distribution as determined by the incore detectors. AS! is utilized as an input to the TM/LP trip function where it is used to ensure that the measured axial power profiles are bounded by the axial power
*profiles used in the development of the T;niet limitation of LCO 3.4.1. An adjustment of the excore channel is necessary only if individual excore channel AS! differs from AXIAL OFFSET, as measured by the incores, by greater than 0.02.
A Note indicates the Surveillance is not required to have been performed until 12 hours after THERMAL POWER is
~ 25% RTP. Uncertainties in the excore and incore measurement process make it impractical to calibrate when THERMAL POWER is < 25% RTP. The 12 hours allows time for plant stabilization, data taking, and instrument
calibration.
The 31 day Frequency is adequate, based on operating experience of the excore linear amplifiers and the slow burnup of the detectors. The excore readings are a strong function of the power produced in the peripheral fuel bundles and do not represent an integrated reading across the core. Slow changes in neutr*on flux during the fuel cycle can also be detected at this Frequency. SR 3.3.1.5 A CHANNEL FUNCTIONAL TEST is performed on each RPS instrument channel, except Loss of Load and High Startup Rate, every 92 days to ensure the entire channel will perform its intended function when needed. For the TM/LP Function, the constant.s associated with the Thermal Margin Monitors must be verified to be withi~ tolerances .
Palisades Nuclear Plant B 3.3.1-30 05/30/99
RPS Instrumentation B 3.3.1
_B_AS_E~S~~~~--'"-~~~~~~~~~~~-:-~~~~~~~~~~~
SURVE I LLANCE SR 3.3.1.5 (continued) REQUIREMENTS A successful test of the required contact(s) of a channel relay may be performed by the verification of the change of state of a single contact of the relay. This clarifies what is an acceptable CHANNEL FUNCTIONAL TEST of a relay. This is acceptable because all of the other required contacts of the relay are verified by other Technical Specifications and non-Technical Specifications tests at least once per refueling interval with applicable extensions. Any setpoint adjustment must be consistent with the assumptions of the current setpoint analysis. The Frequency of 92 days is based on the reliability analysis present~d in topical report CEN-327, "RPS/ESFAS Extended Test Interval Evaluation" (Ref. 5). SR 3.3.1.6 A calibration check of the power range excore channels using the internal test circuitry is required every 92 days. This SR uses an internally generated test signal to check that the 0% and 50% levels read within limits for both the upper and lower tietector, both on the analog meter and on the TMM scteeh .. This check verifies that neither the zero point nor the amplifier gain adjustment have undergone excessive drift since the previous complete CHANNEL CALIBRATION. The Frequency of 92 days is acceptable, based on plant operating experience, and takes into account indications and alarms available to the operator in the control room. SR 3.3.1.7 A CHANNEL FUNCTIONAL TEST on the Loss of Load and High Startup Rate channels is performed prior to a reactor startup to ensure the entire channel will perform its intended function if required. A successful test of the required contact(s) of a channel relay may be performed by the verification of the change of state of a single contact of the relay. This clarifies what is an acceptable CHANNEL FUNCTIONAL TEST of a relay. This is acceptable because ~11 of the other required contacts of the relay are verified by other Technical Specifications and
non-Technical Specifications tests at least once per refueling interval with applicable extensions.
Palisades Nuclear Plant B 3.3.1-31 05/30/99
RPS Instrumentation B 3.3.1
BASES SURVEILLANCE REQUIREMENTS SR 3.3.1.7 (continued)
The High Startup Rate trip is actuated by either of the Wide Range Nuclear Instrument Startup Rate channels. NI-1/3 sends a trip signal to RPS channels A and C; NI-2/4 to channels B and D. Since each High Startup Rate channel would cause a trip on two RPS channels, the High Startup Rate trip is not tested when the reactor is critical. The four Loss of Load Trip channels are all actuated by a single pressure switch monitoring turbine auto stop oil pressure which is not tested when the reactor is critical. Operating experience has shown that these components usually pass the Surveillance when performed at a Frequency of once per 7 days prior to e~ch reactor startup. SR 3.3.1.8 SR 3.3.1.8 is the performance of a CHANNEL CALIBRATION every 18 months. CHANNEL CALIBRATION is a complete check of the instrument.
channel including the sensor (except neutron detectors) .
The Surveillance verifies that the channel responds to a measured parameter within the necessary range and accuracy. CHANNEL CALIBRATION leaves the channel adjusted to account for instrument drift between successive calibrations to ensure that the channel remains operational between successive tests. CHANNEl CALIBRATIONS must be consistent with the setpoint analysis. The bistable setpoints must be found to trip within the
*Allowable Values specified in the LCO and left set consistent with the assumptions of the setpoint analysis.
The Variable High Power Trip setpoint shal.l be verified to reset properly at several indicated power levels during (simulated) power increases and power decreases. The as found and as left values must also be recorded and reviewed for consistency with the assumptions of the setpoint analysis. As part of the CHANNEL CALIBRATION of the wide range Nuclear Instrumentation, automatic removal of the ZPM Bypass for the Low PCS Flow, TM/LP must be verified to assure that these trips are available when required .
Palisades Nuclear Plant B 3.3.1-32 05/30/99
RPS Instrumentation B 3.3.1
_BA_S_E_S~.;_-'--'--'-~----"~..;._~~..;.___;;~~~~~~~~~~~~~~~
SURVEILLANCE SR 3.3.1.8 (continued) REQUIREMENTS The Frequency is based upon the assumption of an 18 month calibration interval for the determination of the magnitude of equipment drift. This SR is modified by a Note which states that it is not necessary to calibrate neutron detectors because they are passive devices with minimal drift and because of the difficulty of simulating a meaningful signal. Slow changes in power range excore neutron detector sensitivity are
compensated for by performing the daily calorimetric calibration (SR 3.3.1.3) and the monthly calibration using the incore detectors (SR 3.3.1.4). Sudden changes in detector performance would be noted during the required CHANNEL CHECKS (SR 3.3.1.1).
REFERENCES 1. 10 CFR 50, Appendix A, GDC 21
2. 10 CFR 100
3.
4. 5. IEEE FSAR~ Stan~ard 279-1971, April 5, 1972 Chapter 14 CEN-327,* June 2, 1986, including Supplement 1, March 3, *1989
Palisades Nuclear Plant. B 3.3.1-33 05/30/99
RPS Instrumentation B 3.3.1 Table B 3.3.1-1 (page 1 of 1)
REQUIRED INSTRUMENT CHANNELS Instruments Affecting Multiple Specifications Source Range NI-1/3 & 2/4, Count Rate Signal AFFECTED SPECIFICATIONS 3.3.9 3.9.2 Source Range NI-1/3, Count Rate Indication @ C-150 Panel 3.3.8 #1 Wide Range NI-1/3 & 2/4, Flux Level 10~ Bypass 3.3.1 #3,6,7,9,&12 Wide Range NI-1/3 & 2/4, Startup Rate 3.3.1 #2 Wide Range NI-1/3 & 2/4, Flux Level Indication 3.3.7 #3 3.3.9 Power Range NI-5, 6, 7 & 8, Tq 3.2.1 3.2.3 Power Range NI-5, 6, 7 & 8, Q Power 3.3.1 #1 & 9 Power Range NI-5, 6, 7 & 8, ASI 3.3.1 #9 3.2.1 3.2.4 Power Range NI-5, 6, 7 & 8, Loss of Load/High Startup Rate Bypass 3. 3 .1 #2 & 10 PCS TC TT-0112 & 0122 CC & CD, Temperature Signal (SMM) 3.3.7 #5 PCS TC TT-0112 & 0122 CA, CB, CC & CD, Temperature Signal (Q Power & TMM) 3.3.1 #1 & 9 3.4.1.b PCS TC TT-0112CA & 0122CB, Temperature Signal (LTOP) 3.4.12.b.1 PCS TC TT-0112CC & 0122CD (PTR-0112 & 0122) Temperature Indication 3.3.7 #2 PCS TC TT-0112CA, Temperature Signal (SPI ~T Power for PDIL Alarm Circuit) 3.1.6 PCS TC TT-0122CB, Temperature Signal (PIP ~T Power for PDIL Alarm Circuit) 3.1.6 PCS TH TT-0112 & 0122 HC & HD, Temperature Signal (SMM) 3.3.7 #5 PCS TH TT-0112HC & 0122HD (PTR-0112 & 0122) Temperature Indication 3.3.7 #1

PCS TH TT-0112 & 0122 HA, HB, HC & HD, Temperature Signal (Q Power)
PCS TH TT-0112HA, Temperature Signal (SPI ~T Power for PDIL Alarm Circuit) PCS TH TT-0122HB, Temperature Signal (PIP ~T Power for PDIL Alarm Circuit) Thermal Margin Monitor PY-0102A, B, C, & D Pressurizer Pressure PT-0105A & B, Pressure Signal (WR Indication & LTOP) Pressurizer Pressure PT-0102A, B, C & D, Pressure Signal (RPS & SIS) 3.3.1 #1 & 9 3.1.6 3 .1.6 3.3.1 #1 & 9 3.3.7 #5 3.3.4.12.b.1 3.3.1 #8 & 9 3.3.3 #1.a Pressurizer Pressure PT-0104A & B, Pressure Signal (LTOP & SDC Interlock) 3.4.12.b.1 3.4.14 Pressurizer Pressure PI-0110, Pressure Indication @ C-150 Panel 3.3.8 #2 SG Level LT-0751 & 0752 A, B, C & D, Level Signal (RPS & AFAS) 3.3.1 #4 & 5 3.3.3 #6.a & 6.b SG Level LI-0757C & 0758C, Wide Range Level Indication @ C-150 Panel 3.3.8 #10 & 11 SG Level LI-0757 & 0758 A & B, Wide Range Level Indication 3.3.7 #11 & 12 SG Pressure PT-0751 & 0752 A, B, C&D, Pressure Signal (RPS & SG Isolation) 3.3.1 #6 & 7 3.3.3 #4.a & 4.b SG Pressure PIC-0751 & 0752 A, B, C & D, Pressure Indication 3.3.7 #13 & 14 SG Pressure PI-0751E & 0752E, Pressure Indication @ C-150 Panel 3.3.8 #8 & 9 Containment Pressure PS-1801, 1802, 1803&1804, Switch Output (RPS) 3.3.l'#ll Containment Pressure PS-1801, 1802A, 1803 & 1804A, Switch Output (ESF Actuation) 3.3.3 #2.a Containment Pressure PS-1801A, 1802, 1803A & 1804, Switch Output (ESF Actuation) 3.3.3 #2.b Note: The information provided in this table is intended for use as an aid to distinguish those instrument channels which provide more than one required function and to describe which specifications they affect. The information in this table should not be taken as inclusive for all instruments nor affected specifications. Palisades Nuclear Plant B 3.3.1-34 05/30/99
RPS Logic and Trip Initiation B 3.3.2
B 3.3 INSTRUMENTATION B 3.3.2 Reactor Protective System (RPS) Logic and Trip Initiation BASES BACKGROUND The RPS initiates a reactor trip to protect against violating the acceptable fuel design limits and reactor coolant pressure boundary integrity during Anticipated Operational Occurrences (AOOs). By tripping the reactor, the RPS also assists the Engineered Safety Features (ESF) systems in mitigating accidents.
The protection and monitoring systems have been designed to ensure safe operation of the reactor. This is achieved by specifying Limiting Safety System Settings (LSSS) in terms of parameters directly monitored by the RPS, as well as LCOs on other reactor system parameters and equipment performance. The LSSS, defined in this Specification as the Allowable Value, in conjunctio~ with the LCOs, establish th~ threshold for protective system action to prevent exceeding acceptable limits during Design Basis Accidents (DBAs). During AOOs, which are those events expected to occur one or more times during the plant life, the acceptable limits are:
The Departure from Nucleate Boiling Ratio (DNBR) shall be maintained above the Safety Limit (SL) value to prevent departure from nucleate boiling;

Fuel centerline melting shall not occur; and

The Primary Coolant System (PCS) pressure SL of 2750 psia shall not be exceeded.
Maintaining the parameters within the above values ensures that the offsite dose will be within the 10 CFR 50 (Ref. 1) and 10 CFR 100 (Ref. 2) criteria during AOOs. Accidents are events that are analyzed even though they are not expected to occur during the plant life. The acceptable limit during accidents is that the offsite dose shall be maintained within an acceptable fraction of 10 CFR 100 (Ref. 2) limits. Different accident categories allow a different.fraction of these limits based on probability of occurrence. Meeting the acceptable dose limit for an accident category is considered having acceptable consequences for that event.
**\
Palisades Nuclear Plant B 3.3.2-1 05/30/99
RPS Logic and Trip Initiation B 3.3.2 ---* BASES BACKGROUND The RPS is segmented into four interconnected modules. (continued) . These modules are:
Measurement channels (or pressure switches);

Bistable trip units;

Matrix Logic; and

Trip Initiation Logic.
This LCO addresses the RPS Logic (Matrix Logic and Trip Initiation Logic), including Manual Trip capability. LCO 3.3.1, "Reactor Protective System (RPS) Instrumentation," provides a description of the role of the measurement channels and associated bistable trip units in the RPS. The RPS Logic is summarized below: RPS Logic The RPS Logic, consisting of Matrix Logit and Trip Initiation Logic, employs a scheme that provides a reactor trip when trip units in any twb of the four channels sense the same 'nput parameter trip. This is called a two-out-of-four trip logic. This logic and the clutch power supply configuration are shown - in FSAR Figure 7-1 (Ref. 3). - Bistable trip unit relay contact outputs from the four channels are configured into s1x logic matrices. Each logic matrix checks for a coincident trip in the same parameter in two trip unit channels. The matrices are designated the AB~ AC, AD, BC, BO, and CD matrices to reflect the bistable trip unit cnannels being monitored. Each logic matrix contains four normally energized matrix relays. When a coincidence is detected, consisting of a trip in the* same Function in the two channels being monitored by the logic matrix, all four matrix relay coils de-energize. The matrix re 1ay contacts are arrang*ed into trip paths, with one of the four matrix relays in each matrix opening contacts in one of the four trip paths. Each trip path provides power to one of the four normally energized clutch power supply "M-contactors" (Ml, M2, M3, and M4). The trip paths thus each have six contacts in series, one from each matrix, and perform a logical OR function; de-energizing the M-contactors if any one or more of the six logic matrices indicate a coincidence
condition .
Palisades Nuclear Plant B 3.3.2-2 05/30/99
RPS Logic and Trip Initiation B 3.3.2
BASES BACKGROUND (continued)
When a coincidence occurs in two RPS channels, all four matrix relays in the affected matrix de-energize. This in turn de-energizes all four M-contactors, which interrupt AC input power to the four clutch power supplies, allowing the full-length control rods to insert by gravity. Manual reactor trip capability is afforded by two main control panel-mounted pushbuttons. One of these (on Control Panel C0-2) opens contacts in series with each of the four trip paths, de-energizing all M-contactors. The other pushbutton (on Control Panel C0-6) opens circuit breakers which provide AC input power to the Mcontactor contacts and downstream clutch power supplies. Thus depressing either pushbutton will cause a reactor trip. De-energizing the M-contactors removes AC power to the four clutch power supply inputs. Contacts from M~contactors Ml and M2 are in series with each other and in the AC power supply path to clutch power supplies PSl and PS2 (these *constitute a 11 trip leg"). M3 and M4 are similarly arranged with respect to clutch power supplies PS3 and PS4 (these constitute a second "trip leg"). Approximately half of the control rod clutches receive power from auctioneered clutch power supplies 1 and 3. The remaining control rod clutches receive clutch power from auctioneered ~lutch power s~pplies 2 and 4. Matrix Logic refers to the matrix power supplies, trip channel bypass contacts, and interconnecting RPS cabinet matrix wiring between bistable and auxiliary trip unit relay contacts, including the matrix relays. Contacts in the bistable and auxiliary trip units are excluded from the Matrix Logic definition, since they are addressed as part of the
. instrumentation channel.
The Trip Initiation Logic consists of the M-contactor isolation transformers, all interconnecting wiring, and the M-contactors. Manual trip circuitry includes both manual reactor trip pushbuttons C0-2 and C0-6, and the interconnecting wiring necessary to effect de-energization of the clutch power supplies. '
Palisades Nuclear Plant B 3.3.2-3 05/30/99
RPS Logic and Trip Initiation B 3.3.2
BASES BACKGROUND (continued)
Neither the clutch power supplies nor the AC input power source to these supplies is considered as safety related. Operation may continue with one or two selective clutch power supplies de-energized. It is possible to change the two-out-of-four RPS Logic to a two-out-of-three logic for a given input parameter in one channel at a time by Trip Channel Bypassing the RPS Trip unit output contacts in the Matrix Logic "Ladder." . Trip Channel Bypassing a trip unit effectively shorts the trip unit relay contacts in the three matrices associated with that channel. Thus, the bypassed trip units will function normally, producing normal channel trip indication and annuntiation, but a reactor trip will not occur unless two additional channels indicate a trip condition. Trip Channel Bypassing can be simultaneously performed on any number of parameters in any number of channels, providing each parameter is bypassed in only one channel at a time. A single bypass key for each trip function interlock prevents simultaneous Trip Channel Bypassing of the same parameter in more than one channel. Trip Channel Bypassing is normally employed during maintenance or testing *
Functional testing of the entire RPS, from trip unit inp~t through the de-energizing of individual sets of clutch power supplies, can be performed either at power or during shutdown and is normally performed on a quarterly basis. FSAR Section 7.2 (Ref. 4) explains RPS testing in more detail.
APPLICABLE Reactor Protective System (RPS) Logic SAFETY ANALYSES The RPS Logic provides for automatic trip initiation to avoid exceeding the SLs during AOOs and to assist the ESF systems in .I ensuring acceptable consequences during accidents. All transients and accidents that call for a reactor trip assume* the RPS Logic. is functioning as designed. Manual Trip There are no accident analyses that take credit for the Manual Trip; however, the Manual Trip is part of the RPS circuitry. It is used by the operator to shut down the reactor whenever any parameter is rapidly trending toward its trip setpoint. *A Manual Trip accomplishes the same results as any one of the automatic trip Functions. The RPS Logic and Trip Initiation satisfy Criterion 3 of 10 CFR 50.36(c).(2). Palisades Nuclear Plant B 3.3.2-4 05/30/99
RPS Logic and Trip Initiation B 3.3.2
BASES LCO Reactor Protective System (RPS) Logic Failures of individual trip unit relays and their* contacts are addressed in LCO 3.3.1. This Specification addresses fai,lures of the Matrix Logic not addressed in the above, such as the failure of matrix relay power supplies or the failure of the trip channel bypass contact in the bypas~ condition.
Loss of a single vital bus will *de-energize one of the two power supplies in each of three matrices. Because of power supply auctioneering, all four matrix relays will remain energized in each affected matrix. This de-energization of up to three matrix power supplies due to a single failure is to be treated as a single channel failure. Each of the four Trip Initiation Logic channels de-energizes one set of clutch power supplies if any of the six coincidence matrices de-energize their associated matrix relays. They thus perform a logical OR function. Trip Initiation Logic channels 1 and 2 receive AC power from preferred AC bus Y-30. Trip Initiation Logic channels 3 and 4 receive AC input power from preferred AC bus Y-40. Because of clutch power supply .output auctioneering, it is possible to de-energize either input bus without de-energizing control rod clutches. *
1. Matrix Logic This LCO requires six channels of Matrix Logic to be OPERABLE in MODES 1 and 2, and in MODES 3, 4, and 5 when more than one full-length control rod is capable of being withdrawn and the PCS boron concentration is less.than REFUELING BORON CONCENTRATION.
*(
2. Trip Initiation Logic This LCO requires four channels of Trip Initiation Logic to be OPERABLE in MODES 1 and 2, and in MODES 3, 4, and 5 when more than one full-length control rod is capable of being withdrawn and the PCS boron concentration is less than REFUELING BORON CONCENTRATION *
Palisades Nuclear Plant B 3.3.2-5' 05/30/99
RPS Logic and Trip Initiation B 3.3.2
BASES LCO (continued)
3. Manual Trip.
The LCO requires both Manual Trip channels to be OPERABLE in MODES 1 and 2, and in MODES 3, 4, and 5 when more than one full-length control rod is capable of being withdrawn and the PCS boron concentration is less than REFUELING BORON CONCENTRATION. Two independent pushbuttons are provided. Each pushbutton is considered a channel. Depressing either pushbutton interrupts power to all four clutch power suppli~s. tripping the reactor. APPLICABILITY The RPS Matrix Logic, Trip Initiation Logic, and Manual Trip are required to be OPERABLE in MODES 1 and 2, and in MODES 3, 4, and s*when more than one full-length control rod capable of being withdrawn and the PCS boron concentration is less than REFUELING BORON CONCENTRATION. This ensures the reactor can be tripped. when necessary, but allows for maintenance and testing when the reactor trip is not needed . In MODES 3, 4, and 5 with no more than one full-length control rod capable of being withdrawn or the PCS boron concentration at REFUELING BORON CONCENTRATION, these Functions do not have to be OPERABLE. However, LCO 3.3.9, "Neutron Flux Monitoring Channels," does require'neutron flux monitoring capability under these conditions. , ACTIONS When the number of inoperable channels in a trip Function exceeds that specified in any related Condition associated with the same trip Function, then the plant is outside the safety analysis. Therefore, LCD 3.0.3 is immediately entered if applicable in the current MODE of operation .
Palisades Nuclear Plant B 3.3.2-6 05/30/99
RPS Logic and Trip Initiation B 3.3.2
BASES ACTIONS (continued)
A.1 Condition A applies if one Matrix Logic channel is .inoperable in any applicable MODE. The channel must be restored to OPERABLE status within r 48 hours. The Completion Time of 48 hours provides the operator time to take appropriate actions and still ensures that any risk involved in operating with a failed channel is acceptable. Operating experience has demonstrated that the probability of a random failure of a second Matrix Logic channel is low during any given 48 hour interval. If the channel cannot be restored to OPERABLE status within 48 hours, Condition E is entered.
Condition B applies if one Trip Initiation Logic channel is inoperable in any applicable MODE. The Required Action require de-energizing the affected clutch power supplies. This removes the need for the affected channel by performing its associated safety function. With the clutch power supplies associated with one initiation .logic channel de-energized, the remaining two clutch power supplies prevent control rod clutches* from de-energizing.* The remaining clutch power supplies are in a one-out-of-two logic with respect to the remairiing initiation
. logic channels in the clut~h power supply path. This meets redundancy requirements, but testing on the OPERABLE channels cannot be performed without causing a reactor trip.
Required Action B.1 provides for de-energizing the affected clutch power supplies associated with the inoperable channel within a Completion Time of J hour *
Palisades Nuclear Plant B 3.3.2-7 05/30/99
RPS Logic and Trip Initiation B 3.3.2
BASES ACTIONS (continued)
C.1 Condition C applies to the failure of one Manual Trip channel in any applicable MODE. With one manual reactor trip channel inoperable operation may continue until the reactor is shut down for other reasons. Repair during operation is not required because one OPERABLE channel is all *that is required for safe operation. No safety analyses assume operation of the Manual trip. The Manual Trip channels are not testable without actually causing a reactor trip, so even if the difficulty were corrected, the post maintenance testing necessary to declare the channel OPERABLE could not be completed during operation. Because of this, the Required Action is to restore the inoperable channel to OPERABLE status prior to entering MODE 2 from MODE 3 during the next plant startup. Condition D applies to the failure of both Trip Initiation Logic channels affecting the same trip leg. The affected control rod drive clutch power supplies must be de-energized immediately. With both channels inoperable, the RPS Function is lost if the affected clutch power supplies are not de-energized. Therefore, immediate action is required to de-energize the affected clutch power supplies. The immediate Completion Time is appropriate since there could be a loss of safety function if the associated clutch power supplies are not de-energized. Entry into LCO 3.0.3 is not an acceptable alternative in this condition. E.1. E.2.1 and E.2.2 Condition E is entered if Required Actions associated with Condition A, B, C, or D are not met within the required Completion Time or if for one or more Functions more than one Manual Trip, Matrix Logic, or Trip Initiation Logic channel is inoperable for reasons other than Condition D. In Condition E the reactor must be placed in a MODE in which the LCO does not apply. The Completion Time of 6 hours to be in MODE 3 is reasonable, based on operating experience, to reach the required MOD! from full power conditions in an orderly manner and without challenging plant systems. Palisades Nuclear Plant B 3.3.2-8 05/30/99
RPS Logic and Trip Initiation B 3.3.2
BASES ACTIONS E.1. E.2.1 and E.2.2 (continued)
Required Actions E.2.1 and E.2.2 allow 6 hours to verify that no more than one full-length control rod is capable of being withdrawn or to verify that PCS boron concentration is at REFUELING BORON CONCENTRATION. The Completion Time is reasonable to place the plant in an operating condition in which the LCO does not apply. SURVEILLANCE SR 3.3.2.1 REQUIREMENTS A CHANNEL FUNCTIONAL TEST on each RPS Logic channel is performed every 92 days to ensure the entire channel will perform its intended function when needed. A successful test of the required contact(s) of a channel relay may be performed by the verification of the change of state of a single contact of the relay. This clarifies what is an acceptable CHANNEL FUNCTIONAL TEST of a relay. This is acceptable because all of the other required contacts of the relay are verified by other Technical Specifications and non-Technical Specifications tests at least once per refueling interval with applicable extensions. This SR addresses the two tests associated with the RPS Logic: Matrix Logic and Trip Initiation Logic. Matrix Logic Tests These tests are performed one matrix at a time. They verify that a coincidence in the two input channels for each Function removes power from the matrix relays. During testing, power is applied to the matrix relay test ~oils and prevents the matrix relay contacts from assuming their de-energized state. The Matrix Logic tests will detect any short circuits around the bistable contacts in the coincidence logic such as may be caused by faulty bistable relay or trip channel bypass contacts .
Palisades Nuclear Plant B 3.3.2-9 05/30/99
RP~ Logic and Trip Initiation B 3.3.2
BASES SURVEILLANCE REQUIREMENTS SR 3.3.2.1 (continued)
Trip Initiation Logic Tests These tests are similar to the Matrix Logic tests, except that test power is withheld from one matrix relay at a time, allowing the initiation circuit to de-energize, de-energizing the affected set of clutch power supplies. The Frequency of 92 days is based on the reliability analysis presented in topical report CEN-327, "RPS/ESFAS Extended Test Interval Evaluation" (Ref. 5). SR 3.3.2.2 A CHANNEL *FUNCTIONAL TEST on the Manual Trip channels is performed prior to a reactor startup to ensure the entire channel will perform its intended function if required. A successful test of the required contact(s) of a channel relay may be performed by the verification of the change of state of a single contact of the relay. This clarifies what is an acceptable CHANNEL FUNCTIONAL TEST of a relay. This is acceptable because all of the other required contacts of the relay are verified by other Technical Specifications and non-Technical Specifications tests at least once per refueling interval with applicable extensions. The Manual Trip Function is not tested at power. However, the simplicity of this circuitry and the absence of drift concern makes this Frequency adequate. Additionally, operating experience has shown that these components usually pass the Surveillance when performed once within 7 days prior to each reactor startup. REFERENCES 1. 10 CFR 50, Appendix A
2. 10 CFR 100

3. FSAR, Figure 7-1

4. FSAR, Section 7.2

5. CEN-327, June 2, 1986, including Supplement 1, March 3, 1989
Palisades Nuclear Plant B 3.3.2-10 05/30/99
ESF Instrumentation B 3.3.3
B 3.3 INSTRUMENTATION B 3.3.3 Engineered Safety Features (ESF) Instrumentation BASES BACKGROUND The ESF Instrumentation initiates necessary safety systems, based upon the values of selected plant parameters, to protect against violating core design limits and the Primary Coolant System (PCS) pressure boundary and to mitigate accidents.
The ESF circuitry generates the signals listed below when the monitored variables reach levels that are indicative of conditions requiring protective action. The inputs to each ESF / actuation signal are also listed. *
1. Safety Injection Signal (SIS).

a. Containment High Pressure (CHP)

b. Pressurizer Low Pressure
2. Steam Generator Low Pressure (SGLP);
a. b. Steam Generator A Low Pressure Steam Generator B Low Pressure
3. Recirculation Actuation Signal (RAS);

a. Safety Injection Refueling Water Tank (SIRWT) Low Level

4. Auxiliary Feedwater Actuation Signal (AFAS);

a. Steam Generator A Low Level

b. Steam Generator B Low Level

5. Containment High Pressure Signal (CHP);

a. Containment High Pressure - Left Train

b. Containment High Pressure - Right Train

6. Containment High .Radiation Signal (CHR);

a. Containment High Radiation Palisades Nuclear Plant B 3.3.3-1 05/30/99
\
ESF Instrumentation B 3.3.3
BASES BACKGROUND 7. Automatic Bypass Removal (continued)
a. Pressurizer Pressure Low Bypass

b. Steam Generator A Low Pressure Bypass

c. Steam Generator B Low Pressure Bypass In the above list of actuation signals, the CHP and RAS are derived from pressure and level switches, respectively.
Equipment actuated by each of the above signals is identified in the FSAR, Chapter 7. (Ref. 1'). The ESF circuitry, with the exception of RAS, employs two-out-of-four logic. Four independent measurement channels are provided for each function used to generate ESF actuation signals. When any two channels of the same function reach their setpoint, actuating relays are energized which, in turn, initiate the protective actions. Two separate and redundant
trains of actuating relays, each powered from separate power suppltes, are utilized. These separate relay trains operate redundant trains of ESF equipment.
RAS logic consists of output contacts of the relays actuated by the SIRWT level switches arranged in a "one-out-of-two taken twice" logic. The contacts are arranged so that at least one low level signal powered from each station battery is required to initiate RAS. Loss of a single battery, therefore, cannot either cause or prevent RAS initiation. The ESF logic circuitry contains the capability to manually block the SIS actuation logic and the SGLP action logic during normal plant shutdowns to avoid undesired actuation of the associated equipment. In each case, when three of the four associated measurement channels are below the block setpoint, pressing a manual pushputton will block the actuation signal for that train. If two of the four of the measurement channels increase above the block setpoint, the block will automatically be removed. The sensor subsystems, including individual channel actuation bistables, is addressed in this LCO. *The actuation logic subsystems, manual a~tuation, and downstream components used to actuate the i ndi vi dual' ESF components are addressed in LCO 3.3.4 . Palisades Nuclear Plant 8 3.3.3-2 05/30/99
ESF Instrumentation B 3.3.3
BASES BACKGROUND Measurement Channels (continued)
Measurement channels, consisting of pressure switches, field transmitters, or process sensors and associated instrumentation, provide a measurable electronic signal based upon the physical characteristics of the parameter being measured. Four identical measurement channels are provided for each parameter used in the generation of trip signals. These are designated Channels A through D. Measurement channels provide input to ESF bi stables within the same ESF channel. In addition, some measurement channels may also be used as inputs to Reactor Protective System (RPS) bistables, and most provide indication in the control room.
When a channel monitoring a parameter indicates an abnormal condition, the bistable monitoring the parameter in that channel will trip. In the case of RAS and CHP, the sensors are latching auxiliary relays from level and pressure switches, respectively, which do not develop an analog input to separate bistables. Tripping two or more channels monitoring the same parameter will actuate both channels of Actuation Logic of the associated ESF equipment.
Three of the four measurement and bistable channels are necessary to meet the redundancy and testability of GDC 21 in Appendix A to 10 CFR 50 (Ref. 2). The fourth channel provides additional flexibility by allowing one channel to be removed from service for maintenance or testing while still maintaining a minimum two-out-of-three logic. Since no single failure will prevent a protective system actuation and no protective channel feeds a control channel, this arrangement meets the requirements of IEEE Standard 279 -1971 (Ref. 3) .
Palisades Nuclear Plant B 3.3.3-3 05/30/99
ESF Instrumentation B 3.3.3
BASES BACKGROUND Measurement Channels (continued)
The ESF Actuation Functions are generated by comparing a single measurement to a fixed bistable setpoint. The ESF Actuation Functions utilize the following input instrumentation:
Safety Injection Si gna 1 (SIS)
The Safety Injection Signal can be generated by any of three inputs: Pressurizer Low Pressure, Containment High Pressure, or Manual Actuation. Manual Actuation is addressed by LCO 3.3.4; Containment High Pressure is discussed below. Four instruments, channels A through D, monitor Pressurizer Pressure to develop the SIS actuation. Each of these instrument channels has two individually adjustable ESF bistable trip devices, one for the bypass remova~ citcuit (discussed below) and one for SIS. Each ESF bistable trip device actuates two auxiliary relays, one for each actuation train. The output contacts from these auxiliary relays form the
logic circuits addressed in LCD 3.3.4. The instrument channels associated with each Pressurizer Low Pressure SIS actuation bistable include the pressure measurement loop, the SIS actuation bistable, and the two auxiliary relays associated with that bistable. The bistables associated with automatic removal of the Pressurizet Low Pressure Bypass are discussed under Function 7.a, below.

Low Steam Generator Pressure Signal (SGLP)
There are two separate Low Steam Generator Pressure signals, one for each steam generator. For each steam generator, four instruments (channels A through D) monitor pressure to develop the SGLP actuation. Each of these instrument channels has two individually adjustable ESF bistable trip devices, one for the bypass removal circuit (discussed below) and one for SGLP. Each Steam SGLP bistable trip device actuates an auxiliary relay. The output contacts from these auxiliary relays form the SGLP logic circuits addressed in LCD 3.3.4. The instrument channels associated with each Steam Generator Low Pressure Signal bistable include the pressure measurement loop~ the SGLP actuation bistable, and the auxiliary relay associated with that bistable. The
bistables associated with automatic removal of the SGLP Bypass are discussed under Function 7.a, below.
Palisades Nuclear Plant B 3.3.3-4 D5/30/99
ESF Instrumentation B 3.3.3
- BASES BACKGROUND Measurement Channels (continued)

Reci rcul ati on Actuation Si gna 1 (RAS)
There are four Safety Injection Refueling Water (SIRW) Tank l~vel instruments used to develop the RAS signal. Each of these instrument channels actuates two auxiliary relays, one for each actuation train. The output contacts from these auxiliary relays form the logic circuits addressed in LCO 3.3.4. The SIRW Tank Low Level instrument channels associated with each RAS actuation bistable include the level instrument and the two auxiliary relays associated with that instrument.
Auxiliary Feedwater Actuation Signal (AFAS)
There are two separate AFAS signals (AFAS channels A and B), each one actuated on low level in either steam* generator. For each steam generator, four level i~strument~ (channels A through D) monitor level .to develop the AFAS actuation. signals. The output contacts from the bi stab 1es on the.se 1eve1 channe 1s form the SGLP logic circuits addresied in LCO 3.3.4. The in~trument channels associated with each Steam Generator Low Level Signal bistable include the level measurement loop and the Low Level AFAS bistable.
Containment High Pressure Actuation (CHP)
The Containment High Pressure signal fs actuated by two sets of four pressure switches, one set for each train. The output contacts from these pressure switches form the CHP logic circuits addressed in LCO 3.3.4. I .
Palisades Nuclear Plant B 3.3.3-5 05/30/99
ESF Instrumentation B 3.3.3
BASES BACKGROUND Measurement Channels (continued)

Containment High Radiation Actuation (CHR)
The Safety Injection Signal can be generated by either of two inputs: High Radiation or Manual Actuation. Manual Actuation is addressed by LCD 3.3.4. Four radiation monitor instruments, channels A through D, monitor containment area radiation level to develop the CHR signal. Each CHR monitor bistable device actuates one auxiliary relay which has contacts in each CHR logic train addressed in LCO 3.3.4. The instrument channels associated with each CHR actuation bistable include the radiation monitor itself and the associated auxiliary relay.
Automatic Bypass Removal Functions Pressurizer Low Pressure and Steam Generator Low Pressure logic circuits have the capability to be blocked to avoid undesired actuation when pressure is intentionally lowered during plant shutdowns. In each case these bypasses are automatically removed when the measured pressure exceeds the bypass permissive setpoint. The measurement channels which provide the bypass removal signal are the same channels which provide the actuation signal. Each of these pressure measurement channels has two bistables, one for actuation and one for the bypass removal Function. The pressurizer pressure channels include an auxiliary relay actuated by the bypass removal bistable. The logic circuits for Automatic Bypass Removal Functions are addressed by LCD 3.3.4.
Several measurement instrument channels provide more than one required function. Those sensors shared for RPS and ESF functions are identified in Table B 3.3.1-1. That table provides a listing of those shared channels and the Specifications which they affect .
Palisades Nuclear Plant B 3.3.3-6 05/30/99
ESF Instrumentation B 3.3.3
BASES BACKGROUND Bistable Trip Units (continued)
Where bistable trip units are used, they receive an analog input from the measurement device, compare the analog input to trip setpoints, and provide contact output to the Actuation Logic.
There are four channels of bistables, designated A through D, for each ESF Function, one for each measurement channel.
The Allowable Values are specified for each safety related ESF trip Function which is credited in the safety analys1s. Nominal trip setpoints are specified in the plant procedures. The nominal setpoints are selected to ensure plant parameters do not exceed the Allowable Value if the instrument loop is performing as required. The methodology used to determine the nominal trip setpoints is also provided in plant documents. Operation with a trip setpoint less conservative than the nominal trip setpoint, but within its Allowable Value, is acceptable. Each Allowable Value specified is more conservative than the analytical limit determined in the safety analysis in brder*to account for uncertainties ~ppropriate to the trip Functio.n. These uncertainties are ad.dressed as described in plant documents~ A channel is inoperable if its actual setpoint is not within its Allowable Value. Setpoints in.accordance with the Allowable Value will ensure that Safety Limits of Chapter 2.0, "SAFETY LIMITS (SLs), are 11 not violated during Anticipated Operational Occurrences (AOOs) and that the consequences of Design Basis Accidents (DBAs) will be acceptable, providing the plant is operated from within the LCOs at the onset of the AOO or OBA and the equip~ent functi6ns as designed. ' ESF Instrument Channel Bypasses The only ESF instrument channels with built in bypass capability are the Low SG Level AFAS bistables. Those bypasses are effected by a key operated switch, similar to the RPS Trip Channel Bypasses. A bypassed Low SG Level channel AFAS bistable cannot perform its specified function and must be considered inoperable. *
Palisades Nuclear Plant B 3.3.3-7 05/30/99
ESF Instrumentation B 3.3.3
BASES BACKGROUND ESF Instrument Channel Bypasses (continued)
While there are no other built-in provisions for instrument channel bypasses in the ESF design (bypassing any other channel output requires opening a circuit link, lifting a lead, or using a jumper), this LCD includes requirements for OPERABILITY of the instrument channels and bistables which provide input to the Automatic Bypass Removal Logic channels required by LCD 3.3.4, ESF Logic and Manual Initiation. 11 11 The Actuation Logic channels for Pressurizer Pressure and Steam Generator Low Pressure, however, have the ability to be manually bypassed when the associated pressure is below the range where automatic protection is required. These actuation logic channel bypasses may be manually initiated when three-out-of-four bypass permissive bistables indicate below their setpoint. When two-out-of-four of these bistables are above their bypass permissive setpoint, the actuation logic channel bypass is automatically removed. The bypass permissive bistables use the same four measurement channels as the blocked ESF function for their inputs . APPLICABLE Many of the analyzed accidents can be detected by one or more SAFETY ANALYSES ESF Functions. One of the ESF Functions is the primary actuation signal for that accident. An ESF Function* may be the primary actuation signal for more than one type of accident. An ESF Function may also be a secondary, or backup, actuation signal for one or more other accidents. Functions not specffically credited in the accident analysis, serve as backups and are part of the NRC approved licensing basis for the plant *
Palisades Nuclear Plant B 3.3.3-8 05/30/99
ESF Instrumentation B 3.3.3
BASES APPLICABLE ESF protective Functions are as follows.
SAFETY ANALYSES (continued) 1. Safety Injection Signal (SIS) The SIS ensures acceptable consequences during Loss of Coolant Accident (LOCA) events, including steam generator tube rupture, and Main Steam Line Breaks (MSLBs) or Feedwater Line Breaks (FWLBs) (inside containment). To provide the required protection, SIS is actuated by a CHP signal, or by two-out-of-four Pressurizer Low Pressure channels decreasing below the setpoint. SIS initiates the following actions:
a. Start HPSI &LPSI pumps;

b. Start component cooling water and service water pumps;

c. Initiate service water valve operations;
d* Initiate component cooling water valve operations;
e. Start containment cooling fans (when coincident with a loss of offsite power);

f. Enable Containment Spray Pump Start on CHP; and

g. Initiate Safety Injection Valve operations.
Each SIS logic train is also actuated by a contact pair on one of the CHP initiation relays for the associated CHP train.
2. Steam Generator Low Pressure Signal (SGLP)
The SGLP ensures acceptable consequences during an MSLB or FWLB by isolating the steam generator if it indicates a low steam generator pressure. The SGLP concurrent with or following a reactor trip, minimizes the rate of heat extraction and subsequent cooldown of the PCS during these events *
Palisades Nuclear Plant

B 3.3.3-9 05/30/99
ESF Instrumentatiori B 3.3.3
BASES APPLICABLE 2. Steam Generator Low Pressure Signal (SGLP) (continued)
SAFETY ANALYSES One SGLP circuit is provided for each SG. Each SGLP circuit is actuated by two-out-of-four pressure channels on the associated SG reaching their setpoint. SGLP initiates the following actions:
a. Close the associated Feedwater Regulating valve and its bypass; and

b. Close both Main Steam Isolation Valves.

3. Recirculation Actuation Signal At the end of the injection phase of a LOCA, the SIRWT will be nearly empty. Continued cooling must be provided by the ECCS to remove decay heat. The source of water for the ECCS pumps is automatically switched to the containment recirculation sump. Switchover from SIRWT to the containment sump must occur before the SIRWT empties to prevent damage to the ECCS pumps and a loss of core cooling capability. For similar reasons, switchover must not occur before there-is sufficient water in the containment sump to support pump suction.
Furthermore, early switchover must not occur to ensure sufficient borated water is injected from the SIRWT to ensure the reactor remains shut down in the recirculation mode. An SIRWT Low Level signal initiates the RAS. RAS initiates the following actions:
a. Trip LPSI pumps (this trip can be manually bypassed);

b. Switch HPSI and containment spray pump suction from SIRWT to Containment Sump by opening sump CVs and closing SIRWT CVs; and

c. Adjust cooling water to component cooling heat exchangers .
Palisades Nuclear Plant B 3.3.3-10 05/30/99
ESF Instrumentation B 3.3.3
BASES APPLICABLE 3. Recirculation Actuation Signal (continued)
SAFETY ANALYSES The RAS signal is actuated by separate sensors from those which provide tank level indication. The allowable range of 21" to 27" above the tank floor corresponds to 1~1% to 3.3% indicated level. Typically the actual setting is near the midpoint of the allowable range.
4. Auxiliary Feedwater Actuation Signal An AFAS initiates feedwater flow to both steam generators if a low level is indicated in either steam generator.
The AFAS maintains a steam generator heat sink during the following events:
MSLB;

FWLB;

5.
LOCA; and Loss of feedwater . Containment High Pressure Signal (CHP) The CHP signal closes all containment isolation valves not required for ESF operation and starts containment spray (if SIS enabled), ensuring acceptable consequences during LOCAs, control rod ejection events, MSLBs, or FWLBs (inside containment).
CHP is actuated by two-out-of-four pressure switches for the associated train reaching their setpoints. CHP initiates the following actions:
a. Containment Spray;

b. Safety Injection Signal;

c. Main Feedwater Isolation;

d. Main Steam Line Isolation;

e. Control Room HVAC Emergency Mode; and
Palisades Nuclear Plant
f. Containment Isolation Valve Closure .
B 3.3.3-11 05/30/99
ESF Instrumentation B 3.3.3
BASES APPLICABLE 6. Containment High Radiation Signal (CHR)
SAFETY ANALYSES (continued) CHR is actuated by two-out-of-four radiation monitors exceeding their setpoints. CHR initiates the following actions to ensure acceptable consequences following a LOCA or control rod ejection event:
a. Control Room HVAC Emergency Mode;

b. Containment Isolation Valve Closure; and

c. Block automatic starting of ECCS pump room sump pumps.
During refueling operations, separate switch-selectable radiation monitors initiate CHR, as addressed by LCO 3.3.6.
7. Automatic Bypass Removal Functions
The logic circuitry provides automatic removal of the Pressurizer Pressure Low and Steam Generator Pressure Low actuation signal bypasses. There are no assumptions in the safety analyses which assume operation of these automatic bypass removal circuits, and no analyzed events result ih conditions where the automatic removal would be required to mitigate the event. The automatic removal circuits are required to assure that logic circuit bypasses will not be overlooked during a plant startup.
The ESF Instrumentation satisfies Criterion 3 of 10 CFR 50.36(c)(2). LCD The LCD requires all channel components necessary to provide an ESF actuation to be OPERABLE. The Bases for the LCO on ESF Functions are addressed below.
1. Safety Iniection Signal (SIS)
This LCO requires four channels of SIS Pressurizer Low Pressure to be OPERABLE in MODES 1, 2, and 3 .
Palisades Nuclear Plant B 3~3.3-12 05/30/99
ESF Instrumentation B 3.3.3
BASES LCO 1. Safety Injection Signal (SIS) (continued)
The setpoint was chosen so as to be low enough to avoid actuation during plant operating transients, but to be high enough to be quickly actuated by a LOCA or MSLB. The settings include an uncertainty allowance which is consistent with the settings assumed in the MSLB analysis (which bounds the settings assumed in the LOCA analysis).
2. Steam Generator Low Pressure Signal (SGLP)
This LCO requires four channels of Steam Generator Low Pressure Instrumentation for each SG to be OPERABLE in MODES 1, 2, and 3. However, as indicated in Table 3.3.3-1, Note (a), the SGLP Function is not required to be OPERABLE in MODES 2 or 3 if all Main Steam Isolation Valves (MSIVs) are closed and deactivated and all Main Feedwater Regulating Valves (MFRVs) and MFRV bypass valves are either closed and deactivated or isolated by closed manual valves . The setpoint was chosen to be low enough to avoid actuation during plant operation, but be close enough to full power operating pressure to be actuated quickly in the event of a MSLB. The setting includes an uncertainty allowance which is consistent with the setting used in the Reference 4 analysis. Each SGLP logic is made up of output contacts from four pressure bistables from the*associated SG. When the logic circuit is satisfied, two relays are energized to actuate steam and feedwater line- isolation. This LCO applies to failures in the four sensor subsystems, including sensors, bistables, and associated equipment. Failures in the actuation subsystems are considered Actuation Logic failures and are addressed in LCO 3.3.4 .
Palisades Nuclear Plant B 3.3.3-13 05/30/99
ESF Instrumentation B 3.3.3
BASES LCO 3; Recirculation Actuation Signal (RAS)
(continued) This LCO requires four channels of SIRWT Low Level to be OPERABLE in MODES 1, 2, and 3. The setpoint was chosen to provide adequate water in the_ containment sump for HPSI pump net positive suction head following an accident, but prevent the pumps from running dry during the switchover. The upper limit on the Allowable Value for this trip is set low enough to ensure RAS does not initiate before sufficient water is transferred to the containment sump. Premature recirculation could impair the reactivity control Function of safety injection by limiting the amount of boron injection. Premature recirculation could also damage or disable the recirculation system if recirculation begins before the sump has enough water. J The lower limit on the SIRWT Low Level trip Allowable Value is high enough to transfer suction to the containment sump prior to emptying the SIRWT.
4. Auxiliary Feedwater Actuation Signal (AFAS)
The AFAS logic actuates AFW to a SG on a SG Low Level in that SG. The Allowable Value was chosen to assure that AFW flow would be initiated while the SG could still act as a heat sink and steam source, and to assure that a reactor trip would not .occur on low level without the actuation of AFW. This LCO requires four channels for each steam generator of Steam Generator Low Level to be OPERABLE in MODES 1, 2, and 3.
Palisades Nuclear Plant B 3.3.3-14 05/30/99
ESF Instrumentation B 3.3.3
BASES LCO 5. Containment High Pressure Signal (CHP)
(continued) This LCO requires four channels of CHP to be OPERABLE for each of the associated ESF trains (left and right) in MODES 1, 2, 3 and 4. The setpoint was chosen so as to be high enough to avoid actuation by containment temperature or atmospheric pressure changes, but low enough to be quickly actuated by a LOCA or a MSLB in the containment.
6. Containment High Radiation Signal (CHR)
This LCO requires four channels of CHR to be OPERABLE in MODES l, 2, 3, and 4. The setpoint is based on the maximum primary coolant leakage to the containment atmosphere allowed by
LCO 3.4.13 and 16the maximum activity allowed by LCO 3.4.16. N concentration reaches equilibrium in containment atmosphere due to its short half-life, but other activity was assumed to build up. At the end of a 24 hour leakage period the dose rate is approximately 20 R/h as seen by the area monitors. A large leak could cause the area dose rate to quickly exceed the 20 R/h setting and initiate CHR.
7. Automatic Bypass Removal The automatic bypass removal logic removes the bypasses which are used during plant shutdown periods, for Pressurizer Low Pressure and Steam Generator Low Pressure actuation signals.
The setpoints were chosen to be above the setpoint for the associated actuation signal, but well below the normal operating pressures. This LCO requires four channels of Pressurizer Low Pressure bypass removal and four channels for each steam generator of Steam Generator Low Pressure bypass removal, to be OPERABLE in MODES l, 2, and 3.
\
Palisades Nuclear Plant B 3.3.3-15 05/30/99
ESF Instrumentation B 3.3.3
BASES APPLICABILITY All ESF Functions are required to be OPERABLE in MODES l, 2, and 3. In addition, Containment High Pressure and Containment High Radiation are required to be operable in MODE 4.
In MODES l, 2, and 3 there is sufficient energy in the primary and secondary systems to warrant automatic ESF System responses to:
Close the main steam isolation valves to preclude a positive reactivity addition and containment overpressure;

Actuate AFW to preclude the loss of the steam generators as a heat sink (in the event the normal feedwater system is not available);

Actuate ESF systems to prevent or limit the release of fission product radioactivity to the environment by isolating containment and limiting the containment.
pressure from exceeding the containment design pressure during a design basis LOCA or MSLB; and
Actuate ESF systems to ensure sufficient borated inventory to permit adequate core cooling and reactivity control during a design basis LOCA or MSLB accident.
The CHP and CHR Functions are required to be OPERABLE in MODE 4 to limit leakage of radioactive material from containment and limit operator exposure during and following a OBA. The SGLP Function is not required to be OPERABLE in MODES 2 and 3, if all MSIVs are closed and deactivated and all MFRVs and MFRV bypass valves are either closed and deactivated or isolated by closed manual valves, since the SGLP Function is not required to perform any safety functions under these conditions. In lower MODES, automatic actuation of ESF Functions is not required, because adequate time is available for plant operators to evaluate plant conditions and respond by manually operating the ESF components, if required. LCO 3.3.6 addresses automatic Refueling CHR isolation during CORE ALTERATIONS or during movement of irradiated fuel .
Palisades Nuclear Plant B 3.3.3-16 05/30/99
ESF Instrumentation B 3.3.3
BASES APPLICABILITY In MODES 5 and 6, ESFAS initiated systems are either (continued) reconfigured or disabled for shutdown cooling operation.
Accidents in these MODES are slow to develop and would be mitig~ted by manual operation of individual components. ACTIONS The most common causes of channel i noperabil i ty *are outright failure qf loop components or drift of those loop components which is sufficient to exceed the tolerance provided in the plant setpoint analysis. Loop component failures are typically identified by the actuation of alarms due to the channel failing to the."safe" condition, during CHANNEL CHECKS (when the instrument is compared to the redundant channels), or during the CHANNEL FUNCTIONAL TEST (when an automatic component might *not respond properly). Typically, the drift of the ioop components is found to be small and results in ~delay of actuation rather than a total loss of function. Excessive loop component drift would, most likely, be identified during a CHANNEL CHECK (when the instrument is *compared to the redundant channels) or during a CHANNEL CALIBRATION (when instrument loop components are checked against reference standards). Typically, the dtift is small and results i~ ~ delay of
. actuation rather than a total loss of function. Determination of setpoint (;!rift is generally made during the performance of a CHANNEL FUNCTIONAL TEST when the process instrument is set up for adjustment to bring it to within specification. If the actual trip setpoint is not within the Allowable Value in Table 3.3.3-1, the channel is inoperable and the appropriate Condition(s) are entered.
In the event a channel's trip setpoint is found nonconservative with respect to the Allowable Value in Table 3.3.3-1, or the sensor, instrument loop, signal processing electronics, or*ESF bistable is found inoperable, then all affected Functions provided by that channel must be declared inoperable and the plant must enter the Condition statement for the particular protection Function affected .
Palisades Nuclear Plant B 3.3.3-17 05/30/99
ESF Instrumentatio~ B 3.3.3
BASES ACTIONS When the number of inoperable channels in a trip Function (continued) exceeds those specified in any related Condition associated with the same trip Function, then the plant is outside the safety analysis. Therefore, LCO 3.0.3 should be immediately entered if applicable in the current MODE of operation.
A Note has been added to clarify the application of the Completion Time rule~. The Conditions of this Specification may be entered independently for each Function in Table 3.3.3-1. Completion Times for the inoperable channel of a Function will be tracked separately. Condition A applies to the failure of a single bistable or associated instrumentation channel of one or more input parameters in each ESF Function except the RAS Function. Since the bistable and associated instrument channel combine to perform the actuation function, the Condition is also appropriate if both the bistable and associated instrument channel are inoperable. ESF coincidence logic is normally two-out-of-four. If one ESF channel is inoperable, startup or power operation is allowed to continue as long as action is taken to restore the design level of redundancy*. If one ESF channel is inoperable, startup or power operation is allowed to continue, providing the inoperable channel actuation bistable is placed in trip within 7 days. The provision of four trip channels allows one channel to be inoperable in a non-trip condition up to the 7 day Completion Time allotted to place the channel in trip. Operating with one failed channel in a non-trip condition during operations, places the ESF Actuation Logic in a two-out-of-three coincidence logic. If the failed channel cannot be restored to OPERABLE status in 7 days, the associated bistable is placed in a tripped condition. This places the function in a one-out-of-three configuration .
Palisades Nuclear Plant B 3.3.3-18 05/30/99
ESF Instrumentation B 3.3.3
BASES ACTIONS A.1 (continued)
In this configuration, common cause failure of the dependent channel cannot prevent ESF actuation. The 7 day Completion Time is based upon operating experience, which has demonstrated that a random failure of a second channel occurring during the 7 day period is a low probability event. Condition A is modified by a Note which indicates it is not applicable to the SIRWT Low Level Function. B.1 and B.2 Condition B applies to the failure of two channels in any of the ESF Functions except the RAS Function. With two inoperable channels, one channel actuation device must be placed in trip within the 8 hour Completion Time. Eight hours is allowed for this action since it must be accomplished by a circuit modification, or by removing power from a circuit
component. With one channel of protective instrumentation inoperable. the ESF Actuation Logic Function is in two-out-of-three logic, but with another channel inoperable the ESF may be operating with a two-out-of-two logic. This is outside the assumptions made in the analyses and should be corrected. To correct the problem, the second channel is placed in trip. This places the ESF in a one-out-of-two logic.
If any of the other OPERABLE channels receives a trip signal, ESF actuation will occur. One of the failed channels must be restored to OPERABLE status within 7 days, and the provisions of Condition A still applied to the remaining inoperable channel. Therefore, the channel that is still inoperable after completion of Required Action B.2 must be placed in trip if more than 7 days has elapsed since the channel 1 s initial failure. Condition B is modified by a Note which indicates that it is not applicable to the SIRWT Low Level Function. The Required Action is also modified by a Note stating that LCO 3.0.4 is not applicable. The Note was added to allow the changing of MODES even though two channels are inoperable, with one channel tripped. MODE changes in this configuration are allowed, to permit maintenance and testing on one of the inoperable channels. In this configuration, the protection system is in a one-out-of-two logic, and the probability of a common cause failure affecting both of the OPERABLE channels during the 7 days permitted is remote. Palisades Nuclear Plant B 3.3.3-19 05/30/99
ESF Instrumentation B 3.3.3
BASES ACTIONS C.l and C.2 (continued)
Condition C applies to one RAS SIRWT Low Level channel inoperable. The SIRWT low level circuitry is arranged in a "l-out-of-2 taken twice logic rather than the more frequently used 2-out-of-4 logic. Therefore, Required Action C.l differs from other ESF functions .. With a bypassed SIRWT low level channel, an additional failure might disable automatic RAS, but would not initiate a premature RAS. With a tripped channel, an additional failure could cause a premature RAS, but would not disable the automatic RAS. Since considerable time is available after initiation of SIS until RAS is required and there is quite a tolerance on the time when RAS must be initiated, and since a premature RAS could damage the ESF pumps, it is preferable to bypass an inoperable charinel and risk loss of automatic RAS than to trip a channel and risk a premature RAS. The Completion Time of 8 hours allowed is reasonable because
the Required Action involves a circuit modification .
Required Action C.2 requires that the inoperable channel be restored to OPERABLE status within 7 days. The Completion Time is reasonable based upon operating experience, which has demonstrated that a random failure of a second channel occurring during the 7 day period is a low probability event. D.l and D.2 If the Required Actions and associated Completion Times of Condition A, B, or C are not met for Functions 1, 2, 3, 4, or 7, the plant must be brought to a MODE in which the LCO does
not apply. To achieve this status, the plant must be brought to at least MODE 3 within 6 hours and to MODE 4 within 30 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems .

Palisades Nuclear Plant B 3.3.3-20 05/30/99
ESF Instrumentation B 3.3.3
BASES ACTIONS E.1 and E.2 (continued)
If the Required Actions and associated Completion Times of Condition A, B, or C are not met for Functions 5 or 6, the plant must be brought to a MODE in which the LCO does not apply., To achieve this status, the plant must be brought to at least MODE 3 within 6 hours and to MODE 5 within 36 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems. SURVEILLANCE The SRs for any particular ESF Function are found in the REQUIREMENTS SRs column of Table 3.3.3-1 for that Function. Most functions are subject to CHANNEL CHECK, CHANNEL FUNCTIONAL TEST, and CHANNEL CALIBRATION. While Palisades is not committed to performing all testing discussed in ANSI/IEEE Standard 338-1977, CHANNEL CHECKS, CHANNEL FUNCTIONAL TESTS, AND CHANNEL CALIBRATIONS are performed in accordance with the guidance of ANSI/IEEE Standard 338-1977, which is endorsed by Regulatory Guide 1.118. SR 3.3.3.1 A CHANNEL CHECK is performed once every 12 'hours on each ESF input channel which is provided with an indicator to provide a qualitative assurance that the channel is working properly and that its readings are within limits. A CHANNEL CHECK is not performed on the CHP and SIRWT Low Level channels because they have no associated control room indicator. Performance of the CHANNEL CHECK once every 12 hours ensures that a gross failure of instrumentation has not occurred. A CHANNEL CHECK is normally a comparison of the parameter indicated on one channel to a similar parameter on other channels. It is based on the assumption that instrument channels monitoring the same parameter should read approximately the same value. Significant deviations between instrument channels could be an indication of excessive instrument drift in one of the channels or of something even more serious. CHANN~~ CHECK will detect gross channel failure; thus, it is key to verifying the instrumentation continues to
operate properly between each CHANNEL CALIBRATION .
Palisades Nuclear Plant B 3.3.3-21 05/30/99
ESF Instrumentation B 3.3.3 BASES SURVEILLANCE SR 3.3.3.1 (continued) REQUIREMENTS Agreement criteria are determined by the plant staff based on a combination of the channel instrument uncertainties, including indication and readability. If a channel is outside the criteria, it may be an indication that the sensor or the signal processing equipment has drifted outside its limit. If the channels are within the criteria, it is an indication that the channels are OPERABLE. If the channels are normally off ~cale during times when Surveillance is required, the CHANNEL CHECK will only verify that they are off scale in the same direction. Offscale low current loop channels are verified to be reading at the bottom of the range and not failed downscale. The Frequency of about once every shift is based on operating experience that demonstrates channel failure is rare. Since the probability of two random failures in redundant channels in any 12 hour period is extremely low, the CHANNEL CHECK minimizes the chance of loss of protective function due to failure of redundant channels. The CHANNEL CHECK supplements
less formal, but more frequent, checks of CHANNEL OPERABILITY during normal operational use of displays associated with the LCO required channels.
*sR 3.3.3.2
A CHANNEL FUNCTIONAL TEST is performed every 92 days to ensure the entire channel will perform its intended function when needed. A successful test of the required contact(s) of a channel relay may be performed by the verification of the change of state of a single contact of the relay. This clarifies what is an acceptable CHANNEL FUNCTIONAL TEST of a relay. This is acceptable because all of the other required contacts of the relay are verified by other Technical Specifications and non-Technical Specifications tests at least once per refueling interval with applicable extensions.
This test is required to be performed each 92 days on ESF input channels provided with on-line testing capability. It is not required for the SIRWT Low Level channels since they have no built in test capability. The CHANNEL FUNCTIONAL TEST for
SIRWT Low Level channels is performed each 18 months as part of the required CHANNEL CALIBRATION.
The CHANNEL FUNCTIONAL TEST tests the individual channel~ using an analog test input to each bistable. Palisades Nuclear Plant B 3.3.3-22 05/30/99
ESF Instrumentation B 3.3.3
BASES SURVEILLANCE SR 3.3.3.2 (continued)
REQUIREMENTS ' Any setpoint adjustment shall be consistent with the assumptions of the current setpoint analysis .. The Frequency of 92 days is based on the reliability analysis presented in topical report CEN-327, "RPS/ESFAS Extended Test Interval Evaluation" (Reference 5). SR 3.3.3.3 CHANNEL CALIBRATION is a complete check of the instr~ment channel, including the sensor. The Surveillance verifies that the channel responds to a measured parameter within the necessary range and accuracy. CHANNEL CALIBRATION leaves the channel adjusted to account for instrument drift between suctessive calibrations to ensure that the channel remains operational between successive surveillances. CHANNEL CALIBRATIONS must be performed consistent with the setpoint analysis .
The as found and as left values must also be recorded and reviewed for consistency with the assumptions of the extension analysis. The requirements for this review are outlined in Reference 5.
The Frequency is based upon the assumption of an 18 month ca 1i brati on i nterva 1 for the determination of the magnitude .of equipment drift in the setpoint analysis. REFERENCES 1. FSAR, Chapter 7
2. 10 CFR 50, Appendix A

3. IEEE Standard 279-1971

4. FSAR, Chapter 14

5. CEN-327, June 2, 1986, including Supplement 1, March 3, 1989
\ '
Palisades Nuclear Plant B 3.3.3-23 05/30/99
ESF Logic and Manual Initiation B 3.3.4
B 3.3 INSTRUMENTATION B 3.3.4 Engineered Safety Features (ESF) Logic and Manual Initiation BASES BACKGROUND The ESF Instrumentation initiates necessary safety systems, based upon the values of selected plant parameters, to protect against violating core design limits and the Primary Coolant System (PCS) pressure boundary and to mitigate accidents.
The ESF circuitry generates the following signals listed below when the monitored variables reach levels that are indicative of conditions requiring protective action. The inputs to each ESF Actuation Signal are also listed.
1. Safety Injection Signal (SIS);

a. Containment High Pressure (CHP)

b. Pressurizer Low Pressure
2. Steam Generator Low Pressure Signal (SGLP) a.
b. Steam Generator A Low Pressure Steam Generator B Low Pressure
3. Recirculation Actuation Signal (RAS);

a. Safety Injection Refueling Water Tank (SIRWT) Low Level

4. Auxiliary Feedwater Actuation Signal (AFAS)

a. Steam Generator A Low Level

b. Steam Generator B Low Level

5. Containment High Pressure Signal (CHP);

a. Containment High Pressure - Left Train

b. Containment High Pressure - Right Train
Palisades Nuclear Plant I
B 3.3.4-1 05/30/99
ESF Logic and Manual Initiation B 3.3.4
BASES BACKGROUND (continued)
6. Containment High Radiation Signal (CHR)

a. Containment High Radiation In the above list of actuation signals, the CHP and RAS are derived from pressure and level switches, respectively.
Equipment actuated by each of the above signals is identified in the FSAR, Chapter 7 (Ref. 1). The ESF circuitry, with the exception of RAS, employs two-out-of-four logic. Four independent measurement channels are provided for each function used to generate ESF actuation signals. When any two channels of the same function reach their setpoint, actuating relays initiate the protective actions. Two separate and redundant trains of actuating* relays, each powered from separate power supplies, are utilized. These separate relay trains operate redundant trains of ESF equipment. The actuation relays are. considered part of the actuation logic addressed by this LCO. RAS logic consists of output contacts of the relays actuated by the SIRWT Low Level switches arranged in a "one-out-of-two taken twice" logic. The contacts are arranged so that at least one low level signal powered from each station battery is required to initiate RAS. Loss of a single battery, therefore, cannot either cause or prevent RAS initiation.
The sensor subsystem, including individual channel bistables, is addressed in LCO 3.3.3, "Engineered Safety Features (ESF)
Instrumentation." This LCO addresses the actuation subsystem manual actuation, and downstream components used to actuate the individual ESF functions, as defined in the following section. ESF Logic Each of the six ESF actuation signals in Table 3.3.4-1 operates two trains of actuating relays. Each train is capable of initiating the ESF equipment to meet the minimum requirements to provide all functions necessary to operate the system associated with the plant's capability to cope with abnormal events .
Palisades Nuclear Plant B 3.3.4-2 05/30/99
ESF Logic and Manual Initiation B 3.3.4
BASES BACKGROUND ESF Logic (continued)
The SGLP logic circuitry includes provisions such that the SGLP automatic actuation Function may be bypassed if three-out-of-four Steam Generator (SG) pressure channels are below a bypass permissive setpoint. Similarly, the SIS automatic actuation on Pressurizer Low Pressure may be bypassed when three-out-of~four channels are below a permissive setpoint. This actuation bypassing is performed when the ESF Functions are no longer required for protection. These actuation bypasss are enabled manually when the permissive conditions are satisfied . All actuation bypasss are automatically removed when enabling conditions.are no longer satisfied. If an SIS or SGLP automatic actuation channel is bypassed, other than as a~lowed by Table 3.3.4-1, the channel cannot perform its required
~afety function and must be considered to be i~operable.
Testing of a major portion of the ESF circuits is accomplished. while the plant is at power. More extensive sequencer and load testing may be done with the reactor shut down. The test
circuits are designed to test the redundant circuits separately such that the correct operation of each circuit may be verified by either equipment operation or by_sequence lights.
Manual Ini ti ati on Manual ESF initiation capability is provided to permit the operator to manually actuate an ESF System when necessary. Two control room mounted manual actuation switches are provided for SIS actuation, one for each train. Each SIS manual actuation switch affects one actuation channel, which actuates one train of SIS equipment. There are no single manual controls provided to actuate CHP, however, CHP may be manually initiated using individual component controls. Two control room mounted manual actuation switches are provided for CHR actuation, each switch affects both actuation channels, which actuates both CHR trains. There are no single m~nual controls provided to actuate SGLP, however, SGLP may be manually initiated using individual component controls . Palisades Nuclear Plant B 3.3.4-3 05/30/99
ESF Logic and Manual Initiation B 3.3.4
BASES BACKGROUND (continued)
Manual Initiation There are no single manual controls provided to actuate RAS, however, RAS may be manually initiated using individual component controls. Manual actuation of AFW may be accomplished through pushbutton actuation of each AFAS channel or by use of individual pump and valve controls. Each automatic AFAS actuation channel starts the AFW pumps in their starting sequence (if P-8A fails to start, a P-8C start signal is generated, and if P-8C also fails to start, a P-8B start signal is generated) and opens the associated flow control valves. APPLICABLE Many of the analyzed accidents can be detected by one or more SAFETY ANALYSES ESF Functions. One of the ESF Functions is the primary actuation signal for that accident. An ESF Function may be the primary actuation signal for more than one type of accident. An ESF Function may also be a secondary, or backup, actuation
signal for one or more other accidents. Functions such as Manoal Initiation, not specifically credited in the accident analysis, serve as backups to Functions and are part of the NRC staff approved licensing basis for the plant.
The manual initiation is not required by the accident analysis. The ESF logic must function in all situations where the ESF . function is required (as discussed in the Bases for LCO 3.3.3}. The ESF satisfies Criterion 3 of 10 CFR 50.36(c)(2). LCO The LCO requires that all components necessary to provide an ESF actuation be OPERABLE. The Bases for the LCO on ESF automatic actuation Functions are addressed in LCO 3.3.3. Those associated with the Manual Initiatitin or Actuation Logic are addressed below. ESF Logic and Manual Initiation Functions are required to be OPERABLE in MODES 1; 2, and 3, or in MODES .1, 2, 3, and 4, as appropriate, when the associated automatic initiation channels addressed by LCO 3.3.3 are required.
\
Palisades Nuclear Plant B 3.3.4-4 05/30/99
ESF Logic and Manual Initiation B 3.3.4
BASES LCO (continued)
1. Safety Injection Signal (SIS)
SIS is actuated by manual initiation, by a CHP signal, or by two-out-of-four Pressurizer Low Pressure channels decreasing below the setpoint. Each Manual Initiation channel consists of one pushbutton which directly starts the SIS actuation logic for the associated train. Each SIS logic train is also actuated by a contact pair on one of the CHP initiation relays for the associated CHP train.
a. Manual Initiation This LCO requires two channels of SIS Manual Initiation to be OPERABLE.

b. Actuation Logic This LCO requires two channels of SIS Actuation Logic to be OPERABLE. Failures in the actuation subsystems are addressed in this LCO .
c. CHP Logic Trains The CHP initiation relay (5P-x) input to the SIS logic is considered part of the SIS logic. Two channels, one per SIS train, must be OPERABLE.
d. Automatic Bypass Removal This LCO requires twQ channels of the automatic bypass removal logic for SIS P~essurizer Low Pressure to be OPERABLE in MODES 1, 2, and 3. If an SIS automatic actuation channel is bypassed, other than as allowed by Table 3.3.4-1, the channel cannot perform its required safety function and must be considered to be inoperable.
The Pressurizer Low Pressure logic train for each SIS train can be bypassed when three-out-of-four channels indicate below 1700 psia. This bypass prevents undesired actuation of SIS during a normal plant cooldown. The bypass signal is automatically removed when two-out-of-four channels exceed the setpoint, in accordance with the philosophy of removing bypasses when the enabling conditions are no longer satisfied .

Palisades Nuclear Plant B 3.3.4-5 05/30/99
ESF Logic and Manual Initiation B 3.3.4
BASES LCD 1. Safety Injection Signal (SIS) (continued)
The bypass permissive is set low enough so as not to be enabled during normal plant operation, but high enough to allow bypassing prior to reaching the trip setpoint.
2. Steam Generator Low Pressure Signal (SGLP)

a. Manual Initiation This LCD requires two channels of SGLP Manual Initiation to be OPERABLE. There is no manual control which actuates the SGLP logic circuits. The actuated components must be individually actuated using control room manual controls.

b. Actuation Logic This LCO requires two channels of SGLP Actuation
c.
Logic to be OPERABLE, one for each SG . Automatic Bypass Removal This LCO requires two channels, one for each SG, of the SGLP automatic bypass removal logic to be OPERABLE in MODES 1, 2, and 3. If an SIS automatic actuation channel is bypassed, other than as allowed by Table 3.3.4-1, the channel cannot perform its required safety function and must be considered to be inoperable. The SGLP from each SG may be bypassed when three-out-of-four channels indicate below 565 psia. This bypass prevents undesired actuation during a normal plant cooldown. The bypass signal is automatically removed when two-out-of-four channels exceed the setpoint, in accordance with the philosophy of removing bypasses when the enabling conditions are no longer satisfied. The bypass permissive is set low enough so as not to be enabled during normal plant operation, but high enough to ~llow bypassing prior to reaching the trip set point. *
Palisades Nuclear Plant B 3.3.4-6 05/30/99
ESF Logic and Manual Initiation B 3.3.4
BASES LCO (continued)
3. Recirculation Actuation Signal (RAS)

a. Manual Initiation This LCO requires two channels of RAS Manual Initiation to be OPERABLE. There is no manual control of which actuates the RAS logic circuits.
The actuated components must be individually actuated using manual controls.
b. Actuation Logic This LCO requires two channels of RAS Actuation Logic to be OPERABLE. /

4. Auxiliary Feedwater Actuation Signal (AFAS)

a. Manual Ini ti ati on This LCO requires two channels of AFAS Manual
Initiation to be OPERABLE. Each train of AFAS may be manually initiated with either of two sets of controls. Only one set of manual controls is required to be OPERABLE for each AFW train. One set of controls are the pushbuttons provided to actuate each train on the C-11 panel; the other set of controls are those manual controls provided on C-01 for each AFW pump and flow control valve.
b. Actuation Logic This LCD requires two channels of AFAS Actuation Logic to be OPERABLE.

5. Containment High Pressure Signal (CHP)

a. Manua 1 Ini ti ati on This LCD requires the manual controls necessary to actuate those valves and components actuated by an automatic CHP to be OPERABLE.

b. Actuation L.oqi c This LCO requires two channels of CHP Actuation Logic to be OPERABLE.
Palisades Nuclear Plant B 3.3.4-7 D5/30/99
ESF Logic and Manual Initiation B 3.3.4

BASES LCO (continued)
6. Containment High Radiation Signal (CHR)

a. Manual Initiation This LCO requires two channels of CHR Manual Initiation to be OPERABLE. Pushbuttons are available for manual actuation of *each CHR logic train.

b. Actuation Logic This LCO requires two channels of CHR Actuation Logic to be OPERABLE.
APPLICABILITY ESF Functions are required to be OPERABLE in MODES 1, 2, and 3 or MODES l, 2, 3, and 4 as specified in Table 3.3.4-1. In MODES 1, 2, and 3, there is sufficient energy in the primary and secondary systems to warrant automatic ESF System responses I. to:
.Close the MSIVs to preclude a positive reactivity addition and containment overpressure; Actuate AFW to preclude the loss of the steam generators as a heat sink (in the event the normal feedwater system is not available);
Actuate ESF systems to prevent or limit the release of fission product radioactivity to the environment by isolating containment and limiting the containment pressure from exceeding the containment design pressure during a design basis LOCA or MSLB; and

Actuate ESF systems to ensure sufficient borated inventory to permit adequate core cooling and reactivity control during a design basi.s LOCA or MSLB accident.
The CHP and CHR Functions are also required to be OPERABLE in MODE 4 to limit leakage of radioactive material from containment and limit operator exposure during and following a OBA .
Palisades Nuclear Plant B 3.3.4-8 05/30/99
ESF Logic and Manual Initiation B 3.3.4
BASES APPLICABILITY (continued)
The SGLP Function is not required to be OPERABLE in MODES 2 and 3, if all MSIVs are closed and deactivated and all MFRVs and MFRV bypass valves are either closed and deactivated or isolated by closed manual valves, since the SGLP Function is not required to perform any safety function under these conditions. In MODES 5 and 6, automatic actuation of ESF Functions is not required, because adequate time is available for plant operators to evaluate plant conditions and respond by manually operating the ESF components if required. In these MODES, ESF initiated systems are either reconfigured or disabled for shutdown cooling operation. Accidents in these MODES are slow to develop and would be mitigated by manual operation of individual components. ACTIONS When the number of inoperable channels in a trip Function exceeds those specified in any related Condition associated with the same trip Function, then the plant is outside the safety analysis. Therefore, LCO 3.0.3 should be immediately entered, if applicable in the current MODE of operation. A Note has been added to the ACTIONS to clarify the application of the Completion Time rules. The Conditions of this Specification may be entered independently for each Function in Table 3.3.4-1 in the LCO. Completion Times for the inoperable channel of a Function will be tracked separately. Condition A applies to one Manual Initiation, Bypass Removal, or Actuation Logic channel inoperable. The channel must be restored to OPERABLE status to restore redundancy of the ESF Function. The 48 hour Completion Time is commensurate with the importance of avoiding the vulnerability of a single failure in the only remaining OPERABLE channel.
\ .
Palisades Nuclear Plant B 3.3.4-9 05/30/99
ESF Logic and Manual Initiation B 3.3.4
BASES ACTIONS (continued)
B.1 and B.2 If two Manual Initiation, Bypass Removal, or Actuation Logic channels inoperable for Functions l, 2, 3, or 4, or if the Required Action and associated Completion Time of Condition A cannot be met for Function 1, 2, 3, or 4, the reactor must be brought to a MObE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 6 hours and to MODE 4 within 30 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems .
.C.1 and C.2 Condition C is entered when one or more Functions have two Manual Initiation or Actuation Logic channels inoperable for Functions 5 or 6, and when the Required Action and associated Completion Time of Condition A are not met for Functions 5 or
6. If Required Action A.l cannot be met within the required Completion Time, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 6 hours and to MODE 5 within 36 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.
SURVEILLANCE SR 3.3.4.1 REQUIREMENTS A functional test of each SIS actuation channel must be performed each 92 days. This test is to be performed using the installed control room test switches and test circuits for both "with standby power" and "without standby power". When testing the "with standby power" circuits, proper operation of the SIS-X relays must be verified; when testing the "without standby power" circuits, proper operation of the OBA sequencer and the associated logic circuit must be verified. The test circuits are designed to block those SIS functions, such as injection of concentrated boric acid, which would interfere with plant
operation.
The Frequency of 92 days is a feature of the initial Palisades
license .
Palisades Nuclear Plant B 3.3.4-10 05/30/99
ESF Logic and Manual Initiation B 3.3.4
BASES SURVEILLANCE REQUIREMENTS SR 3.3.4.2 (continued) A CHANNEL FUNCTIONAL TEST of each AFAS Actuation Logic Channel is performed every 92 days to ensure the channel will perform its intended function when needed. A successful test of the required contact(s) of a channel relay may be performed by the verification of the change of state of a single contact of the rel~y. This clarifies what is an acceptable CHA~NEL FUNCTIONAL TEST of a relay. This is acceptable because all of the other required contacts of the relay are verified by other Technical Specifications and non-Technical Specifications tests at least once per refueling interval with applicable extensions.
Instrumentation channel tests.are addressed in LCO 3.3.3. SR 3.3.4.2 addresses Actuation Logic tests of the AFAS using the installed test circuits. The Frequency of 92 days for SR 3.3.4.2 is in agreement with the conclusions of the reliability analysis presented in topical report CEN-327, "RPS/ESFAS Extended Test Interval
Evaluation" (Ref. 2) .
SR 3.3.4.3 A CHANNEL FUNCTIONAL TEST is performed on the manual ESF actuation channels, bypass removal channels, and Actuation Logic channels for certain ESF Functions, providing actuation of the Function. A successful test of the required, contact(s) of a channel relay may be performed by the verification of the change of state of a single contact of the relay. This clarifies what is an acceptable CHANNEL FUNCTIONAL TEST of a re 1ay. This is acceptab 1e because a11 of the other required contacts of the relay are verifie9 by other Technical Specifications and non-Technical Specifications tests at least once per refueling interval with applicable extensions. This Surveillance verifies that the Manual Initiation Functions are OPERABLE. This Surveillance also verifies that the entire channel of the Manual Actuation Logic will perform its intended I I i function when needed .
Palisades Nuclear Plant B 3.3.4-11 05/30/99
ESF Logic and Manual Initiation B 3.3.4
BASES SURVEILLANCE REQUIREMENTS SR 3.3.4.3 (continued)
The 18 month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power. Operating experience has shown these components usually pass the Surveillance when performed at a Frequency of once every 18 months. REFERENCES 1. FSAR, Chapter 7
2. CEN-327, June 2, 1986, including Supplement 1, March 3, 1989
Palisades Nuclear Plant B 3.3.4-12 05/30/99
DG - UV Start B 3.3.5
B 3.3 INSTRUMENTATION B 3.3.5 Diesel Generator (DG) - Undervoltage Start (UV Start)
BASES BACKGROUND The DGs provide a source of emergency power when offsite power is either unavailable or insufficiently stable to allow safe plant operation. Undervoltage protection will generate a UV Start in the event a Loss of Voltage or Degraded Voltage condition occurs. There are two UV Start Functions for each 2.4 kV vital bus. Undervoltage protection and load shedding features for safety-related buses at the 2,400 V and lower voltage levels are designed in accordance with 10 CFR 50, Appendix A, General Design Criterion 17 (Ref. 1) and the following features:
1. Two levels of automatic undervoltage protection from loss or degradation of offsite power sources are provided.
The first level (loss of voltage) provides normal loss of voltage protection. The second level of protection (degraded voltage) has voltage and time delay set points selected for automatic trip of the offsite sources to protect safety-related equipment from sustained degraded voltage conditions at all bus voltage levels. Coincidence logic is provided to preclude spurious trips.
2. The undervoltage protection system automatically prevents load shedding of the safety-related buses when the emergency generators are supplying power to the safeguards loads.

3. Control circuits for shedding of Class lE and non-Class lE loads during a Loss of Coolant Accident (LOCA) themselves are Class lE or are separated electrically from the Class lE portions~
Palisades Nuclear Plant B 3.3.5-1 05/30/99
DG - UV Start B 3.3.5
BASES BACKGROUND (continued)
Description Each 2,400 V Bus (IC and 10) is equipped with two levels of undervoltage protection relays (Ref. 2). The first level (Loss of Voltage Function) relays 127-1 and 127-2 are set at approximately 77% of rated voltage with an inverse time relay. One of these relays measures voltage on each of the three phases. They protect against sudden loss of voltage as sensed on the corresponding bus using a three-out-of-three coincidence logic. The actuation of the associated auxiliary relays will trip the associated bus incoming circuit breakers, start its associated DG, initiate bus load shedding, and activate annunciators in the control room. The DG circuit breaker is closed automatically upon establishment of satisfactory voltage and frequency by the use of associated voltage sensing relay 1270-1 or 1270-2. The second level of undervoltage protection (Degraded Voltage Function) relays 127-7 and 127-8 are set at approximately 93% of rated voltage, with one relay monitoring each of the three phases. These relays protect against sustained degraded
voltage conditions on the corresponding_bus using a three-out-of-three coincidence logic. These relays have a built-in 0.65 second time delay, after which the associated DG receives a start signal and annunciators in the control room are actuated. If a bus undervoltage exists after an additional six seconds,.the associated bus incoming circuit breakers will be tripped and a bus load shed will be initiated.
Trip Setpoints The trip setpoints are based on the analytical limits presented in References 3 and 4, and justified in Reference 5. The selection of these trip setpoints is such that adequate protection is provided when all sensor and processing time delays are taken into account. To allow for calibration tolerances, instrumentation uncertainties, and instrument drift, setpoints specified in SR 3.3.5.2 are conservatively adjusted with respect to the analytical limits. A ~etailed analysis of the degraded voltage protection is provided in References 3 and 4. Palisades Nuclear Plant B 3.3.5-2 05/30/99
DG - UV Start B 3.3.5
BASES BACKGROUND Trip Setpoints (continued)
The specified setpoints will ensure that the consequences of accidents will ~acceptable, providing the plant is operated from within the LCOs at the onset of the accident and the equipment functions as designed. APPLICABLE The DG - UV Start is required for Engineered Safety Features SAFETY ANALYSES (ESF) systems to function in any accident with a loss.of offsite power. Its design basis is that of the ESF Systems. Accident analyses credit the loading of the DG based on a loss of offsite power during a LOCA. The diesel loading has been included in the delay time associated with each safety system component requiring DG supplied power following a loss of offsite power. This delay time includes contributions from the DG start, DG loading, and Safety Injection System component actuation. The required channels of UV Start, in conjunction with the ESF systems powered from the DGs, provide plant protection in the event of any of the analyzed accidents discussed in
Reference 6, in which a loss of offsite power is assumed.
UV Start channels are required to meet the redundancy and testability requirements of GDC 21 in 10 CFR 50, Appendix A* (Ref. 1). The delay times assumed in the safety analysis for the ESF equipment include the 10 second DG start delay and the appropriate sequencing delay, if applicable. The response times for ESFAS actuated equipment include the appropriate DG loading and sequencing delay. The DG - UV Start channels satisfy Criterion 3 of 10 CFR 50.36(c)(2). . The LCO for the DG - UV Start requires that three channels per bus of each UV Start instrumentation Function be OPERABLE when the associated DG is required to be OPERABLE. The UV Start supports safety systems associated with ESF actuation. l ! Palisades Nuclear Plant B 3.3.5-3 05/30/99
DG - UV Start B 3.3.5
BASES LCO The Bases for the trip setpoints are as follows:
The voltage trip setpoint is set low enough such that spurious trips of the offsite source due to operation of the undervoltage relays are not expected for any combination of plant loads and normal grid voltages. This setpoint at the 2,400 V bus and reflected down to the 480 V buses has been verified through an analysis to be greater than the minimum allowable motor voltage (90% of nominal voltage). Motors are the most limiting equipment in the system. MCC contactor pickup and drop-out voltage is also adequate at the setpoint values. The analysis ensures that the distribution system is capable of starting and operating all safety-related equipment within the equipment voltage rating at the allowed source voltages. The power distribution system model used in the analysis has been verified by actual testing (Refs. 5 and 7). *
The time delays involved will not cause any thermal damage as the setpoints are within voltage ranges for sustained

operation. They are long enough to preclude trip of the offsite source caused by the starting of large motors and yet do 1not exceed the time limits of ESF actuation .assumed in FSAR Chapter 14 (Ref. 6) and validated by Reference 8.
Calibration of the undervoltage relays verify that the time delay is sufficient to avoid spurious trips. APPLICABILITY The DG - UV Start actuation Function is required to be OPERABLE whenever the associated DG is required to be OPERABLE per LCO 3.8.1, "AC Sources - Operating," or LCO 3.8.2, "AC Sources
- Shutdown," so that it can perform its function on a loss of power or degraded power to the vital bus.
ACTIONS A DG - UV Start channel is inoperable when it does not satisfy the OPERABILITY criteria for the channel *s Function. In the event a channel s trip setpoint is found nonconservative 1 with respect to the Specified Setpoint, or the channel is found inoperable, then all ~ffected Functions provided by that channel must be decla~ed inoperable and the LCO_Condition entered. The required channels are specified on a per DG basis. Palisades Nuclear Plant B 3.3.5-4 05/30/99
DG - UV Start B 3.3.5 BASES ACTIONS A.1 (continued) Condition A applies if one or more of the three phase UV sensors or relay logic is inoperable for one or more Functions (Degraded Voltage or Loss of Voltage) per DG bus. The affected DG must be declared inoperable and the appropriate Condition(s) entered. Because of the three-out-of-three logic in both the Loss of Voltage and Degraded Voltage Functions, the appropriate means of addressing channel failure is declaring the DG inoperable, and effecting repair in a *manner consistent with other DG failures. Required Action A.1 ensures that Required Actions for the affected DG inoperabilities are initiated. Depending upon plant MODE, the actions specified in LCO 3.8.1 or LCO 3.8.2, as applicable, are required immediately. SURVEILLANCE SR* 3.3.5.1
REQUIREMENTS A CHANNEL FUNCTIONAL TEST is performed on each UV Start logic channel every 18 months to ensure that the logic channel will perform its intended function when needed. The Under Voltage sensing relays are tested by SR 3.3.5.2. A successful test of the required contact(s) of a channel relay may be performed by th~ verification of the change of state of a single contact of the relay. This clarifies what is an acceptable CHANNEL FUNCTIONAL TEST of a relay. This is acceptable because all of the other required contacts of the relay are verified by other Technical Specifications and non-Technical Specifications tests
. at least once per refueling interval with applicable
extensions.
The Frequency of 18 months is based on the plant conditions necessary to perform the test .
Palisades Nuclear Plant, B 3.3.5-5 05/30/99
OG - UV Start B 3.3.5
BASES SURVEILLANCE REQUIREMENTS SR 3.3.5.2 (continued) A CHANNEL CALIBRATION performed each 18 months verifies the accuracy of each component within the instrument channel. This includes calibration of the undervoltage relays and demonstrates that the equipment falls within the specified operating characteristics defined by the manufacturer.
The Surveillance verifies that the channel responds to a measured parame~er within the necessary range and accuracy. CHANNEL CALIBRATION leaves the channel adjusted to account for instrument drift between successive calibrations to ensure that the channel remains operational between successive tests. CHANNEL CALIBRATIONS must be performed consistent with the setpoint analysis. The Frequency of 18 months is based on the plant conditions necessary to perform the test.
REFERENCES 1.
2. 3. 10 CFR 50, Appendix A GDCs 17 and 21 FSAR, Section 8.6 CPCo Analysis EA-ELEC-VOLT-033
4. CPCo Analysis EA-ELEC-VOLT-034

5. CPCo Analysis EA-ELEC-VOLT-17

6. FSAR, Chapter 14

7. CPCo Analysis EA-ELEC-VOLT-13

8. CPCo Analysis A-NL-92-111
Palisades Nuclear Plant B 3.3.5-6 05/30/99
Refueling CHR Instrumentation B 3.3.6
B 3.3 INSTRUMENTATION B 3.3.6 Refueling Containment High Radiation (CHR) Instrumentation BASES BACKGROUND This LCO addresses Refueling CHR actuation. When the Refueling CHR Monitors are enabled, a CHR actuation may be automatically initiated by a signal from either of the Refueling CHR monitors or manually by actuation of either of the control room CHR 11 Manual Initiate pushbuttons (pushing either Manual Initiate 11 pushbutton will actuate both trains of CHR). A CHR signal initiates the following actions:
a. Control Room HVAC Emergency Mode;

b. Containment Isolation Valve Closure; and

c. Block automatic starting of Engineered Safeguards pump room sump pumps.
The Refueling CHR signal provides automatic containment
isolation valve closure during refueling operations, using two radiation monitors located in the refueling area of the containment (elevation 649 ft). The monitors are part of the
*plant area monitoring system and employ one-out-of-two logic for isolation. During normal operation these monitors will not initiate a CHR signal. A switch is provided so that CHR actuation can be initiated by these monitors during refueling only.
Each monitor actuates one train of CHR logic when. containment radiation exceeds the setpoint. Two separate enabling keylock switches, one per train, enable the Refueling CHR input to the CHR logic when switched to the Refueling Mode. Each 11 11 Refueling CHR channel, associated keylock switch, and initiation circuit input to the CHR logic thus forms a one-out-of-one logic input to its associated CHR actuation logic train. The Refueling CHR isolation instrumentation is separate from the CHR instrumentation addressed in LCO 3.3.3, 11 ESF Instrumentation. However, the Refueling CHR 11 Instrumentation does operate the same CHR actuation relays as the two-out-of-four CHR logic addressed in LCO 3.3.4. This LCO is not included in LCOs 3.3.3 and 3.3.4 because of the differences in APPLICABILITY and the single channel nature of the Refueling CHR inp~t. The Refueling CHR signal performs the automatic containment isolation valve closure Function during refueling operations required by LCO 3;9.3, Containment 11 Penetrations. 11 Palisades Nuclear Plant B 3.3.6-1 05/30/99 .
Refueling CHR Instrumentation B 3.3.6
BASES BACKGROUND (continued)
The Refueling CHR Instrumentation provides protection from release of radioactive)gases and particulates from the containment in the event a fuel assembly should be severely damaged during handling. The Refueling CHR Instrumentation will detect any abnormal radiation levels in the containment refueling area and will initiate purge valve closure to limit the release of radioactivity to the environment. The containment purge supply and exhaust valves are closed on a CHR signal when a high radiation level in containment is detected. The Refueling CHR Instrumentation includes two independent, redundant actuation subsystems, as described above. Reference 1 describes the Refueling CHR circuitry. Trip Setpoint No required setpoint is specified because these instruments are not assumed to function by any of the safety analyses. Typically, the instruments are set at about 25 mR/hr above exp*ected background for planned operations (including movement of the reactor vessel head or internals).
APPLICABLE The Refueling CHR Instrumentation isolates containment in the SAFETY ANALYSES event that area radiation exceeds an established level following a fuel handling accident. This ensures the radioactive materials are not released directly. to the environment and significantly reduces the offsite doses from those calculated by the safety analyses, which do not credit containment isolation (Ref. 2). Either way, i.e., with or without containment isolation, the offsite doses remain within the guidelines of 10 CFR 100.
The Refueling CHR Instrumentation is not required by the fuel handling accident analyses to maintain offsite doses within the guidelines of 10 CFR 100, but operating experience indicates that containment isolation provides significant reduction of the resulting offsite doses. Therefore, the Refueling CHR Instrumentation satisfies the requirements of Criterion 4 of 10 CFR 50.36(c)(2). .
\
Palisades Nuclear Plant B 3.3.6-2 05/30/99
Refueling CHR Instrumentation B 3.3.6
BASES LCO The LCO for the Refueling CHR Instrumentation requires that two channels of refueling CHR instrumentation and two channels of CHR manual initiation be OPERABLE, including the logic components necessary to initiate Refueling CHR Isolation. The CHR setpoint is chosen to be high enough to avoid inadvertent actuation in the event of normal background radiation fluctuations during fuel handling and movement of the reactor internals, but low enough to alarm and isolate the containment in the event of a Design Basis fuel handling accident.
APPLICABILITY In MODE 5 or 6, the Refueling CHR isolation of containment isolation valves is not normally required to be OPERABLE. However, during CORE ALTERATIONS or during movement of irradiated fuel within containment, there is the possibility of a fuel handling accident requiring containment isolation on high radiation in containment. Accordingly, the Refueling tHR Instrumentation must be OPERABLE during CORE ALTERATIONS and when moving any irradiated fuel in containment. In MODES 1, 2, 3 and 4, both the Containment High Pressure (CHP) and CHR signals provide containment isolation as discussed in the Bases for LCO 3.3.3 and LCO 3.3.4. ACTIONS A.1. A.2.1~ and A.2.2 Condition A applies to the failure of one Refueling CHR monitor channel, one CHR Manual Initiate channel, or one of each. The Required Action allows either initiation of a CHR signal by placing the inoperable channel in trip (which accomplishes the safety function of the inoperable channel), or suspension of CORE ALTERATIONS and movement of irradiated fuel assemblies within containment (which places the plant in a condition where the LCO does not apply). The Completion Time of 4 hours is warranted because one additional channel of each Function remains operable during that period. The suspension of CORE ALTERATIONS and fuel movement shall not preclude completion of movement of a component to a safe position .
Palisades Nuclear Plant B 3.3.6-3 05/30/99
Refueling CHR Instrumentation B 3.3.6
BASES ACTIONS (continued)
B.1 and B.2 Condition B applies when either no automatic Refueling CHR or no Manual CHR (or neither) is available. The Required Action is to immediately suspend CORE ALTERATIONS and movement of irradiated fuel assemblies within containment. This places the plant in a condition where the LCO does not apply. The Completion Time is warranted on the basis that at least one containment isolation Function is completely lost. The suspension of CORE ALTERATIONS and fuel movement shall not preclude completion of movement of a component to a safe position. SURVEILLANCE SR 3.3.6.1 REQUIREMENTS Performance of the CHANNEL CHECK once every 12 hours ensures that a gross failure of instrumentation has not occurred. A CHANNEL CHECK is normally a comparison of the parameter
indicated on one channel to a similar parameter on other channels. It is based on the assumption that instrument channels monitoring the same parameter should read approximately the same value.
Significant deviations between the two instrument channels could be an indication of excessive instrument drift in one of the channels or actual differing radiation levels at the two detector locations. CHANNEL CHECK will detect gross channel failure; thus, it is key to verifying the instrumentation continues to operate properly between each CHANNEL CALIBRATION. The Frequency, about once every shift, is based on operating experience that demonstrates the rarity of channel failure. Since the probability.of two random failures in redundant channels in any 12 hour period is low, the CHANNEL CHECK minimizes the chance of loss of protective function due to failure of redundant channels. The CHANNEL CHECK supplements less formal, but more frequent, checks of channel OPERABILITY during normal operational use of the displays associated with the LCO required channels. *
Palisades Nuclear Plant B 3.3.6-4 05/30/99
Refueling CHR Instrumentation B 3.3.6
BASES SURVEILLANCE REQUIREMENTS SR 3.3.6.2 (continued) A CHANNEL FUNCTIONAL TEST is performed on each Refueling CHR channel to ensure the entire channel will perform *its intended function. A successful test of the required contact(s) of a channel relay may be performed by the verification of the change of state of a single contact of the relay. This clarifies what is an acceptable CHANNEL FUNCTIONAL TEST of a relay. This is acceptable because all of the other required contacts of the relay are verified by other Technical Specifications and non-Technical Specifications tests at least once per refueling interval with applicable extensions.
The Frequency of 31 days is based on plant operating experience with regard to channel OPERABILITY, which demonstrates that failure of more than one channel of a given Function in any 31 day interval is a rare event. SR 3.3.6.3
A CHANNEL FUNCTIONAL TEST is performed on each CHR Manual Initiation channel to ensure it will perform its intended .
function. The Frequency of 18 months is based on plant operating experience with regard to channel OPERABILITY, and is consistent with the testing of other manually actuated functions. SR 3.3.6.4 A CHANNEL CALIBRATION is a complete check of the instrument channel including the sensor. The Surveillance verifies that the channel responds to a measured parameter within the necessary range and accuracy. CHANNEL CALIBRATION leaves the channel adjusted to account for instrument drift between successive calibrations to ensure that the channel remains operational between successive tests. The Frequency is based upon the assumption of an 18 month calibration interval in the setpoint determination. REFERENCES 1. FSAR, Section 7.3
2. FSAR, Section 14.19 Palisades Nuclear Plant B 3.3.6-5 05/30/99
PAM Instrumentation B 3.3.7
B 3.3 INSTRUMENTATION B 3.3.7 Post Accident Monitoring (PAM) Instrumentation BASES BACKGROUND The primary purpose of the Post Accident Monitoring (PAM) instrumentation is to display plant variables that provide information required by the control room operators during accident situations. This information provides the necessary support for the operator to take the manual actions, for which no automatic control is provided, that are required for safety systems to accomplish their safety Functions for Design Basis Events.
The OPERABILITY of the PAM instrumentation ensures that there is sufficient information available on selected plant parameters to monitor and assess plant status and behavior following an accident. The availability of PAM instrumentation is important so that responses to corrective actions can be observed and the need for, and ma*gni tu de of, further actions can be determined . These essential instruments are identified in the FSAR Appendix 7C (Ref. l) and address the recommendations of Regulatory Guide -1.97 (Ref. 2), as required by Supplement 1 to NUREG-0737, "TMI Action Items" (Ref. 3). Type A variables are included in this LCO because they provide the primary information required to permit the- control room operator to take specific manually controlled actions, for which no automatic control is provided, that are required for safety systems to accomplish their safety functions for Design. Basis Accidents (DBAs).
Category I variables are the key *variables deemed risk significant because they are needed to:

Determine whether other systems important to safety are performing their intended functions;

Provide information to the operators that will enable them to determine the potential for causing a gross breach of the barriers to radioactivity release; and

Provide information regarding the release of radioactive materials to allow for early indication of the need to initiate action necessary to protect the public and for

Palisades Nuclear Plant an estimate of the magnitude of any impending threat .
B 3.3.7-1 05/30/99
PAM Instrumentation B 3.3.7
BASES BACKGROUND (continued)
These key variables are identified in Reference 1 by plant specific Regulatory Guide 1.97 analyses (Ref. 1). This analysis identified the plant specific Type A and Category 1 variables and provided justification for deviating from the NRC proposed list of Category I variables. The specific instrument Functions listed in Table 3.3.7-1 are discussed in the LCD Bases. APPLICABLE The PAM instrumentation ensures the OPERABILITY of Regulatory SAFETY ANALYSES Guide 1.97 Type A variables, so that the control room operating staff can:
Perform the diagnosis specified in the emergency operating procedures. These variables are restricted to preplanned actions for the primary success path of DBAs; and

Take the ipecified, preplanned, manually controlled actions, for which no automatic control is provided, that are required for safety systems to accomplish their safety functions.
The .PAM instrumentation also ensures OPERABILITY of Category I, non-Type A variables. This ensures the control room operating staff can: *
Determine whether systems important to safety are performing their intended functions;

Determine the potential for causing a gross breach of the barriers to radioactivity release;

Determine if a gross breach of a barrier has occurred; and

Initiate action necessary to protect the public as well as to obtain an estimate of the magnitude of any impending threat.
PAM instrumentation that satisfies the definition of Type A in Regulatory Guide 1.97 meets Criterion 3 of 10 CFR 50.36(c)(2) .
Palisades Nuclear Plant B 3.3.7-2 05/30/99
PAM Instrumentation B 3.3.7
BASES APPLICABLE SAFETY ANALYSES Category I, non-Type A PAM instruments are retained in the Specification because they are intended to assist operators (continued) in minimizing the consequences of accidents. Therefore, these Category I variables are important in reducing public risk.
LCO LCO 3.3.7 requires at least two OPERABLE channels for all but one Function to ensure no single failure prevents the operators from being presented with the information necessary to determine the status of the plant and to bring the. plant to, and maintain it in, a safe condition following that acci~ent. Furthermore, provision of at least two channels allows a CHANNEL CHECK during the post accident phase to confirm the validity of displayed information. The exception to*the two channel requirement is Containment Isolation Valve Position. In this case, the important information is the status of the containment penetrations. The LCO requires one position indicator for each active containment isolation valve. This is sufficient to redundantly verify the isolation status of each isolable penetration either via indicated status of the active valve and prior knowledge of the passive valve or via system boundary status. If a normally active containment-isolation valve is known to be closed and deactivated, position indication is not needed to determine status. Therefore, the position indication for valves in this state is not required to be OPERABLE. *I Listed below are discussions of the specified instrument Functions listed in Table 3.3.7-1. Component identifiers of the sensors, indicators, power supplies, displays, and recorders in each instrument loop are found in Reference 1. 1, 2. Primary Coolant System (PCS) Hot and Cold Leg Temperature (wide range) PCS wide range Hot and Cold Leg Temperatures are Type B, Category 1 variables provided for verification of core cooling and long term surveillance. Reactor outlet temperature inputs to the PAM are provided by two wide range resistance elements and associated transmitters (one in each loop). The channels provide indication over a range of 50°F to 700°F .
Palisades Nuclear Plant B 3.3.7-3 05/30/99
PAM Instrumentation B 3.3.7
BASES LCO (continued)
3. Wide Range Neutron Flux Wide Range Neutron Flux indication is a Type B, Category 1 variable, and is provided to verify reactor shutdown.

4. Containment Floor Water Level (wide range)
Wide range Containment Floor Water Level is a Type B, Category 1 variable, and is provided for verification and long term surveillance of PCS integrity.
5. Subcooled Margin Monitor The Subcooled Margin Monitor (SMM) is a Type A, Category 1 variable used to identify conditions which require tripping of the primary coolant pumps and throttling of safety injection flows. Each SMM channel uses a number of PCS pressure and temperature inputs to determine the degree of PCS subcooling or superheat .
6. Pressurizer Level (Wide Range)
Pressurizer Level is a Type A, Category 1 variable, and is used to determine whether to terminate Safety Injection (SI), if still in progress, or to reinitiate SI if it has been stopped. Knowledge of pressurizer water level is also used to verify the plant conditions necessary to establish natural circulation in the PCS and to verify that the plant is maintained in a safe shutdown condition.
7. Containment Hydrogen Monitors Containment Hydrogen Monitors are provided to detect high hydrogen concentration conditions (a Type A, Category 1 variable) that represent a potential for containment breach and are used to determine when to place the hydrogen recombiners in operation. This variable is also important in verifying the adequacy of mitigating actions *
Palisades Nuclear Plant B 3.3.7-4 05/30/99
PAM Instrumentation B 3.3.7
BASES LCO (continued)
8. Condensate Storage Tank (CST) Level CST Level is a Type D, Category 1 variable, and is provided to ensure water supply for AFW. The CST provides the safety grade water supply for the AFW System. Inventory is monitored by a 0 to 100% level indication. CST Level is displayed on a control room indicator. In addition, a control room annunciator alarms on low level.
The CST is the initial source of water for the AFW System. However, as the CST is depleted, manual operator action is necessary to replenish the CST.
9. Primary Coolant System Pressure (wide range)
PCS wide range pressure is a Type A, Category 1 variable provided for verification of core cooling and PCS integrity long term ~urveillance. Wide range PCS loop pressure is measured by pressure
transmitters with a span of 0 psia to 3000 psig. Redundant monitoring capability is provided by two channels of instrumentation. Control room indications are provided on Cl2 and C02.
10. Containment Pressure (wide range)
Wide range Containment Pressure is a Type C, Category 1 variable, and is provided for verification of PCS and containment OPERABILITY. It is also an input to decisions for initiating containment spray. 11, 12. Stearn Generator Water Level (wide range) Wide range Steam Generator Water Level is a Type A, Category 1 variable, and is provided to monitor operation of decay heat removal via the steam generators. The steam generator level instrumentation covers a span extending from the tube sheet to the steam separators, with an indicated range of -140% to
+150%. Redundant monitoring capability is provided by two channels of instrumentation for each SG .. , Palisades Nuclear Plant B 3.3.7-5 05/30/99
PAM Instrumentation B 3.3.7
BASES LCD 11, 12. Steam Generator Water Level (wide range) (continued)
Operator action for maintenance of heat removal is based on the control room indication of Steam Generator Water Level. The indication is used during a SG tube rupture to determine which SG has the ruptured tube. It is also used to determine when to initiate once through cooling on low water level. 13, 14. SG Pressure Steam Generator. Pressure is a Type A. Category 1 variable used in accident identification, including Loss of Coolant, and Steam Line Break. Redundant monitoring capability is provided by two channels of instrumentation for each SG.
15. Containment Isolation Valve Position Containment Isolation Valve (CIV) Position is a Type B, Category 1 variable, and is provided for verification of
containment OPERABILITY .
CIV position is provided for verification of containment integrity. In the case of CIV position, the important information is the isolation status of the containment penetration .. The LCD requires one channel of valve position
. indication in the control room to be OPERABLE for each active CIV in a containment penetration flow path.; This is sufficient to redundantl_y verify the isolation status of each isolable penetration via indicated status of the active valve, as applicable, and prior knowledge of passi~e valve or system boundary status. If a penetration flow path is isolated, position indication for the CIV(s) in the associated penetration flow path is not needed to determine status.
Therefore, as indic&ted in Note (a) the position indication for valves in an isolated penetration flow path is not required to be OPERABLE. Palisades Nuclear Plant B 3.3.7-6 05/30/99
PAM Instrumentation B 3.3.7
BASES LCO (continued) 16, 17, 18, 19. Core Exit Temperature Core Exit Temperature is a Type C, Category 1 variable, and is provided for verification and long term surveillance of core cooling.
Each Required Core Exit Thermocouple (CET) channel consists of a single environmentally qualified thermocouple. The design of the Incore Instrumentation System incl~des a . Type K (chromel alumel) thermocouple within each of the incore instrument detector assemblies. The junction of each thermocouple is located above the core exit, inside the incore detector assembly guide tube, that supports and shields the incore instrument detector assembly string from flow forces in the .outlet plenum region. These core exit thermocouples monitor the temperature of the reactor coolant as it exits the fuel assemblies.
The core exit thermocouples have a usable tem~erature range from 32°F to 2300°F, although accuracy is reduced at temperatures above 1800°F.
20. Reactor Vessel Water Level Reactor Vessel Water Level is monitored by the Reactor Vessel Level Monitoring System (RVLMS) and is a Type B, Category 1 variable provided for veri fi ca ti on and 1ong term survei 11 ance of core cooling.
The RVLMS provides a direct measurement of the collapsed liquid level above the fuel alignment plate. The collapsed level represents the amount of liquid mass that is in the reactor vessel above th~ core. Measurement of the collapsed water level is selected because it is a direct indication of the water inventory. The collapsed level is obtained over the same temperature and pressure range as the saturation measurements, thereby encompassing all operating and accident conditions where it must function. Also, it functions during the recovery interval~ Therefore, it is designed to survive the high steam temperature that may occur during the preceding core recover~ interval .
Palisades Nuclear. Plant
( B 3.3.7-7 05/30/99.
PAM Instrumentation B 3. 31. 7
BASES LCO 20. Reactor Vessel Water Level (continued)
The level range extends from the top of the vessel down to the top of the fuel alignment plate. A total of eight Heated Junction Thermocouple (HJTC) pairs are employed in each of the two *RVLMS channels. Each pair consists of a heated junction TC and an unheated junction TC. The differential temperature at each HJTC pair provides discrete indication of uncovery at the HJTC pair location. This indication is displayed using LEDs in the control room. This provides the operator with adequate indication to track the progression of the accident and to detect the consequences of its mitigating actions or the functionality of automatic equipment. A RVLMS channel consists of eight sensors in a probe. A channel is OPERABLE if four or more sensors, two or more of the upper four and two or more of the lower four, are OPERABLE.
21. Containment Area Radiation (high* range)
High range Containment Area Radiation is a Type E, Category 1 variable, and is provided to monitor for the potential of significant radiation releases and to provide release assessment for use by operators in determining the nee~ to 1 invoke site emergency plans. APPLICABILITY The PAM instrumentation LCO is applicable in MODES 1~ 2, and 3. These variables are related to the diagnosis and preplanned actions required to mitigate DBAs. The applicable DBAs are assumed to occur in MODES 1, 2, and 3. In MODES 4, 5, and 6, plant conditions are such that the likelihood of an event occurring that wou~d require PAM instrumentation is low; therefore, PAM instrumentation is not required to be OPERABLE in these MODES .
Palisades Nuclear Plant B 3.3.7-8 05/30/99
PAM Instrumentation B 3.3.7
BASES ACTIONS Note 1 has been added in the ACTIONS to exclude the MODE change restriction of LCO 3~0.4. This exception allows entry into the applicable MODE while relying on the ACTIONS, even though the ACTIONS may eventually require plant shutdown. This exception is acceptable due to the passive function of the instruments, the operator's ability to monitor an accident using alternate instruments and methods, and the low probability of an event requiring these instruments.
Note 2 has been added in the ACTIONS to clarify the application of Completion Time rules. The Conditions of this Specification may be entered independently for each Function listed in Table 3.3.7-1. The Completion Time(s) of the inoperable channel(s) of a Function will be tracked separately for each Function, starting from the time the Condition was entered for that Function. When one or more Functions have one required channel that is inoperable, the required inoperable channel must be restored to OPERABLE status within 30 days. The 30 day Completion Time is based on operating experience and takes into account the remaining OPERABLE channel, the passive nature of the instrument (no critical automatic action is assumed to occur from these instruments), and the low probability of an event requiring PAM instrumentation during this interval. This Required Action specifies initiation of actions in accordance with Specification 5.6.6, which requires a written report to be submitted to the Nuclear Regulatory Commission. This report discusses the results of the root cause evaluation of the inoperability and identifies proposed restorative Required Actions. This Required Action is appropriate in lieu of a shutdown requirement, given the likelihood of plant conditions that would require information provided by this instrumentation. Also, alternative Required Actions are identified before a loss of functional capability condition occurs .
Palisades Nuclear Plant B 3.3.7-9 05/30/99
I PAM Instrumentation B 3.3.7
BASES ACTIONS (continued)
C.1 When one or more Functions have two required channels inoperable (i.e., two channels inoperable in the same Function), one channel in the Function should be restored to OPERABLE status within 7 days. The Completion Time of 7 days is based on the ,relatively low probability of an event requiring PAM instrumentation operation and the availability of alternate means to obtain the required information. Continuous operation with two required channels inoperable in a Function is not acceptable because the alternate indications may not fully meet all performance qualification requirements applied to the PAM instrumentation. Therefore, requiring restoration of one inoperable channel of the Function limits the risk that the PAM Function will be in a degraded condition should an accident occur. Condition C is modified by a Note which indicates it is not applicable to hydrogen monitor channels .
Condition D applies when two hydrogen monitor channels are inoperable. Required Action D.1 requires restoring one hydrogen monitor channel to OPERABLE status within 72 hours.
The 72 hour Completion Time is. reasonable based on the backup capability of the Post Accident Sampling System to monitor the hydrogen concentration for evaluation of core damage and to provide information for operator decisions. Also, it is unlikely that a LOCA (which would cause core damage) would occur during this time. E.1 This Required Action directs entry into the appropriate Condition referenced in Table 3.3.7-1. The applicable Condition referenced in the Table is Function dependent. Each time Required Action C.1 or D.1 is not met, and the associated Completion Time has expired, Condition E is entered for that channel and provides for transfer to the appropriate subsequent Condition .
Palisades Nuclear Plant B 3.3.7-10 05/30/99
PAM Instrumentation B 3.3.7
BASES ACTIONS (continued)
F.1 and F.2 If the Required Action and associated Completion Time of Condition C or D are not met, and Table 3.3.7-1 directs entry into Condition F, the plant must be brought to a MODE in which the requirements of this LCO do not apply. To achieve this status, the plant must be brought to at least MODE 3 within 6 hours and to MODE 4 within 30 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems. G.1 Alternate means of monitoring Reactor Vessel Water Level and Containment Area Radiation have been developed and tested. These alternate means may be temporarily installed if the normal PAM channel cannot be restored to OPERABLE status within the allotted time. If these alternate means are used, the Required Action is not to shut down the plant, but rather to follow the directions of Specification 5.6.6. The report provided to the NRC should discuss the alternate means used, describe the degree to which the alternate means are equivalent to the installed PAM channels, justify the areas in which they are not equivalent, and provide a schedule for restoring the normal PAM cnannels. SURVEILLANCE A Note at the beginning of the Surveillance Requirements REQUIREMENTS specifies that the following SRs apply to each PAM instrumentation Function in Table 3.3.7-1. SR 3.3.7.1 Performance of the CHANNEL CHECK once every 31 days ensures that a gross failure of instrumentation has not occurred. A CHANNEL CHECK is normally a comparison of the parameter indicated on one channel to a similar parameter on other channels. It is based on the assumption that instrument channels monitoring the same parameter should read approximately the same value. Significant deviations between the two instrument channels could be an indication of excessive instrument drift in one of the channels or of something even more serious. A CHANNEL CHECK will detect gross channel failure; thus, it is key to verifying the instrumentation continues to operate properly between each CHANNEL CALIBRATION. Palisades Nuclear Plant B 3.3.7-11 05/30/99
PAM Instrumentation B 3.3.7
BASES SURVEILLANCE REQUIREMENTS SR 3.3.7.1 (continued)
( Agreement criteria are determined by the plant staff based on a combination of the channel instrument uncertainties, including indication and readability. If a channel is outside the criteria, it may be an indication that the sensor or the signal processing equipment has drifted outside its limit. If the channels are within the criteria, it is an indication that the channels are OPERABLE. If the channels are normally off scale during times when surveillance is required, the CHANNEL CHECK will only verify that they are off scale in the same direction. Off scale low current loop channels are verified to be reading at the bottom of the range and not failed downscale. As indicated in the SR, a CHANNEL CHECK is only required for those channels which are normally energized.
The Frequency of 31 days is based upon plant operating experience with regard to channel OPERABILITY and drift, which demonstrates that failure of more than one cha~nel of a given Function in any 31 day interval is a rare event. The CHANNE~
CHECK supplements less formal, but more frequent, checks of channel during normal operational use of the displays associated with this* LC0 s required channels. 1 SR 3.3.7.2 A CHANNEL CALIBRATION is performed every 18 months or approximately every refueling. CHANNEL CALIBRATION is typically a complete check of the instrument channel including the sensor. Therefore, this SR is modified-by a Note which states that it is not necessary to calibrate neutron detectors because of the difficulty of simulating a meaningful signal. Wide range and source range nuclear instrument channels are not calibrated to indicate the actual power level or the flux in the detector location. The circuitry is adjusted so that wide range and source range readings may be used to determine the approximate reactor flux level for comparative purposes. The Surveillance verifies the channel responds to the measured parameter within the necessary range and accuracy. For the core exit thermocouples, a CHANNEL CALIBRATION is . performed by substituting a known voltage for the thermocouple.
Palisades Nuclear Plant B 3.3.7-12 05/30/99
PAM Instrumentation B 3.3.7
BASES SURVEILLANCE REQUIREMENTS SR 3.3.7.2 (continued)
The Frequency is based upon operating experience and consistency with the typical industry refueling cycle and is justified by an 18 month calibration interval for the determination of the magnitude of equipment drift. REFERENCES 1. FSAR, Appendix 7C, 11 Regulatory Guide 1.97 Instrumentation 11
2. Regulatory Guide 1.97

3. NUREG-0737, Supplement 1
Palisades Nuclear Plant B 3.3.7-13 05/30/99
Alternate Shutdown System B 3.3.8
- B 3.3 INSTRUMENTATION B 3.3.8 Alternate Shutdown System BASES BACKGROUND The Alternate Shutdown System provides the control room operator with sufficient instrumentation and controls to maintain the plant in a safe shutdown condition from a location other than the control room. This capability is necessary to protect against the possibility that the control room becomes inaccessible. A safe shutdown condition is defined as MODE 3.
With the plant in MODE 3, the Auxiliary Feedwater (AFW) System and the steam generator safety valves or the steam generator atmospheric dump valves can be used to remove core decay heat and meet all safety requirements. The long term supply of water for the AFW System and the ability to borate the Primary Coolant System (PCS) from outside the control room allow extended operation in MODE 3. In order to ensure use of sufficient components of the AFW System and sufficient process information to permit reactor
MODE 3 control in the event a fire damages equipment and circuitry of the main.feedwater system or the AFW System in the control room, cable spreading room, Engineered Safeguards Auxiliary Panel C-33 room, or the corridor between Switchgear Room 1-C and the charging pump rooms, auxiliary Hot Shutdown Control Panels (C-150/C-150A) have been provided and located in the southwest electrical penetration room. These panels are comprised of two enclosures, the main enclosure C-15Q*and an auxiliary enclosure C-150A. The description below combines these two enclosures into one entity "Panel C-150."
From this panel, control of the AFW flow control valves and control of AFW turbine steam supply Valve B can be enabled. Indication of AFW flow frbm the steam driven AFW pump to both Steam Generators (SGs), water level of both SGs, and pressurizer level are enabled by transfer. In addition, primary coolant pressure (pressurizer pressure) is displayed by a primary sensor dedicated to this use. Transfer of the above-mentioned systems is annunciated in the control room. See FSAR Section 7.4 (Ref. 1) for operation via Panel C-150. The equipment controls that are required are listed in Table 3.3.8-1 .
Palisades Nuclear Plant B 3.3.8-1 05/30/99
Alternate Shutdown System B 3.3.8
BASES BACKGROUND (continued)
Switches, which transfer control or instrument functions from the control room to the auxiliary shutdown control panel, alarm in the control room when the devices in the alternate hot shutdown panel are enabled. APPLICABLE The Alternate Shutdown System is required to provide equipment SAFETY ANALYSES at appropriate locations outside the control room with a capability to maintain the plant in a safe condition in MODE 3. The criteria governing the design and the specific system requirements of the Alternate Shutdown System are located in 10 CFR 50, _Appendix A, GDC 19, and Appendix R (Ref. 2). The Alternate Shutdown System has been identified as an important contributor to the reduction of plant risk to accidents and, therefore, satisfies the requirements of Criterion 4 of 10 CFR 50.36(c)(2) .
LCD The Alternate Shutdown System LCO provides the requirements for the OPERABILITY of one channel of the instrumentation and controls necessary to maintain the 'plant in .MODE 3 from a location other than the control room. The instrumentation and controls required are listed in Table 3.3.8-1 in the accompanying LCO.
Equipment controls that are required.by the alternative dedicated method of maintaining MODE. 3 are as follows:
1. AFW flow control valves (CV-0727 and CV-0749); and

2. Turbine-driven AFW pump.
Instrumentation systems displayed on the Auxiliary Hot Shutdown Control Panel are:
1. Source range flux monitor;

2. AFW flow (HIC-0727 and HIC-0749C);.

3. Pressurizer pressure;

4. Pressurizer 1eve h
5.
. Palisades Nuclear Plant SG level and pressure; B 3.3.8-2 05/30/99
Alternate Shutdown System B 3.3.8
BASES LCD
' (continued)
6. Primary coolant temperatures (hot and cold legs);

7. Turbine-driven AFW pump low-suction pressure warning 1i ght; and

8. SIRW tank level.
A Function of an Alternate Shutdown System is OPERABLE if* all instrument and control channels needed to support the remote shutdown Functions are OPERABLE. The Altern~te Shutdown System instrumentation and control circuits covered by this LCD do not need to be energized to be considered OPERABLE. This LCD is intended to ensure that the instrument and control circuits will be OPERABLE i.f *plant conditions require that the Alternate Shutdown System be placed in operation. Table 3.3.8-1 Indication Channel 1, Source Range Nuclear
. Instrumentation, uses the same detector and preamplifier as *the* *control room channel. Optical isolation is provided between
the control room*and AHSDP (Alternate Hot Shut Down Panel)
. portions of the circuit. When .the control switches are changed to the "AHSDP" position, the detector and preamplifier is iso.lated from its normal power supply and connected into the AHSDP power supply.
Table 3.3~8-l Indication Channels 2 and 12 are provided with their own presstire and level transmitter. The associated circuitry is energized when the AHSDP is energ1zed. The other Table 3.3.8-1 Indication Channels in Table 3.3.8~1 use a transmitter which also serves normal control room instrumentation. When th~ control switches are changed to the "AHSDP" (Alternate Hot Shut Down Panel) ~osition, the transmitter is isolated from- its normal power supply and
. circuitry, and connected into the C-150 or C-150A panel circuit; control for AFW flow control valves CV-0727 and CV-0749 is also transferred to C-150. The transfer switches are alarmed in the control room .
Palisades Nuclear Plant B 3.3.8-3 05/30/99
Alternate Shutdown System B 3.3.8
- BASES APPLICABILITY The Alternate Shutdown System LCO is applicable in MODES l, 2, and 3. This is required so that the plant can be maintained in MODE 3 for an extended period of time from a location other than the control room.
This LCO is not applicable in MODE 4, 5, or 6. In these MODES, the plant is already subcritical and in the condition of
*reduced PCS energy. Under these co~ditions, considerable time is available to restore necessary instrument control Functions if control room instruments or control become unavailable.
ACTIONS A Note has been included that excludes the MODE change restrictions of LCO 3.0.4. This exception allows entry into an applicable MODE while relying on the ACTIONS, even though the ACTIONS may eventually require a plant shutdown. This is acceptable due to the low probability of an event requiring this system. The Alternate Shutdown System equipment can generally be repaired during operation without significant risk of spurious trip .
Note 2 has been added in the ACTIONS to clarify the a*pplication of Completion Time rules. - The Conditions of this Specification may be entered independently for each Function listed in Table 3.3.8-1. The Completion Time of the inoperable channel of a Function will be tracked separately for each Function, starting from.the time the Condition was entered for that Function.
Condition A addresses the situation where the required channels of the Remote Shutdown S~stem are inoperable. This includes any Function listed in Table 3.3.8-1 as well as the control and transfer switches. Required Action A. is to restore the channel to OPERABLE status within 30 days. This allows time to complete repairs on the failed channe 1. The Completion Time is based on operating experience and the low probability of an event that would require evacuation of the control room .
Palisades Nuclear Plant B 3.3.8-4 05/30/99
r Alternate
\
Shutdown System B 3.3.8
*- BASES ACTIONS (continued)
B.1 and B.2 If the Required Action and associated Completion Time of Condition A are not met, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 6 hours and to MODE 4 within 30 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required MODE from full power conditions in an orderly manner and without challenging plant systems. SURVEILLANCE SR 3.3.8.1 REQUIREMENTS This SR applies to the startup range neutron flux monitoring channel. The _CHANNEL FUNCTIONAL TEST consists of verifying proper response of the channel to the internal test signals, and verification that a detectable signal is available from the detector. After lengthy shutdown periods flux may be below the range of the channel indication. Signal verification with test equipment is acceptable .
. The CHANNEL FUNCTIONAL TEST of the startup range neutron flux monitoring channel is performed once within 7 days prior to reactor startup. The Frequency is based on plant operating experience that demonstrates channel failure is rare.
SR 3.3.8.2 S~ 3.3.8.2 verifies that each required Alternate Shutdown System transfer switch and control circuit performs its intended function. This verification is performed from AHSDPs C-150 and C-150A and locally, as appropriate. Operation of the equipment from the AHSDPs C-150 and C-150A is not necessary. The Surveillance can be satisfied by performance of a continuity check. This will ensure that if the control room becomes inaccessible, the plant can be maintained in MODE 3 from the auxiliary shutdown panel and the local control stations. The 18 month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transie~t if the Surveillance were performed with the reactor at power. Operating experience demonstrates that Alternate Shutdown
System control channels seldom fail to pass the Surveillance when performed at a Frequency of once every 18 months.
Palisades Nuclear Plant B 3.3.8-5 05/30/99
Alternate Shutdown System B 3.3.8
BASES SURVEILLANCE REQUIREMENTS SR 3.3.8.3 (continued A CHANNEL CALIBRATION is a complete check of the instrument channel including the sensor. The Surveillance verifies that the channel responds to the measured parameter within the necessary range and accuracy.
Performance of a CHANNEL CALIBRATION every 18 months on Functions 1 through 15 ensures that the channels are operating accurately and within specified tolerances. This verification is performed from the AHSDPs and locally, as appropriate. A test of the AFW pump suction pressure alarm (Function 15) is included as part of its CHANNEL CALIBRATION. This will ensure that if the control room becomes inaccessible, the plant can be maintained in MODE 3 from the AHSDPs and local control stations. The 18 month Frequency is based upon the need to perform this Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power .
Operating experience demonstrates that Alternate Shutdown System instrumentation channels seldom fail to pass the Surveillance when performed at a Frequency of once every 18 months. Therefore, the Frequency was concluded to be
~cceptable from a reliability standpoint.
This SR is modified by two Notes. Note 1 states that the SR is not required for Functions 16, 17, and 18; Note 2 states that it is not necessary to calibrate neutron detectors because of the difficulty of simulating a meaningful signal. Wide range and source range nuclear instrument channels are not calibrated to indicate the actual power level or the flux in the detector location. The circuitry is adjusted so that wide range and source range readings may be used to determine the approximate reactor flux level for comparative purposes. REFERENCES 1. FSAR, Section 7.4, 0ther Safety Related Protection, 11 Control, and Display Systems 11
2. 10 CFR 50, Appendix A, GDC 19 and Appendix R.
Palisades Nuclear Plant B 3.3.8-6 05/30/99
Neutron. Flux Monitoring Channels B 3.3.9
B 3~3 INSTRUMENTATION B 3.3.9 Neutron Flux Monitoring Channels BASES BACKGROUND The neutron flux monitoring channels consist of two combined source range/wide range channels, designated NI-1/3 and NI-2/4.
The wide range portions, (NI-3 and NI-4) provide neutron flux power indication from < lE-7% RTP to > 100% RTP. The source range portions, designated NI-1 and NI-2, provide source range indication over the range of 1 to 1E+5 cps. This LCO addresses MODES 3, 4, and 5. In MODES 1 and 2, the neutron flux monitoring requirements are addressed by LCO 3.3.1, "Reactor Protective System (RPS} Instrumentation." When the plant is shutdown, both neutron flux monitoring channels must be available to monitor neutron flux. If only one section of a neutron flux monito~ing channel (source range or wide range) is functioning, the neutron flux monitoring channel may be considered OPERABLE if it is capable of
detecting the existing reactor neutron flux. In this application, the RPS channels need not be OPERABLE since the reactor trip Function is n6t required. By monitoring neutrtin.
flux power, loss of SOM caused by boron dilution can be detected as an increase in flux. Two channels must be OPERABLE . to provide single failure protection and to facilitate detection of channel failure by providing CHANNEL CHECK capability. APPLICABLE The wide range neutron flux monitoring channels are necessary SAFETY ANALYSES to monitor core reactivity changes. They are the primary means for detecting and triggering operator actions to respond to reactivity transients initiated from conditions in which the RPS is not required to be OPERABLE. The neutron flux monitoring channel s LCO requirements support compliance with 1 10 CFR 50, Appendix A, GDC 13 (Ref. 1). The FSAR, Chapters 7 and 14 (Refs. 2 and 3, respectively), describes the specific neutron flux monitoring channel features that are critical to comply with the GDC. The OPERABILITY of neutron flux monitoring channels is necessary to meet the assumptions of the safety analyses and provide for the detec'ti on of reduced SOM. I
The neutron flux monitoring channels satisfy Criterion 4 of 10 CFR 50.36(c)(2).
Palisades Nuclear Plant B 3.3.9-1
05/30/99 I
Neutron Flux Monitoring ChanneJs B 3.3.9
-- BASES LCO The lCO on the neutron flux monitoring channels ensures that adequate information is available to verify core reactivity conditions while shut down.
Two neutron flux monitoring channels are required to be OPERABLE. If only one section of a neutron flux monitoring channel (source range or ~ide range) is functioning, the neutron flux monitoring channel may be considered OPERABLE if it is capable of detecting the existing reactor neutron flux. For example, with the source range count rate indicator functioning properly within its range, and in reasonable agreement with the other source range, a neutron flux monitor channel may be considered OPERABLE even though its wide range indicator is not functioning. The source range nuclear instrumentation channels, NI-1 and NI-2,*provide neutron flux coverage extending an additional one to two decades below the wide range channels for use during refueling, when neutron flux may be extremely low. This LCO does not require OPERABILITY of the High Startup Rate
Trip Function or the Zero Power Mode Bypass Removal Furiction.

Those functions*are addressed in LCO 3.3.1, RPS Instrumentation.
APPLICABILITY In MODES 3, 4*, and5, neutron flux monitoring c;:hannels must be OPERABLE to monitor core *power for reactivity changes. In MODES 1 and 2, neutron flux monitoring channels are address~d as part of the RPS in LCO 3~3.1. The requirements for source range neutron flux monitoring in MODE 6 are addressed in LCO 3.9.2, "Nuclear Instrumentation."
Palisades Nuclear Plant B 3.3.9-2 05/30/99
Neutron Flux Monitoring Channels B 3.3.9
BASES ACTIONS A.1 and A.2 With one required channel inoperable, it may not be possible to perform a CHANNEL CHECK to verify that the other required channel is OPERABLE. Therefore, with one or more required channels inoperable, the neutron flux power monitoring Function cannot be reliably performed. Consequently, the Required Actions are the same for one required channel inoperable or more than one required channel inoperable. The absence of reliable neutron flux indication makes it difficult to ensure SOM is maintained. Required Action A.l, therefore, requires that all positive reactivity additions that are under operator control, such as boron dilution or PCS temperature changes, be*
halted immediately, preserving SOM. SOM must be verified periodically to ensure that it is being maintained. The initial Completion Time of 4 hours and once every 12 hours thereafter to perform SOM verification takes into consideration that Required Action A.l eliminates many of the means by which SOM can be reduced. These Completion Times are also based on operating experience in performing the Required Actions and the fact that plant conditions will change slowly. SURVEILLANCE SR 3.3.9.1 REQUIREMENTS SR 3.3.9.1 is the performance of a CHANNEL CHECK on each required channel every 12 hours. A CHANNEL CHECK is normally a comparison of the parameter indicated on one channel to a similar parameter on other channels. It is based upon the assumption that instrument channels monitor.ing the same parameter should read approximately the same value. Significant deviations between instrument channels could be an indication of excessive instrument drift in one of the channels or of something even more serious. CHANNEL CHECK will detect gross channel failure; thus, it is key to verifying that the instrumentation continues to operate properly between each CHANNEL CALIBRATION.
Agreement criteria are determined by the plant staff and should be based on a combination of the channel instrument uncertainties including indication, and readability. If a channel is outside the criteria, it may be an indication that the transmitter or the signal processing equipment has drifted outside its limits. If the channels are

within the criteria, it is an indication that the channels are OPERABLE.
Palisades Nuclear Plant B 3.3.9-3 05/30/99
Neutron Flux Monitoring Channels B 3.3.9
BASES SURVEILLANCE REQUIREMENTS SR 3.3.9.1 (continued)
The Frequency, about once every shift, is based on operating experience that demonstrates the rarity of channel failure. Since the probability of two random failures in redundant channels in any 12 hour period is extremely low, CHANNEL CHECK minimizes the chance of loss of protective function due to failure of redundant channels. CHANNEL CHECK supplements less formal, but more frequent, checks of channel OPERABILITY during normal operational use of displays associated with the LCO required channels. SR 3.3.9.2 SR 3.3.9.2 is the performance of a CHANNEL CALIBRATION. A CHANNEL CALIBRATION is performed every 18 months. The Surveillance is a complete check and readjustment of the neutron flux channel from the preamplifier input through to the remote indicators. This SR is modified by a Note which states that it is not necessary to calibrate neutron detectors because of the difficulty of simulating a meaningful signal. Wide range and
source range nuclear instrument channels are not calibrated to indicate the actual power level or the flu~ in the detector location. The circuitry is adjusted so that wide range and source range readings may be used to determine the approximate reactor flux level for comparative purposes.
This LCO does not require the OPERABILITY of the High Startup Rate trip function or the Zero Power Mode Bypass removal function. The OPERABILITY of those functions does not have to be verified during performance of this SR. Those functions are addressed in LCO 3.3.1, RPS Instrumentation. This Frequency is the same as that employed for the same channels in the other applicable MODES. REFERENCES 1. 10 CFR 50, Appendix A, GDC 13
2. FSAR, Chapter 7

3. FSAR, Chapter 14' ,
Palisades Nuclear Plant B 3.3.9-4 05/30/99
ESRV Instrumentation B 3.3.10
B 3.3 INSTRUMENTATION B 3.3.10 Engineered Safeguards Room Ventilation (ESRV) Instrumentation BASES BACKGROUND This LCO addresses the instrumentation which provides isolation of the ESRV System (Ref. 1). The ESRV Instrumentation high radiation signal provides automatic damper closure, using two radiation monitors. One radiation monitor is located in the ventilation system duct work associated with each of the Engineered Safeguards (ES) pump rooms. Upon detection of high radiation, the ESRV Instrumentation actuates isolation of the associated ES pump room by closing the dampers in the ventilation system inlet and discharge paths. Typically, high radiation would only be expected due to excessive leakage during the recirculation phase of operation following a Loss of Coolant Accident (LOCA). The ESRV System is addressed by LCO 3.7.13, "Engineered Safeguards Room Ventilation (ESRV)
Dampers."
APPLICABLE SAFETY ANALYSES The ESRV Instrumentation isolates the ES pump rooms in the event of high radiation in the pump rooms due to leakage during the recirculation phase. The analysis for a Maximum Hypothetical Accident (MHA) described in FSAR, Section 14.22 (Ref. 2), assumes a reduction factor in the potential radioactive releases from the ES pump rooms due to plateout following automatic isolation.* However, no specific value is assumed in the MHA for the actuation of the isolation. The results indicate that the potential MHA offsite doses would be less than 10 CFR 100 guidelines.
The ESRV Instrumentation satisfies the requirements of Criterion 3 of 10 CFR 50.36(c)(2). LCO The LCO for the ESRV Instrumentation requires both channels to be OPERABLE to initiate ES pump room isolation when high radiation exceeds the trip setpoint. The ESRV Instrumentation Setpoint is specified as ~ 2.2E+5 cpm. This setpoint is high enough to avoid inadvertent actuation in the event of normal background radiation fluctuations during t~sting, but low enough to isolate the ES pump room in the event of radiation levels indicative of a LOCA
and excessive leakage during recirculation of primary coolant through the ES pump room.
Palisades Nuclear Plant B 3.3.10-1 05/30/99
ESRV Instrumentation B 3.3.10 BASES APPLICABILITY The ESRV Instrumentation must be OPERABLE in MODES 1~ 2, 3, and 4. In these MODES, the potential exists for an accident that could release fission product radioactivity into the primary coolant which could subsequently be released to the environment by leakage from the ES systems which are recirculating the coolant. While in MODE 5 and in MODE 6, the ESRV Instrumentation need not be OPERABLE since the potential for radioactive releases is minimized and operator action is sufficient to ensure post accident offsite doses are maintained within the 10 CFR 100 guidelines. ACTIONS The most common causes of channel inoperability are outright failure of loop components or drift of those loop components which is sufficient to exceed the tolerance provided in the plant setpoint analysis. Loop component failures are typically identified by the actuation of alarms due to the channel failing to the "safe" condition, during CHANNEL CHECKS (when the instrument is compared to the redundant channels), or
during the CHANNEL FUNCTIONAL TEST (when an automatic component might not respond properly). Typically, the drift of the loop components is found to be small and results in a delay of actuation rather than a total loss of function. Excessive loop component drift would, most likely, be identified during a CHANNEL CHECK (when the instrument is compared to the redundant channels) or during a CHANNEL CALIBRATION (when instrument loop components are checked against reference standards).
A Note has been added to the ACTIONS to clarify the application of the Completion Time rules .. The Conditions of this Specification may be entered independently for each channel since each channel serves to isolate a different Engineered Safeguards Room. The Completion Times of each inoperable channel will be tracked separately, starting from the time the Condition was entered .
Palisades Nuclear Plant B 3.3.10-2 05/30/99
ESRV Instrumentation B 3.3.10
_BA_S_E_S~~~~~~~~~~~~~~~~~~~~~~~~~~~~-
ACT IONS A.1 (continued) Condition A addresses the failure of one or both ESRV Instrumentation high radiation monitoring channels. Operation may continue as long as action is immediately initiated to isolate the ESRV System. With the inlet and exhaust dampers closed, the ESRV Instrumentation is no longer required since the potential pathway for radioactivity to escape to the environment has been removed. The Completion Time for this Required Action is commensurate with the importance of maintaining the ES pump room atmosphere isolated from the o~tside environment when the ES pumps are circulating primary coolant. SURVEILLANCE SR 3.3.10.1 REQUIREMENTS Performance of the CHANNEL CHECK once every 12 hours ensures that a gross failure of instrumentation has not occurred. A CHANNEL CHECK is normally a comparison of the parameter indicated on one channel to a similar parameter on other channels. It is based on the assumption that instrument channels monitoring the same parameter should read approximately the same value. Significant d~viations between the two instrum~nt channels could be an indication of excessive instrument drift in one of the channels or of something even more serious. CHANNEL CHECK will detect gross channel failure; thus, it is key to verifying the instrumentation continues to operate properly between each CHANNEL CALIBRATION. Agreement criteria are determined by the plant staff, based on a combination of the channel instrument uncertainties, including indication and readability. If a channel is outside the criteria, it may be an indication that the transmitter or the signal processing equipment has drifted outside its limits. The Frequency, about once every shift, is based on operating experience that demonstrates the rarity of channel failure. The CHANNEL CHECK supplements less formal, but more frequent, checks of channel OPERABILITY during normal operational use.of the displays associated with the LCO required channels .
Palisades Nuclear Plant B 3.3.10-3 05/30/99
ESRV Instrumentation B 3.3.10
_BA_S_E_S_________________________________________________________
SURVEILLANCE SR 3.3.10.2 REQUIREMENTS
.(continued) A CHANNEL FUNCTIONAL TEST is performed on each ESRV Instrumentation channel to ensure the entire channel will perform its intended function. A successful test of the required contact(s) of a channel relay may be performed by the verification of the change of state of a single contact of the relay. This clarifies what is an acceptable CHANNEL FUNCTIONAL TEST of a relay. This is acceptable because all of the other required contacts of the relay are verified by other Technical Specifications and non-Technical Specifications tests at least once per refueling interval with applicable extensions.
Any setpoint adjustment must be consistent with the assumptions of the setpoint analyses. The Frequency of 31 days is based on plant operating experience with regard to channel OPERABILITY, which demonstrates that failure of more than one channel of a given Function in any 31 d~y. interval is a rare event. SR 3.3.10.3 CHANNEL CALIBRATION is a complete check of the instrument channel including the sensor. The Surveillance verifies that the channel responds to a measured parameter within the necessary range and accuracy. CHANNEL CALIBRATION leaves the channel adjusted to account for instrument drift between successive calibrations to ensure that the channel remains operational between successive tests. CHANNEL CALIBRATIONS must be performed consistent with the setpoint analysis. The Frequency is based upon the assumption of an 18 month calibration interval for the determination of the magnitude of equipment drift in the setpoint analysis. REFERENCES 1. FSAR, Section 7.4.5.2
2. FSAR, Section 14.22
Palisades Nuclear Plant B 3.3.10-4 05/30/99
ATTACHMENT 3 PALISADES NUCLEAR PLANT SECTION 3.3 - INSTRUME~ITATIONS CTSMARKUP AND DISCUSSION OF CHANGES
3. 3.1
3.~
3.3.l Reoct6r Protective System !RPSl Instrumentot1on INSTRUMENTATION lHST~MS I RAI I3.3.1-6 RAI Tob le 3.3.1-1 3.3.1-1 ond Footnote o [.l"i11;g':2I With one RPS trip unit or associated instrument channel inoperable for [ Cond A l one or more functions: [ RA A.1 1 a)* Place the affected trip unit in the tripped condition within
7 days.
1)'<17.)':31 With two RPS trip units or associated instrument channels inoperable for [ Cond. E ] one or more functions: [RA E.1 l a)
< Add RA E NOTE ~
Place one inoperable trip unit in the tripped condition within 1 hour, and RAI 3.3.1-1 [ Cond.F] b) If two Power Range Nuclear Instrument channels are inoperable, [ RA F.1 ] limit power to s 70% RATED POWER within 2 hours, and [RA E.2 l c)* Restore one-RPS trip un;t and associated instrument channel to OPERABLE status within 7 days. SEE 3.3.2> 1)1:17 .):61 [ Cond. G ] [ RA G.1 l [ RA G.2.1 l RAI 3.3.1-1 Amendment No. ~. 186 Verify PCS boron concentrotion is ot REFUELING BORON CONCENTRATION. Add Cond. B ond C >-@.6 Add Cond. 0 >@
< Add Required Actions 8.1, ond C.1 < Add Required Actions 0.1, ond 0.2 M.1 Page 1 of 7
3.3.l
3.17 INSTRUpr!ENTATIQN SYSTEMS Table [)i;)(ij(jj.1-1 )
Instrymentatjgo Oper1tioq Regytremeots fgr Reactor ProtecS;1 ye System A. 5 Required Permissible RPS Operational E1i10,tiao1l LIOU 'blDD~]S Bypasses
1. (#nu?rrj( 7
..____...__....._,_____.__......._--<..,_......._____.__.....c..-____.__...._____..d_o.....&.n?<_*__.KsEE 3.3.2)
[1] 2. Variable High Power (Trip ) 4 None. [2]3. High Start Up Rate 4 Below 10....%1cl or fT3.3.H l above 13% RATED POWER. l£ootnote lb:j [9]4. ThenRal Margin/ Low Pressure 4 (b)~ 3~~l-1 A.5 3.3.1-6 (8]5. High-Pressurizer 4* Pressure [3Hi. LQW PCS Flow 4 (b)(K@-@
(10] 7. Loss of Load<d> 4 Below 17% RATED POWER.
[4] 8 *. Low *A* Ste111 4 None. Generator Level [5] g. Low *e* Stea 4 None. Generator Level [6~ 10. Low *A* Steam 4 (b)rn-@ Generator Pressure (7111. Low *e* Steam 4 (b)~* Generator Pressure (11]12. 4 None. 13. SEE 3.3.2) 14. [12J~ Two OPERABLE Wide Range Nuclear Instru11ent channels are required if the Zero
Power Mode bypass is used. ~
[(ell~- Be 1ow 10-4%-<cl RA~~POWER l#d aYsHujDOij/eoRPR co}iCENjl(AT Ip! foj' thi'CoLPi ISHU10QWiVcond'f tl B ass shall be automat1call removed when WIDE RANGE POWER is >10- 4 % RTP. RAI 3.3.1-1 3.3.1-6
<Add Footnotes A & B> Amendment No. 118, 139, 136, 162 3-65 Page 2 of 7
3. J. I 1
4.17 INSTBUMENTATIOH SYSTEMS Syryeillance Regyirement TEST~
Surveillance testing of instrument channels, logic channels, and control channels listed in Tables 3.17.1 through 3.17.6 shall be performed as specified in Tables 4.17.1 through 4.17.6, respectively
. iI Amendment No. 37 1 69, &9, 1§2, 162

4-75
~"1e... 3 October 26, 1994 °.f 7
3.3.1
l" 4 .17 I NSTRUMEtUATI ON SfiTEMS TE SIS ~.....--,,.___-..
Table lf'l~(E.1-1 ) Instrumentatjon Syryeillance Requirements for Reactor protectjve System CSR 3.3.1.lJ cs~H;NRrL.51 [SR 3.3.1.8J CHANNEL FUNCTIONAL CHANNEL Fynctjonal Unit CHECK TEST CALIBRATION
1. [)4nuif Trji) / / 7 /ftA 7 7 )!a)/ 7 7 N~sEE 3.3.~
[!] 2. Variable High Power 12 hours days (b, c, &d) [2] 3. High Start Up Rate 12 hours (a) 18 months C*l (g] 4. Thermal Margin/ 12 hours 18 months Low Pressure [8] 5. High Pressurizer 12 hours days 18 months Pressure RAI [3] 6 .. Low PCS Flow 12 hours days 18 months 3.3.1-6 [10]7. Loss of Load NA (a) 18 months
[4] 8.
[5] 9. Low 11 A* SG Level , Low "Bil SG Level [6] 10. Low "A* SG Pressure (7) 11. Low "Bil SG Pressure 12 hours 12 hours 12 hours 12 hours days days days days 18 months 18 months is months 18 months (11) 12. High Containment Pressure NA days 18 months 13. 14.
~h?2-<SEEi3.2>
[SR 3.3.1.5] 15. Thermal Margin Monitor; Verify constants each 92 days. . A.1 1 3 .~~~ 8 [SR 3.3.1.21 16. e Ma in ni r: Verify Control Room Temperature s 90"F each 12 hours. [SR 3.3.1. 71 (a) Once w;thin 7 days prior to each reactor startup. <Add SA 3.3.1.3 Note& [SR 3.3.1.3] (b) Calibrate with Heat Balance each 24 hours, when> 15% RATED POWER.~ * [SR 3.3.1.6] (c) Calibrate Excores channels with test signal each~ [SR 3.3.1.81 (d) CHANNEL CALIBRATION each 18 months. CSR 3.3.1.8] (e) Include verification of automatic Zero Power Mode Bypass removal. Apply to ITS Function 9r '-"
\~
Amendment No. ti&, -i-3-9, 4-76
~. -rse, ~. 1-64.~.
Poge 4 of 7 186
3'. 3'. I 1
2.0 2.1 Safety Limjt - Reactor Core The Minimum DNBR of the reactor re shall be maintained greater than or equal to the DNB correlation safet limit.
Correlation Safetv Ljmit XNB 1.17 ANFP 154 HTP 1. 41 is applicable in HOT STANDBY a POWER OPERATION .
2.2.1 1n the reactor ..
2.3 .llo.Lllii+-11-1.u:a...111.11.1..&.11.1~ol.ll..:li1111..11&1U1~......,;;,.."1.liio~~~~~&..11.~:i.lll (RPS) be as tated in Table 2 3 .1. limiting Saf Syst* Settin of Table 2.3.I re applicable w n the associated RPS hannels are req red to be OPE LE by Specificat on 3.17.1. 2.3.l If RPS tnstrwnent set ng ts not within e allowable ttings of Table .3.1, innedtately lace the 1nstrume inoperable d complete corrective action as directe by Specification 3.17.L
Amendment No. a.!-, a., 43, -Hi, .W, ~. ~. 174 2-1 October 31, 1996 Po..~e 5 o.f 7
? . 3. I
~
G .
13.~.H R
TABLE ).l-I 4.IO.d REACTOR PROTECTIVE SYSTEM TRIP SETTING LIMITS
~. 3' l-Z..)
Variable High s15' above core power, Power with a minimum of s3~ RATED POWER and a maximum of s106.5' RATED POWER. C3J 2. PCS Flow ~95' Full PCS Flow. ceJ 3. High Pressure Pressurizer s2255 psia. s2255 [g] 4. Thermal Margin/ (a) (a) Low Pressure 0)sJI 5. Steam Generator ~25.~ Low Water Level Narrow Range al 11 6. Steam Generator ~500 psh. ~500 psia.
$3~g.
Low Pressure
Li 1] 7. Containment High :s3.70 psig.
Pressure c-o,3.1- z.] (a) The pressure setpo1nt for the Thermal Margin/low Pressure Trip, Ptrip* is the higher of two nlues, Pinn and Pv., both in psia: Pinn
1750 Pv*

2012(QA)(QR1 ) + 17 .O(T1n}

9493 where:
QA * -0.720(ASI) .+ 1.028; when °0.628 :s ASI < -0.100 QA * -0.333(ASI) + 1.067; when -0.100 :s ASI < +0.200 QA * +0.37S(ASI) + 0.925*; when +0.200 :s ASI :s +0.565 ASI .. ASI
Measured ASI o.o when Q_~ 0.0625 when Q < 0.0625 QR,--

0*.412(Q) + 0.588; when Q :s 1.0 QR, * . Q; when Q > 1.0 Q

Core Powtr/RATED POWER Tin

Max1ntt111 primary coolant inlet temperature, in *F.
ASI, Tin, and Q are the existing values as measured by the associated instrument channel. ' -
Amendment No. a.I-, 89, -Ha, Ha, ~. 162 October 26, 1994 2-2
j,:?. I
/::--'\ \8.:l, 1
4.18.l 4.18.1.l The detection system shall be demonstrated operable:
a. -By erformance of a Channel C eek prior to its use fol a erat1on and at least once per 7 days during power r, quired for the functions isted in Section 3.11.1.
At least once per refuelf g by performance of a Ch which exempts the neutro detectors but includes components. he incore alarm system is ~emonstrated operable thr ugh use of the datalogger ~or alarm. The ~o alarm is demonstratea-operable*o( per refueling by pert~ ance of a Channel Check. 4.18.2 Excore Mon1tor1ng System
a. and excore monitoring llowable power leve determin using excore and incor detector readings state n r equilibrium condition *
[5R.3.3.1.4] b. [S'll. 3.3,1,1./ .Oo+e. "0 (..See o..lso 3 ,'?.)
c. Tf lt shall be compar. d to the tncore meas red Quadrant Power T lt. If the dtfferenc ts greate than 2', ta excore monitoring ystem shall be recalt rated.
RA1. 33.1-B Amendment No. 37, 68, 118, 162 October 26, 1994 4-83
3 .17 INSTRUMENTATION SYSTEMS ' Re.a_i:..+o-r Pr"o-tec.+wc:. 5"1 s+e.M (~f's)
Le 1t*c. t..-.t"'I 'Tv-** p .:Cr1 ,*+t C.:t; o"' 3.17.1 [J..oAJi) c.] 3.17.1.l With one Manual Reactor Trip channel inoperable: [gA C. l] a) Restore the channel to OPERABLE status prior to the next reactor startup. 3.17.1.2 (With \ one RPS trip unit or associated instrument channel inoperable*{or* one or more functioA,_s: * . ~ I '\ '*, a)* P.lace the affecte' trip unit in the da)\-
3.17.1.3 With two RPs, trip units or sociated instrum t channels one or more functions:
a) Place one
" " hour, and perable trip u t in the trippe condition within b) --~f two Power Ran Nuclear Instr ent channels ar inoperable, lf~ power to~ 7 RATED POWER w in 2 hours, a *c)* Re!~~~one RPS trip u t and associated instrument channel to *-... PERABLE status within 7 da s. .
[2'J.Jb A] 3.17.1.4 With one RPS Matrix logic channel inoperable: Add. Co~b
lY f}A A* jJ a) Restore the channel to OPERABLE status within 48 hours.
[C.01-1D '3] 3.17.1.S With one RPS Initiation Logic channel inoperable: c.~F\ B. D a) De-energize the affected clutch power supplies within 1 hour. [0:l~\) E.] 3.17.1.6 a) The reactor shall b) e.'<\~ ~"' IMOV'E' 1-1\11.." () t'lt~ ~ ....+o( t-od C-£pa.b/<. Amendment No. ~. o+ 'oe."'"'"t w1+tu:fv-~"' M':' fcs bo.-o ... en ... c.
,*s <11" 1<<-Fua.11.JU. 6tJiiJ1.J Cc-uc:=.-J..J1~lfno11
3. 17 INSTRUMENTATION SYSTEMS i
Table 3.17.1 Instrumentation Ooeratinq Requirements for Reactor Protective System Required RPS Functional Unit Channels (}. 3. z.J 1. Manua 1 Trip 2
2. Power
\
3. 10*~1c1 or l 3'- RA TED POWER*]

4. 4 2 (b) , (c I

5. 4 2 None.

6. 4 (b) & (c).

7. 4 Below 17i RATED P I

8. Low "A" Steam 4 2 None.
IGenerator Level
9. Low "B" Steam 4 2 Generator Level

10. Low "A" Steam 4 2 Generator Pressure

11. Low "B" Steam 2 Generator Pressure

12. High Containment Pressure _ _ _ ___i.____ _ _2_ _ _ _ None.
~.... -,
j -\,f, 2J 13. RPS Matrix Logic 6
;3,J,'Z.j I 14, RPS Initiation Logic 4 (a) o OPERABLE Wide Range Nuc P er Mode bypass is used.
(b) Bel 10~1* RATED POWER and at SHUTDOWN BORON CONCENTRA ON for the *coLD SHUT N condition. (c) POWER PHYSICS TESTING, set oint may be increased fr m 10*~ to S TDOWN BORON CONCENTRATION s not required. * (d) Amendment No. 118, 139, 13&, 162 October 26, 1994
. 3-65

4 .17 INSTRUMENTATION SYSTEMS TESTS Syryeillance*gegyjreroent\
~ ' --~~~~~~~~~~~~~~~~---...
Surv illanct te'~ng of inst~~t chann1~, channt listed ; Tables 3.17. through 3. logic chan~*~*nd ~ntrol shall be p~rmed~
\Specifi in Tables .17.1 throug 4.17.6, resp tively i
I I J,
I Alltndlltnt No. 37, &Q, 691 1§2, 162 October 26, 1994 4-75
; ~ -
4.17 INSTRUMENTATION SYSTEMS TESTS Table 4.17.l Instrumentation Syrveillance Requirements for Reactor Protective System CHANNEL CHANNEL FUNCTIONAL CHANNEL Fynct1onal Un1t CHECK TEST CALIBRATION jeJ,:i',2.-zJ 1. Manual Trip NA (a) NA
2. ower 12 hours 31 days (b, d)

3. 12 hours (a)

4. 31 days

5. 31 days

6. 31 days

7. (a) 18 months

8. 31 days 18 months

9. 31 days 18 months

10. *A* SG Pressure 18 months

11. 31 18 months

12. High Containment Pressur NA 31 18 months
/. ...,
J_f?.. *n,z.. lj 13. RPS Matrix Logic NA~days L.I 11 NA (}iJ.~, z.. D 14. RPS Initiation Logic NA days NA
15. Thennal Monitor; Verify constan s each 92 days.

16. Thena Monitor: Verify Control
each 12 ho
[SR 3.1.2. ."fl (a) (b) (c) (d) {e)
Amendment No.
4-76
~. ~ * .i.a.i, ~. ~. -t-i4, ~.
3.3.3
~~~-----....
ITS 3.17 INSTRU!l£NTATIQN SYSTEM) Engineered Sof ety F eotures (ESFl Instrumentot1on (A.b-,
- t Applfcabflftv &;tf QO . . ,__ . -<( Add ACTIONS Note )---3}
3.17.2.1 ECond Al 3.17.2.2 W1th one ESF fnstruaent channel inoperable for one or more functions, [NOIEJ---1!xcept SIRWT LevelF
ERA A.lJ a) Pl ace the trip unit for each affected ESF function 1n the tripped condition within 7 days.
ECond Bl 3.11:2.J With two ESF instrument channels inoperable for one or more functions, [NOTEr-1except S IRWT Levell: (Add Cond. B Required Actions Note)--@ ERA B.ll a) Place one channel trfp unit for each affected ESF functf on in the tripped condition within 8 hours, and ERA B.21 b) Restore one channel to OPERABLE status within 7 days. [Cond Cl3.17.2.4 With one SIRWT Level channel inoperable: [RA C.ll a) Bypass the level switch within 8 hours, and ERA c.21 b) Restore the _channel to OPERABLE status within 7 days. [Cond Dl 3 .17 *2. 5 ERA O.ll 1) The reactor shall be placed in ~~.:..:...:...::...z..:..:~ and ERA 0.2] b) M.2
3-66 Alllendllent No. ~. 180 Poge 1 of 8
:-: 3.17 INSTRUMENTATION SYSTEMS Table~--
~
Instrumentation Operating Requirements for Engineered Safety Features MifS'mum 'ti\, 1 ORiit\Sl.E Functional Unit ~
1. Safety Injection Signal (SIS}
Initiation 2 l None. l None.
c. CHP Signa SIS Initiation (SP Relay O tput}
!1". ::<...~ d. Pressurizer Pressure [Leo 1.J.3] 4 Instrument Channels
2. Recirculation Actuation Signal (RAS}
{j. a.-J a.
b. c. Manu l RAS Log Initiation SI RWT Leve 1 Switches ~J.3.'3] 4 l 1
§--8
3. Auxiliary Feedwater Actuation Signal (AFAS}

a. Manual Initiation 1 None

b. 1 None
~.Ci..J c. 11 A11 Steam Generator Leve 1 B. b] d. 11 811 Steam Generator Level Amendment No. ~
3-67
3.3.3
ITS 3.17 INSTRUMENTATION SYSTEMS Engineered Sof etlj Feotures <ESFl lnstrumentot1on SEE 3.3.~
AQpl1cab111ty Action ....,4-----( Add ACTIONS Note re 3.3.~ 3.17.3.1 [Cond AJ 3.17.3.2 - With one Isolation Function instrument channel inoperable for one or 110re functions: ERA .A.lJ a) Place the trip unit for each affected Isolation Function in the ~ tripped condition within 7 days. [Cond Bl 3.17.3.3
With two Isolation Function instru111ent channels inoperable for one or 110re functions: ( Add Cond. 8 Required Actions Note~ \~
[RA B.lJ a) Place one channel trip unit for each affected Isolation Function in the tripped condition within 8 hours, and [RA B.2J b) Restore one channel to OPERABLE status within 7 days. 3.17.3.4 [Cond El 3.17.3.5 [RA 0.1 & E.lJ RA! 3.3.4-q M.4
3-68 Amendment No. 162 Poge 3 of 8
3. 17 INSTRUMENTATION SYSTEMS* ~
. ~J.3.1-1)
Table~ . Instrumentation Operating Regyirements for
* . Isolation Functions 1}3.3'.3-\]
1. Containment High Pressyrt (CHP) a'. ,pip l~c Trains ~ 1 \ None~*~ 3.~,4.
~
[5 .;(! b. Containment Pressure ~ r-::
.~,;I *,~* o.._.*
Switches - Left Train
c. Containment Pressure Switches - Right Train J r- , ..,-,
4 /llo3 .*. ".Jj
*~
2. Containment High Radiation (CHR)
\ a. 1
b. CHR Lo ic Trains 1
c. Containment Area 4 (1c.o 3.3.;'] ~ ~-
Radiation Monitors
3. Steam Generator Low Pressyre (SGLP)

a. Ma nu 1 set

b. 1 J(h0on~j
~
[?,a...] c. "A" Steam Generator Pressure 12.b] I d. "B* Ste111 Generator
- I Pressure No""'
b.i West Room nitor 1 None. Amendment No. 162 October 26, 1994
3-69
/
~<:
I I J 3 .16 3.16 he Engineered Safety eatures (ESF) syste l 'mi ts shall be as sta d in Table 3.16. A Specif ation 3.16 is applic le when associated For Function instrumentation is r uired to be OPERABLE Specifica ion 3.17.2 or 3.~7.3.
\~
3.16.l I an ESF instrumen setting is not withf the allowable settin Ta e 3.16, i11111ediate declare the instru nt inoperable and co corr tive action as d'rected by specificati n 3.17.
~
TABLE~ Engjneered Safety Features System Instrument Settings l.!.* ~j [ffe.a.,5.b]
~.c.-J 1.
2. Instrument Channel Pressurizer Low Pressure Containment High Pressure Allowable Valye
~
3.70 - 4. 1593 psia
~ ~
psig I
3. Containment High Radiation s 20 R/h ~AI:
rz.().. ,2. b] 4. Steam Generator Low Pressure , ~ 500 psia 3:S.'3-l
~.a/t~ !. Steam Generator Low Level ~ 25.~
Narrow Range (j~~J 6. SIRW Tank Low Level 21 - 27 inches Above Tank Bottom
7. E~red afegaards P~om ~
Ven lation b Radiatio Amendment No. 89, 162 October 26, 1994
3-63
3.~.3
i 4 .17 IPfSTIM!EfTATIQ! SY$IE!'S TESTS Syrv1tll1nc1 81gy1re110S Surv1illanc1 tutlng of lnstn11n1nt ch1nn11s, lo9tc channels, ind control ch1nn1l1 listed 1n Tabl1s 3.17.1 through 3.17.6 shall be p1rfonn1d as sptcif1td \n T1bl1s 4.17.1 throu9h 4.17.6, r1specttv1ly Amtndllftt No. ;;, 111 191 iii, 16Z October t&, 1994 4.75

4 .17 INSTRUMENTATION SYSTEMS TESTS r...,
Tablec~
~(.~.J,3,/
O;
.,.~~
Instrumentation Surveil~Requirements for Engineered Safety Features CHANNEL
;;;cSt.3.l.3.iJ r~.; ~ 3 ~..,
L;'"" 91 "" * ..;i-1 CHANNEL FUNCTIONAL CHANNEL Functional Unit CHECK TEST CALIBRATION
1. Safety Iniection Signal (SIS)

a. NA NA

b. NA NA

c. 18 months NA 11
- .~1 d. Pressurizer Pressure ~
Instrument Channels 12 hours ~days 18 months
'2.
2. R~circulation Actuation Signal (RAS)

a. Manual nitiation
~
18 months NA\r@
. .J,J.~
b. 18 months NA K'Al:
Ji. cg c.
~
SIRWT Level Switches NA 18 months ~J.J-1
3. Ayxiliary F~edwater Actuation Signal (AFAS)

a. Ma nu nlt ia 10n 18 months

b. ic I3 ,a.] c. "A" SG Level 12 hours~ days 18 months ~1'"I..
3,3,3-1 C:! b] d. "B" SG Level 12 hours ~~ - days 18 months L. Sze
,~.
__::.~~~~~~~~~~~~~-,-~~~~~~~~~~~-:--~~~~---- a) Test n~l and emergency power unctions using test circuit~each 92 days. Verify a l automatic actuations d automatic resetting of lo~pressure bloc
'---~h 18 nths.
Amendment No. ~. ~. ++!-, 4-77
- i _,,. _,,
4. 17 INSTRUMENTATION SYSTEMS TESTS Table~"-'-:.:.~
Instrurnentat1on Syrve111ance Regyirements for Isolatjon Fynctjons --r<
,...f<: -I/ \ ,- ~e3,J,J',~ \.7~J.1.3, D CHANNEL r-0~.3',3.3'-3]
CHAHNEL FUNCTIONAL CHANNEL Fynctiontl Un1t CHECK TEST CALIBRATION
1. ~2ot1io111ot ~jgb ec111uc1 (CHP)

a. c£_HP ~gic Trtins ~ 18 months \ NAf~
~ ,C-_;,
l= I b. Conttin.. nt Pressure Switches - Left Trtin NA
~d~s 18 months ~/>..I 3.'33-1 L§.bj I
c. ContaiollM!nt Pressure Switches - Right Train NA days 18 months

2. ~got1io111ot ~igb B1di1tign (CHR)

1. ("****1J:1t11tion 18 mont NA *.
b.lcHR Lo 1 Trains 18 110nths NA
~ ,- I .,h.a.J I c. Contain..nt Area Radiation Monitors 12 hours ~d1ys 18 11anths !.'3.1-1
3. ~t111 G1n1c1tgc LID! ec111uc1 (SGLP)

a. *Manua Actuation 18 months NA

b. SGLP Log c Tratns 18 110nths NA
~.a]I c. *A* Steam Generator* 12 hours ~ d~s 18 110nths Pressure L(. e~-r 18 1110nths IS.l*I j1.b] I d. *a* St1111 Glllerator 12 hours ~ days Pressure
4. \__

a. days

b. West R Mont tor Allendlltnt No. ~. ~. 171 4-78
3.3.4
ITS 3.17 INSTRUM;NTATIQft SYSTEMS Action
.......__-'(Add ACTIONS Note )--3)
[Cond A] 3.17.2.1 With one ESF manual control channel or ESF lo ic channel inoperable for one or 110re functions: [RA A.1] a) Restore the channel to OPERABLE status within 48 hours *
(sEE 3.3.3
[Cond BJ 3.17 .2.5 [RA 8.1] RAJ 3.3.4-9 [RA B.21
\
I
3-66 AlllendlAent No *. *2. 180 Page 1 of 7
~
3.17 *.INSTRUMENTATION SYSTEMS Table~
~3.3.4-J)
Instrumentation Operating Requirements for Engineered Safety Features Permissible Operational Functional Unit Bypasses [1J 1. Safety Injection Signal (SIS)
a. Manual Initiation 2
.{ZPressurr~*r pr~srure-l:;c..11) ~
b. SIS Lo ic 2 (l~ati~ Ac~ion, ~
low essur bloc auto r et
c. CHP Signal SIS Initiation ~
CfS"~ l a)-.J)utpuij)
,*- rfr4/e 3.J.f *I 1
d. 'Pres rizer Pressure 4 2 ~ 1700 psi a ~o+e. C"-l]
Instr ent Channels PCS pressure. [3J 2. Recircylation Actuation Signil (RAS) 3.:!s I
a.
b. Manual Initiation RAS Logic
c. ~IRWT '§vel Switches 2
2 4\
. _ __ _ _ _;i,___ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ ~3 ~
None.
~---------------------\,,~-N-o_n~
[YJ 3. Auxiliary Feedwater Actuation Signal (AFAS)
*~
a. Manual Initiation 2

b. AFAS Logic 2

c. "A" 2 None.

d. "B" enerator Level 2 None.
Amendment No. ~
3-67
l 3.3.4
ITS 3 .17 Two Manual Im t1at1on, bypass removal, and two oc tuat1on (SEE 3.3.3
[3.3.4) 3.17 .3 Appl1cab1lity Actign Add ACTIONS Note>--@ 1 C ESF ~ . A.5 oss removo r{--ES_F_) [Cond Al 3.17 .3.1 With one ilsorlt'1on1Funct1on manual control or IJ'Solaftonl Function logic channel inoperable for one or more functions: [RA A.lJ a) Restore the channel to OPERABLE status within 48 hours. 3.17.3.2
3.17.3.3 (sEE 3.3.3 3.17.3.4 3.17.3.5 CCond Bl (SGLPl
[Cond Cl CRA B.1 & C.lJ RAI CRA B.2 & C.2J 3.3.4-q M.4 [RA B.21 ISGLPl - - - - - - - - - - - - - - - - - 'l...M~O~.v
3-68 Amendllent No. 162 Page 3 of 7
3.3.4
~
3. 17 INSTRUMENTATION SYSTEMS
__£(?. 3.4--0 ~+
£T-3.3.+-Q .
Table~ Instrumentation Operating Requirements for Isolation Functions Permissible Operational Functional Unit Byoasses r~J L:' 1. Containment High Pressure (CHP) e.A.1 3.~."1-$ I
a. CHP logic Trains 2

b. Conta1 ment Pressure None.
Switche - Left Train 4 None.* [to] 2. Containment High Radiation (CHR) RAJ:
'3.'5.'4*5 I
a. Manual Initiation

b. CHR Logic Trains 2
2 (}@ ~
~
c. 4 2
~~ '!.,,"1 *'i'
[2] 3. Steam Generator Low Pressul'l! (SGLP)~
~
a. Manual ActuationCc) c!:~t'l'a.in ~

b. SGLP Logic Trains 2

c. "A* team Generator 2 < 550 psig Press re Steam Pressure.

d. Generator 2 < 550 psig Steam Pressure.
~jt,~ps; ()..
None. None.
Amendment No. 162 October 26, 1994

3-69 AJo To.1t:il e. 3,3.'l-1 IJD+e (d\
p'1-4 e. 4- o+ 7
4.17 INSTBtJM£MTATIQ! SYSTij1S TESTS Syrvt1111nc1 81qu1rC111nt Surv1ill1nc1 t1st1ng of instrument ch1nn1ls, lo;fc channels, 1nd control ch1nntls ltst1d tn Tables l.17.1 through 3.17.6 sn1ll be performed as sptciftld tn Tables 4.~7.1 tnro1J9h 4.17.6, r1specttv1ly Amtndmlnt No. a1, 10, 11, lia, 1sz October 26, l 994 4.75
3,3, 4- r t
ITS 4 .17 INSTRUMENTATION SYSTEMS TESTS Table~:l:~)
-), .4--1) 3 Instrumentation Surveillance Requirements for Engineered Safety Features CHANNEL CHANNEL FUNCTIONAL CHANNEL Functional Unit CHECK TEST CALIBRATION
[_1 J 1. Safety Iniection Signal (SIS} ~l...1,..1. 4-,3]
a. Manual Initiation NA 18 months NA

b. SIS Logic-t
.(21'r2$.Sun~e..- pr.,(fure. -w'-'ll NA /I}.( J.3. ** 'J (a}, ~i: ~. 3 .'l.l] NA (1In-l!iation~ctuat{2n, a.Qd-....,_ ~ow p'l-essure B'N>ck alllo rese!J) ~\\.J,).+.3]
c. 18 months NA

d. Pre surizer Pressure 31 days Inst ment Channels
[sR. 3. 3A.1] (a} [}it 1.J.1.n] Amendment No. ~' ~. ~ 4-77
3.J.4-
@i
4. 17 INSJBUMENTAJION SYSU:MS TESTS Tablec@
-0 j, 4-?0 Instrumentation Syrveillance Reayjrements for Isolatjon Fynctjons CHANNEL CHANNEL FUNCTIONAL CHANNEL Fynction1l Unit CHECK TEST CALIBRATION
1. Cont1jnroent High Pressyre (CHP) /13t ..r.J.~*JJ

a. CHP logic Tr1ins NA 18 months NA

b. ont11n11en ressure 31 d1ys months itches - Left Tr1in

c. 31 days 8 months

2. Cont1jnment High B1di1tion (CHR) ff'-~.? ... Jj

1. t11nual Initi1tion NA 18 rt0nths~ NA

b. CHR Logic Trains NA 18 1110nths NA

c. Cont inaent Area 12 31 days Radia on Monitors
1.3.4.3 J r

3. Ste11 Generator Low Pressyre (SGLP)

a. Manual Actuation NA 18 months NA

b. SGLP Logic Trains NA 18 rt0nths NA

c.
Stea11 Generator \hours 31 days 18 1110nths
d. hours 31 days 18 llOnthS 31 days 18 llOnths Allendll9nt No. ~. -t-i4, 171 4-78
------------------~-
AdJ /\ew Speci(;,c..+:~ ... ) I r5 3. 3' ~":,
3.7 1
"D1'e~e\ G1,er.~.:tor (t)f.\ - l)~d~rvol+tt~ {iJ..V)
ELECTRICAL SYSTEMS* Apolicablljty
.s'J.c.rt
Applies to the vailability of electrical power for the aper, tion of plant compon ts. To def. ne those conditions of electrical power avail to prov*ae for safe reactor operation and the continui g en neered safety features.
3. 7 .1 The primary coolant system shall not b1 hea d or maintained at temperatures above 3oo*F if the following ectrical systems are not operable:

a. Station power transformer 1*2 (24 b.

c. 2400 V engineered safe9u1r lC and 10.
d* e. f. g.
h. Two station tter1es and the d*c systems tncluding at least one battery cha er on each bus.

i. Both die 1 generators, with a minimum of 2500 gallons of fu each d1 tank and a minimum of 16,000 gallons of fuel in th unde ound storage tank.

j. Sw chyard battery and tht d*c syst.. with on* battery

k. 40 Y 1-c power pan1ls Ho. 1 and 2, breaker distribution syst ..s.
2400 V bus lE. 3.7.2 Tht require.. nts of Specification 3.7.1 may bt fied to the extent that on* of the following conditions will be all ed. If any of the provisions of thos1 exceptions art violated, t reactor shall be placed in a hot shutdown condition within 12 hours. f the violation is not corrected within 24 hours, the reactor shal bt placed in a cold shutdown condition within 24 hours. 3-41 Amendment No. ~. 161 August 12, 1994 pl).~ I o .(- I
ca I"\~;..,,.,, ~""'t J+ I~~ !20..J,.~/1. +r &;II'\ (ci-+e.) 1... r-tv ""' e ._fr;._+,;,.,. ITS 3. 8* REFUELING Qp}RAJ'IONS) Ob1ective To llt.:tnimize \be poss;J,>ility o'(.an acci~nt occur"ing dur operat-1.ons that- could~fect o~ic heafttt. and sa~y. 3.8.l Specifications T~ollowing...condit1o shall~satis oerlttons*
a. One door o t e emergency air lock s a Whenever both doors of the personnel air lock are open during refueling operations, the equipment door shall be open and the ventilating system and charcoal filter in the fuel storage building shall be operating *
d.
e. Whenever core ge011etry is being changed, neutron flux shall be continuously monitored by at least two source range neutron monitors, with each 11anitor providing continuous visual indication
-in the control roOll. When core ge011etry is not being changed, at least one source ran e neutron monitor sh 3-46
Amendment No. 34 January 27, 1978
\IS 3.8 REFUELING OPERATIONS (Continued) g.
h.
~~D ~ 3.8.2 3.8.3.
3.8.4 he venti at on system and chareoa ter in the fuel s rage building shall be opera ing whenever irradiat fuel which has deca d less than 30 days is bein handled by either of he following operati
3.8.5.
Th equipment an ge procedures to be utilized during refu ling are disc sed in the AR. Detailed instructions, e above specifi tions. and the de gn of the f 1 handling e 1pment incorpo ting built-in i erlocks and safe features p v1d1 assuran that no incid t could occur ing the refueling 1rat1ons th would resul in a hazard to ublic health an safety.(l) hen1v1r chan s are not be made in core eometry. one fl x monitor is su 1cient. TM
Amendment No. ~~. 81 May 22, 1984

fl. 3. c,,]
3 .17" 3 .17. 6 INSTRUMENTATION SYSTEMS Specification The Safety Function instruments listed in Table 3.17.6 shall be OPERABLE. ~ Apol icabil ity According to the Applicable Conditions column of Table 3.17.6. Action 3.17.6.l Wit Flux onitoring channels a) all positive reactiv ty additions i11111edia and b) HOT SHUTDOWN or below within 15 minutes, a c) Verif SHUTDOWN MARGIN within 4 hours, therea er. 3.17.6.2 With e channel of Rod Positio Indication inoperable r one or mor~ CRDMs: a)
3.17.6.3 With one or two SIRWT Temperature channels inoperable:
a) Provide 3.17.6.4 With one Main 7 days. a) Provide alter te means of flow monitoring wi hours. 3.17.6.5 Provide alternate 24 hours. 3.17.6.6. With one AFW flow indi tor for one or more flow path Deten11ine the OPERABILITY f the associated AFW flow contr ith1n 2 hours. 3.17.6.6. one flow path inoperable:. I Amendment No. 3; &7, 9& 1 98 1 1151 1181 121; 124, 129, 13&, 162 October 26, 1994 3-74
3.17 INSTRUMENTATION SYSTEMS Action (continued) the a ected Boric Acid Tank is wit in limits 3.17.6.15 ation Alarm inoperable:
w Calculate the QU RANT POWER TILT using the at 'I ~ east once each 1 hours.
' I \
Rest e the system to from LO SHUTDOWN. startup j 3.17.6.17 With one r two SOC suction valve int rlock channels inope ble: a) Place ircuit breaker for the asso iated valve operator *n "Racked Out* po ition. The breaker may be r ked in only during o eration of assoc ted valve. e Power Dependant Ins rtion Alarm channel *noperable: a) Ve *fy that each regulatin group is within the limits of Spe *fication 3.10 within 1 minutes after movem t of any re u tin rod. Stop moving f l within the Fue Pool capability is r tored, and b) Re ore monitor to ERABLE status o rovide equivale ca a ility within 72 urs. ' 3.17.6.20 With one Containment refueling Radiation Monitor inoperable: a) Stop REFUELING OPERATIONS in the containment. 3.17.6.21 If any action required by 3.17.6.1 through 3.17.6.18 .is not met AND the a sociated c pletion ti has expire or if the num r of OPERAB'b.E ch nels is le s than sp ified in the inimum OPERABL: Channels';~""- a) and b) The eactor shal e place n equip nt is not required, wi I Amendment No. 3, 67, 96, 98, 1151 1181 121, 124, 129, 136, 162 October 26, 1994 3-76
3 .17 INSTRUMENTATION SYSTEMS Table 3.17.6 (continued}
Instrumentation Ooerating Regyirements for Other Safety Functions Minimum Required OPERABLE Applicable MQ Instrument Channels Channels Condjtions
10. 2/val ve'*

11. Detector l '* 0

12. \4,., 2

13. 1

14. Cone oric Acid Tank tx_tank 0 lo Le 1 Al arm

15. Ex co 1 0 Devi a
/ 16. AXIAL S PE INDEX 2 TEO POWER*
Alarm
17. SOC S~ion Valve 2 0 Interl ks

18. Power Inserti
- \ 2 1 HOT STA~BY and ab~e ~ ,
3, I
19. Fuel Pool Area z'bl . 0 Radiation Monitor
[3.J.t.] 20. Containment Refueling ~ 0 FU~NG'QPERAfIO Radiation Monitor w irradiated fuel Cl111-1*~, ()eE A1.:fl?tl.11t~l4/. 8 in *the Containment. tl,,1'1tl fhU~Uf,kf f>+
!fjJ_ -Sii.t;L a.I 5o (a) Speci(jcations l..,.0.4 and 4'_0.4 are ~ appl i~le~ 3.3.I 3. L.
(b) Specif1ca~ns 3.0. Amendment No. 3, '7 1 9&, 98 7 11&1 118, 121, 124, 129, ll&, 162 October 26, 1994 3-78
S'1!l. '. 17 I~SIB!.:t!E~I!IIQ~ S1SI~ IESIS hbll 4.17.5 (continued)
!n,trumtnt&tlon surz1tUane1 R1aulremtnt1 for Ott\1r S hJ.y }ul\ct1ons CHANNEL CHANNEL FUNCTIONAL CHANNEL IcHi::1i1n1ct ~~rn~ IEH ~!~Ia~TIQ~
11. SWS Brt&k Ott1ctor NA 11 111onths 18 months 1z. 12 hours 31 days

13. NA 14.
15. 15.
@ i..
17. 11 *
[sR3:3.L..Q
[s2 !.3.'-.z.] [~ie J.1 "*D u. zo. (d) Contatn..nt A1fu1ltn9 Radiation 14onltor
~~point v1rlflcatlon only.J ~
24 hours 1'2.
~
M. I 31 days
ll day1 11 months
-© ?, I ,....ndlltnt No. ~. 1-64, 171 s,
Aprn 1996
3 ,3. 7 02\
-.:ri.S 3. 17 INSTB\l!ENTATIQN SYSTQ1S Sote1f1ctt1 0n
[3.3.7] 3. 17. 4 Cfo1<1D A] 3.17.4.l (colJb l-] 3.17.4.Z (co1Jb A] 3.17.4.3 [14.:b i. :u.1-1 "1+-c. cc.J [coAJD B] 3.17.4.4
[c.o~c
[_cDNI'.> A] F-] 3.17.4.5
@out:> l] 3.17.4.6 3.17.4.7
[co t=J ** col!IC)hte Act1on
b. ernate llOD1t rcotJD B] . c.
*$ijbll1t 1 rtport to th* HRC In 1ccon11nc* with Sp1c1f1c1tlon 6.6.7.
(C..OAJD G.]
d. 1 prlo~ startup f
,._.ndlllnt No ..... ~. ~. 174 October 31, 1996
1TS
3. 17 INSTRUMENTATION SYSTEMS Instrumentation Ooerating Requirements for Accident Monitoring Required Instrument Channels

1. Wide Range TH 2

2. 2. Wide Range Tc 2 3 3. Wide Range Flux 2 y 4. Containment Floor Water Level 2
~ 5. Subcooled Margin Monitor 2
(, 6. Wide Range Pressurizer Level 2 7 7. Containment H2 Concentration 2 i 8. Condensate 2 e, 9. Wide Range ....._........,'"'-~~ 2
/o, 10. Wide Range Containment Pressure 2 II 11. Wide Range *A* Steam Generator Level 2
1i 13 I~
/{
12. 13. 14. 15. Wide Range *e* Steam Generator Level Narrow Range *A* Steam Generator Pressure Narrow Range *e* Steam Generator Pressure Position Indication for each 2 2 2 I/valve Containment Isolation Valv~
) lo 16. Core Exit Thermocouples (CET) 4 Quadrant 1
17. Core Exit Thermocouples (CET) 4 Quadrant 2 _

18. Core Exit Thermocouples (CET) 4 Quadrant 3

19. Core Exit Thermocouples (CET) 4 Quadrant 4
'tO 20. Reactor Vessel Water Level (RVWL) 2 ~I 21. High Range Containment Radiation 2 Amendment No. ~. 162 October 26, 1994
3-71
3,3,8 01
[3.3.8]
3. 17 3.17.5 INSTRUMENTATION SYSTEMS Specification The Alternate Shutdown System instrumentation and controls listed in Table 3.17.5 shall be OPERABLE~ ~
Note:
Specifications~ Ll.4, ~ d;- not apply.
Aoolitibilitv ,~ ~ es~eft....the:::?CS ump~ u?i:; s'i: 30§'* fJ MODS I; 2, aj4"' g_
Spec i fi cation 3 .17. 5 app 1i Actjon RA'I. !
~;.s-01~ ~oA] 3.17.5.1 With one or more Alternate Shutdown System channels inoperable: I a L'2.
[f'A A.1_] b) Restore the inoperable channels to OPERABLE status within c~ri aJ 3.17.5.2 If any action reqij1red by 3.17.5.1 is not met AND the associated completion time has expired: ,, ~
~oDc ~~ ~4 a.v a) The reactor shall be placed in eQl;sHU?QQW9 within e.i.J
[RA b) Amendment No. H-2-, Hi, ~. 162 October 26, 1994 3-72 p<lje. \ o.f 3
35.6
~
IT'S (Av:T INSTRUMENTATION SYSTEMS
3. 17
~-3'-6 -0 i Table(~.
I CT3,,,St-t l J Instruminta~ion Ogerating Reguiremint~ for Alternate Shutdown System I [Leo] Required I Instrument or Control Channels
1. Start-up Range Flux l I

l. 2. Pressurizer Pressure l J 3. Pressurizer Level 1 l/- 4. #1 Hot Leg Temperature 1
~ 5. #2 Hot Leg Temperature 1 ~ 6. #1 Cold Leg Temperature 1 7 7. #2 Cold Leg Temperature 1
a. "A"* Steam Generator Pressure 1
<9 9. "B" Steam Generator Pressure 1
10 (1
l"Z.. n 10. 11. 12.
"A" Steam Generator Level "B" Steam'Generator Level SIRW Tank Level.
13 .. AFW Pump P-88 Flow to *A* SG 1 1 1 1 J ti 14. AFW Pump P-88 Flow to *e* SG 1
/> 15. AFW Pum~ P-88 Suction Pressure Alarm 1 ,~ 16. AFW Pump P-88 Steam Va1ve Control 1 11 17. AFW Flow Control *A* SG 1 Jg 18. AFW Flow Control *9* SG 1 19~::s:h~O
20. Tr fer Sw h, C-15 ~
I. Amendment No. 122, 13&, 162 October 26, 1994
3-73
3.:S.~
1
..).. ;5. 4. 17 !HSTB\Jfl!EHJATION SYSTEMS UST$
Table~ [r-1, 3,21-t] tnstryrnentlf ;1nl~i:!fi!l!g5~ ~!~~lm"'"t' for CHANNEL FUNCTIONAL CHANNEL Iostryment or Control TEST ~!Lla~IlQ~
1. Start-up Range Flux ~~l3.~n 18 months

2. Pressurizer Pressure NA 8 months 0~3.3. f'.~

3. Pressurizer L1v1l NA 8 months 4.
5. 11 Hot L19 Tt11C11ratur* 12 Hot L19 TlllC)traturt NA NA 18 11 months months
~rt J.1.1.4]
wit\.. No1'E.
6. 11 Cold Ltg Ttlll)traturt NA 18 months

7. 12 Cold L19 Tt11C1trature NA 18 months

8. "A" SG Prusurt NA 18 months NA. 11 months

9. *a* SG Prusurt I 10. . "A* SG Level NA I months I 11. *r SG Level NA 8 lllOnths
~ Se-e-Qho /
12. SIRW Tank Level NA . I months ?.~

13. P-88 Flow to "A" SG 11 months 18 months

14. P-81 Flow to *a* SG 11 110nth1 11 lllOnthl 18 lllOnthS

15. P-88 Low Suction Ala~ NA 18 11anths lB months

16. P-88 St1aa Yalve Control NA 11 11Gnth1 NA [sr< !J.~.'"Q

17. .AFV Flow Control **A* SQ NA 11 lilonths NA se 3,1.~.3]
tuifli ,(Jote.
11. AFV Flow Control *a* S8 NA 11 llOl'.l.ths NA

19. Transfer S*1tches. C*l50 11 11anths MA

20. Transfer S*1tch. C*lSOA NA 11 months NA
("": 3 ~ 1.-r (a) Onct w1th1n 1 da11 prior to.each reactor startup. l-51" 1 I ,4 I~ I AMndmtnt No. Ha, .ru, ~. -M4, 171
. Apc11 S, 1996 4*10
1\S
0. 3. ~ J

3. 17 .
3 . 17. 6 INSTRUMENTATION SYSTEMS G;ie.v-t-'f".:;"' \:" {v'I.. No~-,--h-'f"...,..,;..-'j-C--:-,-:--\..,-c_-\.'1-1/'1-e-:-W Scee if i cation The Safety Function instruments listed in Table 3.17.6 shall be OPERABLE. Applicability According to the Applicable Conditions column of Table 3.17.6. Action 3.17.6.1 With one or two Neutron Flux Monitoring channels inoperable: a) Stop all positive reactivity additions irrmediately, and b) Qe ~T S'@,tDOWN 01'._below ~hin 15 muutes, ~ (jA A.l._] c) Verify SHUTDOWN MARGIN within 4 hours, and once each 12 hours thereafter. 3.17.6.2 With ne channel of Rod Posi on Indication CRDMs: a) that the associated d group is within the ication 3.10 within 15 inutes after movement
3.17.6.3 With one or two SIRWT Te 3.17.6.4 one Main up .
Provide alternate mean of temperature monit ring within Fee~ater Flow a) rovide alternate means of low monitoring withi 3.17.6.5 With
a) Prov"de alternate means of tem rature monitoring wit 24 hours.
3.17.6.6. e or more flow path a) Deten11in the OPERABILITY of the a ociated AFW flow contr l valve within 2 3.17.6.6. a) The associat control valve shall inne ately be declared inoperable and he requirements of 3.5.2. apply. Amendment No. 3 1 67 1 96, 98 1 11§ 1 118, 1211 12~. 129, 136, 162 October 26, 1994 3-74
. ? -;-. c; INSTRUMENTATION SYSTEMS 3.17 Tab 1e 3 .17. 6 Instrumentation Ooeratjng Regyirements for Other Safety Functions Minimum Required OPERABLE Applicable Mon5 Instrument Channels Channels Conditions Q'3. i] 1. Neutron Flux Monitoring 2
[4pp/i~
2. 1 21al 1
*.\
2/valve1* (a) The,?{visions of/Specificat~s 3.0.4 .atld 4.0.4 aral'not applio-lble. -(~1~ p..lro 1.LI) (b) channels one channel ach of 7a~ 7c, d. (c) be any o channel of l.a('"1b, 7c, (continued) Amendment No. 3 1 &7 1 ~& 1 981 11&1 1181 121, 1241 129, 13&, 162 October 26, 1994
3-77
3.3. 'i 0i f
4.17 !HSTB\l!INJAIIQN sy5IQ!S TESTS Tab lt 4. 17.. 6
!nstrumtnUt1on SucytLJlani:t Rtauiremtnts foe Other Stfety Funct 1 ens CHANNEL CHANNEL FUNCTIONAL CHANNEL !nstryrntnt CHECK TEST CALIBRATION <S.e.~ 11l"5P ')
3,q 1* Neutron Flux Monitortn9 . lZ hours~~3.1.'l.1J lS months@l'l 3,3,'l.U
z. R~osttton I catton l~rs b) see 3,1
/ ~cil~' ,. "3.1$ / 3. SIRW Tank T HA 18 months
5. 18~

6. 18 months 7,

a. 12 hours b*
d. Atr Cooler ond1n11t1 Flow S*ttch
81. Prt*1r1 Sa 1 Valve acousttc 110nttor 18 lllOnthS 1qutr1d 18 months Sb/ Valvt I PORV-

91. tpt t__,1r1turt 9b. RV Acousttcal ttor 9c.
18 months
10. HA (1)
(b) (c) (conttnutd) MlndMnt Ho. Ui, t-64, 171 Apr11 5, 1996
4*81
3.3.ID 1
3. 17
3. 17. 3 3.17.3.1 With ont ls ation Function *anu control or Isolation F ct1on logic channel 1 trablt for one or
functions:
3.17.3.Z Wi one Isolation Fun on instru..nt channel 1 perablt for one or rt functions: a) Phct tht trii'un1t for Heh affected tripped c~tton wtth1n 7 days *
./
3.17.3.3 Wtth two Is,_r1tton Functton instM1111nt rnort funcyons: a) ~l{~, one channel trip unit reach affected Isol1t1on
.-the tr1pptd concUt1on wtth 8 hours, and / ,b-) Restore ont ch1nntl to 3.17.3.4 With ont or two EntinHrtd S1f19u1rds Rooe R1dhtton Monitors 1noperab 1t: AJA AC:rto A) AJo e.+
1) 3.17.3.5 {;co J.o.iJ 1nd Allltnd1111nt No. 16Z October Z6, 1994
3*68

3. l 7 [NSTRUMENTAT!ON SYSTEMS hbh 3.17.J Requir~ Minimu11 Permissible Isolation OPERABLE Oper1tion1l Cb1nn1Js Ch1nnels Bypassu
i. P logic Tr11ns z None.
Containment Pressure 4 2 Nont. S*itcbts
Left Train
c. Containment Prtssurt 4 2 S*itchts
Right Train
z. ~cc~1la1111ct ~lgb B1~11tlga (CH M1nu11 ln1t1atton l i .

b. CHR Logic Tr1tn1 z l

c. Cont1tnment Area 4 z R1di1tion 14onttor

3. (SGLP)
M1nu1l 1 set/tratn l set
b. SGLP 91c Tratns z 1 < SSO pstg Stua Prusurt.
/
c. Stea.m Generator 4 / z < 550 ps1g rusurt / Stua Prtssurt.
. *a* Steaa ;.n1r1tor 4 /// z < 550 ps1g Pressure Stua Pressure. *:. En::::*:::E::I Py* Boat** RW111v_*_: ______.%) . .H_:_:_._.- ~
Allltndmtnt No. 162 October Z6, 1994
3*69
l
4. 17 - INSTRUMENTATION SYSTEMS TESTS hb le 4. 17' 3 CHANNEL CHANNEL CHECK CAL!BBATIOH 1.
Tr11ns NA NA ont11ninent Pressure NA d i)'S 18 lllQnths Switches
Left Train Cont1in111ent Pressure 31 days 18 months Switches

Right Train
2. ~gct1Jcm~ct ~iab B1dl1tlga

i. M1nu11 In it ht 1on HA 18 months NA *.

b. CHR Lo9ie Trains HA 18 months

c. 12 hours Jl days NA NA

b. HA NA 12 hours 18 months
*e* Stu* Generator. 31 days 18 months Pressure
4. Ecgin1er1g S1fegy1cds PU!!!Q Boom High Bagj1tlga
~ozr 12 hours days 18 months
1. Jl

b. u Ro nitor 12 hours days 18 1110nths SR 3.3.10.JJ S/2-J.1.ro;~ 5 2. '3,3,10.~
~nd111nt No. ~. ~' 171 4~78
'7 ..... /'
(
I
----------,.--------------------------**---------------~
3. 16 I

3. 16 Engineered Safety Features instrumentation settin be as stated in T Appljcabjlitx
; Specification 3.16 is a licable when associated ESF or I Function 1nstrumentat1 is required to be OPERABLE by \ Specification J.17.2 r 3.17.3. :\
i Atli.Q.o 'i I3.l6.l i If an ESF instrumen setting Is not within the allowa e settings of Table 3.16, irrmedi ely declare the instrument inop able and complete
.*/*I i corrective actio ~s directed by specification 3. I \ _________L.-----------~ / / TABLE 3.16
I. Low Pressure psi a © ~
2. Pressure 3.70
4.40 psig
3. ontainment High Radiation s 20 R/h Steam Gener1tor Low Pressure ~ 500 psia

s. Steam Gener1tor Low Level/ ~ 25.9%
/
Narrow Range
/
6. SIRW Tank Low Level 21
27 inches Above Tank Bottom fJR 3.3.io.3] 7. Engineered Safegu1rds. Pump Room Ventilation High Radiation s 2.2 x 10 1 cpm r@
Amendment No. SQ, 162 October 26, 1994
. 3*63
3.3 RElocA Ttb
3.8 REFUELING OPEBATIONS Appl i cabil j ty Applies
\
nimize the possibility occurring dur1n refueling\ o rations that could affe and safety.
\
I 3.8.1 Specifications The following condi ons shall bt sat1sf1ed durin any refueling operations: One door of the emer ncy air lock shall be proper Whenever both doors o ht ptrsonntl air lock are o n during t refueling operations, th equiP1119nt door shall be ope and the entilating systeM and cha oal filter in tht fuel stor e building s 11 be operating.
b. valves shall be operable closed *
c.
d. ""lladhtion lev~ in the coftn..t.nMnt an spent fue s
' s~ be monito~ontinuousl~
e. g changed, neutron f be cont uously monitored by least two sourct r e neutron tors, with each rnon or providing conttnu s visual indication n the control roOftl. htn cart gtOflletry ts ot being changed, least one sourct r gt neutron monitor sh be in service.

f. At ltast one in operation *
3*46 P~ Io~ 9 Amendment No. 34 January 27. 19 ;e
2. ! i( lUC. Art:b 3.8 REFUELING OPERATIONS (Continued)
During r~~vessel head removal ancfW-; e re ue ing operations are being perfo ed in the reactor, the r fueling boron. '"~ concentration sh l be maintained in the p
ary coolant syst and
( all be checked by ampling on each shift. I h' Dire co1T111unication bet en personnel in the co rol room and at the re eling machine shal e available whenever c nges in core geometry re takin lace. 3.8.2 are not met, all ref. eling operation to atisfy the requi ed activity of the 3 .8.3. e u~ling operation shall not bt initiated before the reactor core deca: ed for a minimum of 4 hours if the reactor has been operat'itd power evels in excess of 2 rated power. . \ 3.8.4 The venti ation system and cha 011 filter in the fuel\storage buil shall be op rating whenever irra iated fuel which has cte_cayed less th 30 days is b ing handled by eithe of the following oper* tions:
a. eration with the eq pment door open, or
\
Fuel handling in the fuel storage ~ld1ng
both fans are unav ilable, any fuel llOVENts in progress sha be co leted and further el move111ents over th spent fuel storage ol sha be terminated unt1 *one fan is returned o service.
3.8.5. When spe"t fuel which has d ayed less than one ear is placed in the tilt pit* torage racks, the ~ lk water temperatur in the tilt pit storage ar must be monitored ontinuously to ass e that the water temperature oes not exceed 150
Monitoring will ntinue for 24 hours after a addition of fuel the main pool or t tilt pit or when a failure the spent fuel .PO cooling system occ s.
The equipment and g eral procedures t~ be utilized during efueling are dis~ssed in the FSA Detailed instruC.tions, the above spe ifications, and I\and sa ety features prov andling the sign of the fuel equipment ncorporating built* n interlocks e assurance that incident could oc during the refueli~perations that uld result in ah ard to public healt and safety.(1 Whenever changes rt not being made n core geometry, on flux ~
\.monitor is ufficient. This . !., 't
3.47 Amendment No. 3~. 31 May 22, 1984
?ClSe 2 o{ q
3. l 7 !.~STRUMENTATrOff SYSTEMS 3.3 K[Loc.A TED Socc1f1qt1on
3. 17. 6 A
3.17.5.l l) &nd b) c) Vtr1fy. HUTOOtllM l'ARGIN w1th1n 4 ho and once each 12 nours thtrtafttr. 3.17.S.Z Rod Posttton Ind atton 1noptrablt r one or ITIOrt
l) Vt fy that the ass tatld rod group Spec tcat1on 3.10 wt tn 15 *tnut11 af that gr p, w1th1n tht J.17.S.3 With one or tlfO IRVT r..,.rature chlftn1l1 1noptr 11:
l) 1n 7 days. 3.17.5.4 l) 3.17.S.5 W1 l) 3.17.S.S. cator for on1 or llOre f1ow path nop1r&bl1:
~
a) ILITY of tht assoc1attd AFW f w control v&lve 3.17.S.S.
1) Tht a tattd control valve shall 1 tely b* dtcl&rtd lnop1rablt and th1 requtrtmefttl of l.S.2.1 apply.
I AmlndMnt No. 31 '11 911 911 1111 1111 tali 1341 £2\), 13&, 162 Octotltr 26, 1994 3*74
J.17 . [NST~UMENTATtON sysTEMS
~ (continued) 3.17.6.7. Luk 01Uctton ct'l&n 11 (7&, tl, c, or d) &)
3.17.6.7.2 Wit two or tt'lrtt rtqutrtd 1ak Ottectton channels c, or d) i p1rabl1: a) R1stor1 tt'lr11 ct'lann1ls to OP£RAILE status wtth n 30 days. 3.17.6.S W1th one Primary Safety Valvt Posttton Indicator channtl Inoperable, fo ont or mart valvts: a) 1stort tt'lt channtls to °'ERAILE status prior to tn f m COLO SHUT~. 3.17.5.9 W1th onto two PORY Position Indt tor channels 1noptrablt or one or ITIOr"t valvu: a) Restort th chann1l1 to OPERAlll s tu1 prtor to th1 ntxt s fro* COLO S TOO!ilt
3.17.5.10 Wtth ont PORV Block alvt Postt10ft Indicator hlftfttl tnop1rabl1, ont or 1110rt valves:
a) R1stor1 th* channel o OP£RA1t.E status prior o tht n1xt startup fro* COLO SHUTOQWM, a , b)
f the PORV path t 1 rtqu for l TOP or as a PC ~*nt, and tht lvt posttton ltghts tno lilt, verify POAY Bloc Valve Is open u lZ hours.
3.17.6.11 With tht
&)
ERAILE status prtor to th* next channel inop1rabl1:
st1rtuo. ~. a) AmtndMnt No. J 1 '17 1 H1 91 1 1111 1111 1311 lH1 1391 lli, 162 October 26, 1994
1.17 tNSTRUMENTATtOH SYSTEMS
~ (continutd) 3.17.5.14 Wi~th* Cone Bor~c1d Tank Lo~evtl Al&n11 inop rable: "\_
a) erify the leve 1n tht affect Bortc Acid Tan is within l~s e h 12 hour a) Calcula the QUADRANT POWE least nee each 12 hours. or two AXIAL SHAP. INDEX Alar11 cha els inoperable: a) R1stort the sy~ to OP£RAILE st fro* COLO SHUTI)OWN. a)
3.17.S.19 Wtt
&)
b) Tht rt tor shall be laced tn a ca 'tttan wtltrt tht s not nqut , wtthtn 41 urs. AllltndMftt No. a, 111 911 911 1111 1111 Ul1 124 I 129; 13i. 162 October 26. 1994
3.17 !NSTBtJM£NTATION SYSTEMS Tab 11 l. 17.'
!nstrym1nt1t1on Og1c1t1nq B1gu1r1tntnts foe OtbtC Stftty fynct!gns M1n1-B1quirtd 0PERA8LE Aopltcablt InUrym1nt Cbann1l1 Cbanntls Cgndtt1gna 1 .~ron Fl u~i tor1 n9 Btl...10*~ RATED ?OlffR. / wi.tt(futl in tb~actor. ~od Pos1~ ~ Wtltn ~than on~DM 1s 3.
4. nd1cat1on
"' ~ capablt o~ w!tbd~al.
s 5.
7. PCS L11k19e Ottact1oa c.
SUllP Ltvtl l
d. At c Coo Condtnsata 1 Fl ow Sw1 tc:h l

a. Above lOO'F r_.

9. 3/valvt119 Above UO'F T block valvt open or tts pos 1on 1nd1cat1on syst.. ts inop1rabl1.
{a)~* provi'h.ns of~f1cat~0.4 ~4 ~*11~ ( COfttt nued) Amtndlltnt No. a.,..-17 1 H 1 91 1 111; 1111 131, 134, 1391 136, 162 October 26, 1994
3.77
3,3 R£LOCA 7£JJ
l.17 [NSTBUMENTATION sysTEMS Table 3.17.5 (cont1nutd)
In1trumtnt1ttqn Op1c1ttnq B1gy1rtmtnts foe Otbtr S1f1ty fynctlons Mt n f 111Uii BtQUif'td OPERABLE Apol fcablt ti.Q Instrymtnt Channtlt Ch1nn1!1 Cond1t1ons
10. 2/valv7.ltl l/Valv At all t1mts nltss tht PCS 1s 1pr1ssurized a vented 1 1ccord1nc1 111 t Sptc1f at1on 3.1.8.
i 1* 0
~z1o\ f~lux-~r Co~r ~~
13. Bod ~ Stqutnc.,....-- Z /
~ 0
15.~xc:ort
15. AX Alarit Ott~r tYht f Oft Ala SHAPE [NOD 11~'suct1on Y~ o AM99zoo
./""' Inttrloc~ . _ _,;/" ps1a / ...,...........PCS. Pf'usurt/'
18.
~ ~TS~a' 19.
l 20.
/
(b) 1c1'1c~3.~*~0.4~ a~lt.~ Amtndlllnt No> I; 171 91; Hu 111; 1111 lil I 136~t~~:~ u~* 1~~!
3*71 ..
P~e 7 o~ q
33 R[LDCA T D
,,17 INSTBVHEHT!TIOH SySTQ!S TESTS Table 4.17.. S Instrymtntat1on Sury*Lll1nc1 B1ou1rtrntnts for Othtr S*ftJy functj om CHANNEL CHANNEL FUNCTIONAL ln$trym1nt CHECK TEST 12 ho~
3. hs 18 ~n~
Ind1cat1on lZ hours 18 moo s PtS L11kag1 01t1c\1on: ,,
a. . SUllC)
,, Ltvt l lZ hour' 4V b* At~. Gas ii.on1tor Hua1d1t Mon1 tor 19 months Sa. NA af1ty Valve I V19 tallptpt t*mc> ature 9b. PORY Acou 9c. PORY S
10. PO Block Valvt P stt1on /
(continued)
.-..ndlllnt Ho. ~. 1-64, 171 Apri 1 S, l 996 4*11
4.17 !HSTB\J!'!ENT!TIOH SYSTQ!S TESTS Tibl* 4.17.S (continued)
!nstCIJ!Mnhtlon Su5';'1Uane1 R19u1remtnts for i!th1r Lf*tJ F'uni:t LOns CHANNEi.
FUNCTIONAL. UST
11. 18 rnonths eAI I "3.~.\-ZG:> ,___1z_*-~-------~-------_,&.------1

13. 18 months Not B1quir1d Exeort Otv1at1on 11 l!Wlnths lS. !SI Ala,..
17. 11 *
19. ll days zo. ll days
~nt No. ~ ...... 171 Aprn s, 1996
ATTACHMENT 3
AD1\11NISTRATIVE CHANGES (A)
DISCUSSION OF CHANGES SPECIFICATION 3.3.1, RPS INSTRUMENTATION A.l 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 CTS 3 .. 17.1 Applicability is when there is fuel in the reactor, more than one CONTROL ROD is capable of being withdrawn, and the PCS is less than REFUELING BORON CONCENTRATION. In the ITS, the proposed Applicability is MODES 1 and 2, and MODES 3, 4 and 5 with more than one control rod capable of being withdrawn, and the PCS boron concentration is less than REFUELING BORON CONCENTRATION. This Applicability is equivalent to the current Applicability, and the CTS has been revised to reflect the ITS wording. in MODES 1 and 2, more than one rod must be capable of being withdrawn, and therefore the phrase "more than one full length control rod is capable of being withdrawn" *is only associated with MODES 3, 4 and 5. The term "Control Rod" has been revised to read "full length control rod," in footnote (a) of ITS Table 3.3.1-1 (and in the Bases, where appropriate). This change in wording, from "control rod" to (revised) "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. The part length control rods have no clutches, remain fully withdrawn during operation, and are unaffected by RPS functions. The applicability limitation of "when there is fuel in the reactor" is retained in the ITS by the use of MODES in describing the applicable conditions. The ITS (and STS) definition of MODE ihcludes the limitation "with fuel in the reactor vessel." This change only involves a difference in presentation, and is considered administrative .

Palisades Nuclear Plant Page 1of11 05/30/99
ATTACHMENT 3
A.3 DISCUSSION OF CHANGES SPECIFICATION 3.3.1, RPS INSTRUMENTATION A Note was added to the Actions of CTS 3 .17 .1 which allows separate Condition entry for each RPS Function. The Note in ITS 3.3.1 provides explicit instructions for proper application of the Actions for Technical Specification compliance. In conjunction with the proposed Specification 1.3 - "Completion Times, " this Note provides direction consistent with the intent of the existing Actions for the RPS instrumentation. This change is consistent with NUREG-1432 {TSTF-178).
A.4 CTS 3 .17 .1. 6b requires the reactor to be placed in a condition where the affected equipment is not required .. In the ITS, this requirement is presented by specifying the actUal. actions necessary to satisfy this requirement. ITS 3.3.1 Required Actions 1.4.1 and 1.2.2 require that no more than one control rod be capable of being withdrawn or that the PCS boron concentration be of the REFUELING BORON CONCENTRATION. These actions place the plant in a condition where the affected equipment is not required. Since this change only provides more specific actions to be taken, it is considered administrative.
A.5 CTS Table 3.17.1 contains a "Minimum Operable Channels" column. This column is deleted in the ITS because the Actions in the ITS are based on the number of channels
*inoperable, from the total number of channels, which is specified in LCO 3.3.1. The
total number of channels. required to be Operable is unchanged from CTS. ITS conditions do not depend on a specified minimum operable channels for entry into .
actions. The crs use the specified minimum number of channels to initiate entry into. a shutdown action statement. In ITS, when the number of inoperable channels in a trip Fundtion exceeds that specified* in any related Condition, the plant does not meet the LCO and no associated Action is provided, therefore, a shutdown is required in accordance with LCO 3.0.3. is required. The discussion of the change in Required Actions is discussed in DOC M.2. This change only involves a change in presentation format, and is considered administrative. This change is consistent with the presentation format in NUREG-1432. A.6 . CTS Table 3.17.1 Footnote (a) requires two OPERABLE wide range nuclear instrument channels if the Zero Power Mode (ZPM) bypass is used. This Footnote is associated with the automatic ZPM bypass removal Function. ITS 3.3.1 does not include this Footnote, instead, the automatic ZPM bypass removal function has been specified as Function 12 in Table 3.3.1-1. ITS Condition D*provides guidance when bypass channel(s) are inoperable, allowing indefinite continued operation with an inoperable ZPM bypass removal channel only if the ZPM bypass is not used. These two ITS features are the equivalent of the CTS footnote (a) requirement that these bypass removal channels be operable when the bypass is used. Since the ITS requirements are equivalent to the CTS requirements, this change is considered to be
administrative .
Palisades Nuclear Plant Page 2of11 05/30/99
ATTACHMENT 3
A.7 DISCUSSION OF CHANGES SPECIFICATION 3.3.1, RPS INSTRUMENTATION CTS Table 3 .17 .1 Footnote (b) provides a restriction that the low power bypass (i.e., Zero Power Mode Bypass) cannot be enabled unless the SHUTDOWN BORON CONCENTRATION for the COLD SHUTDOWN condition is achieved. This provision is not included in the associated ITS Table 3.3.1-1 Footnote (b) since the SHUTDOWN MARGIN (SDM) requirements are met through ITS Section 3.1. The CTS definition of SHUTDOWN BORON CONCENTRATION required is > 2 % t:.p .
with all control rods inserted in the core and the highest worth control rod fully withdrawn. In ITS Specification 3.1.1, the minimum SDM required during any cooldown is specified in the COLR, and the COLR requirements must be calculated in accordance with approved methodology. Changes to the COLR must be performed in accordance with the requirements of ITS 5.6.5. The COLR SDM requirements provide the same assurance that the reactor will remain shutdown as do those in CTS. Therefore, since appropriate SDM must still be maintained throughout plant operation and cooldown, the removal of this provision can be considered administrative. This change essentially moves requirements from one Technical Specification to another, and therefore, is considered administrative. This change is consistent with NUREG-1432. A.8 CTS Table 4.17.1 Footnotes (b) and (c) require, a calibration of the Variable High Power trip Function channels (excore nuclear power and ~T power) with a heat balance, and calibration of the excore power channels with a test signal. In ITS, a requirement to adjust the instrumentation if the absolute difference between the* instrument readings and the results of the heat balance exceeds 1. 5 %. That requirement corresponds to "the necessary range and accuracy" requirement of the Channel Calibration definition, and does not constitute a change in requirements. Also in ITS, Table 3.3.1-1, these surveillances (SR 3.3.1.3 and SR 3.3.1.6) are also. associated with the Thermal Margin/Low Pressure (TM/LP) Trip Function. In CTS, this association is provided by a table in the bases. This association is necessary since the power signal from the excore power range is an input to the TM/LP trip function setpoint calculator. There is no additional testing required. This change is considered administrative, and is consistent with NUREG-1432. A.9 CTS 2.3 specifies the requirements for the RPS Limiting Safety System Settings. The Applicability of this Specification is when the associated RPS channels are required to be OPERABLE by Specification 3.17.1. In addition, CTS 2.3.1 specifies that if the RPS instrument setting is not within the allowable settings of Table 2.3.1, the instrument must be declared inoperable and complete corrective action as directed by Specification 3 .17 .1. In the ITS the RPS Allowable Values are listed in ITS Table 3. 3 .1-1 and the ITS 3. 3 .1 Actions provide identical protection. This change essentially moves requirements from one Technical Specification to another, and
. therefore, is considered administrative. This change is consistent with NUREG-1432 .
Palisades Nuclear Plant Page 3of11 05/30/99
ATTACHMENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.3.1, RPS INSTRUMENTATION A.10 CTS Table 2.3.1 provides the RPS Trip Setting Limits with four Primary Coolant I Pumps (PCPs) operating and with three PCPs operating. ITS Table 3.3.1-1 provides I the required Allowable Values for four PCP conditions, the three PCP settings have I been omitted. I I With the RPS Operable (including the trip setpoints being adjusted as required by ITS), I but with less than four PCPs operating, the Low Primary Coolant System Flow Trip I will be actuated, ensuring all full length control rods are inserted. Therefore removal I of the "Three Primary Coolant Pumps Operating" trip settings will not result in an I unacceptable operating situation. In addition, ITS LCO 3.4.4, "PCS Loops - I MODES 1 and 2," requires the plant to be placed in MODE 3 if less than four PCPs I are operating (an action the RPS performs automatically). I I The CTS Primary Coolant System Section requires four primary coolant pumps to be in I service, but contains an allowances to operate for up to 12 hours with only three (of the I four installed) primary coolant pumps in service. *The CTS Safety Limits Section I contains RPS setting requirements for both the three and four pump conditions. This I allowance for three pump operation is a holdover from the original Palisades Technical I
Specifications which allowed continuous operation, at specified RPS settings, with two,

three, or four pump operation. The deletion of the remaining 12 hour allowance for three pump operation is addressed in Section 3.4 of the ITS submittal as change 3.4.4 M. l. Both ITS LCOs 3.4.1 and 3.4.4 require four primary coolant pumps to be in operation, rendering the three pump RPS settings unnecessary and unusable. Since I
I I I I
three pump operation will no longer be allowed (due to the more restrictive change in I Section 3.4) deletion of these alternative settings, which can no longer be used, is I considered to be an administrative* change. I I*
A.11 CTS 4.18.2. lb requires a surveillance to compare the indivipual excore channel measured AO (ASI) to the total core AO measured by the incores, and associated calibration of the excore channel if the difference exceeds 0.02. ITS includes this requirement as SR 3.3.1.4 and applies it to the Thermal Margin/Low Power (TM/LP) I Function since these excore channels provide input to this RPS trip. This change is I considered administrative since it provides only a different format for identifying the requirements. A.12 CTS Table 4.17.1, item 16, requires a surveillance to verify control room temperature is acceptable for operability of the Thermal Margin Monitor (TMM). The TMM is*not specifically identified in the CTS LCO, ~or in the ITS. However, the TMM provides an input to both the VHPT and the TM/LP Functions. Therefore, this CTS requirement is reflected in ITS SR 3.3.1.2 which is identified in ITS Table 3.3.1-1 as a required SR for both the VHPT and the TM/LP Functions. This change is considered administrative since it provides only a different format for identifying the requirements. Palisades Nuclear Plant Page 4of11 05/30/99
TECHNICAL CHANGES - MORE RESTRICTIVE (M)
ATTACHMENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.3.1, RPS INSTRUMENTATION M.1 The allowances for channels to be. bypassed and the requirements for automatic removal of these bypasses is treated differently in CTS and ITS. Both treatments allow some
. trip Functions to be bypassed during conditions when the LCO requires them to be Operable. Both treatments require the bypasses on the safety related trips (those assumed to function by the safety analyses) to be automatically removed.
Operability: In CTS, the automatic bypass removal channels (for safety related channel bypasses)
are required to be Operable by footnote (a) of Table 3.17.1. (When the Wide Range Nuclear Instrument channels are Operable, if the indicated power level increases above the setpoint, the permissive signal is removed, automatically removing the bypass, regardless of the position of the manual bypass switch. This bypass is called the Zero Power Mode or "ZPM" bypass.) In ITS, the bypass removal channels are required to be Operable as part of the LCO statement and are specifically listed in Table 3. 3 .1-1 as

required Functions. The ITS LCO and Actions are worded to differentiate between "trip Functions" and "bypass removal Functions." These treatments are equivalent, with the ITS treatment being more explicit.
Applicability: In CTS, the automatic bypass removal channels are required to be Operable "When there is fuel in the reactor, more than one CONTROL ROD is capable of being withdrawn, and PCS boron concentration is less than REFUELING BORON CONCENTRATION, if the Zero Power Mode bypass is used"; in ITS, they are required to be Operable whenever the associated trips are required to be Operable, i..e., in Modes 1 and 2, and in Modes 3, 4, and 5 with more than one full-length control rod capable of being withdrawn and PCS boron concentration less than REFUELING BORON CONCENTRATION. The ITS, therefore, require the bypass removal channels to be Operable over a broader range of plant conditions than do the CTS, making this change More Restrictive . Palisades Nuclear Plant Page 5of11 05/30/99
ATTACHMENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.3.1, RPS INSTRUMENTATION M. l (continued) Bypass Allowance: In CTS, these bypass allowances are provided by footnote (b) of Table 3 .17 .1 which apply to the TM/LP trip, Low PCS Flow trip, and Low SG Pressure trip functions .
.That footnote allows these trips to be manually bypassed if the associated wide range nuclear instrument channels indicate below the specified power level. In ITS, the allowance to bypass the associated trip channels is provided in Table 3. 3 .1-1, footnote (c). These treatments are equivalent, with the ITS treatment being more explicit.
Automatic Bypass Removal:* While not explicitly stated, CTS requires automatic removal of the ZPM bypasses by requiring the associated instrument channels to be Operable: footnote (a) of Table 3.17.1 requires both wide range channels to be Operable if the subject bypass is used. Since the wide range channels, which supply the input for the High Startup Rate
trip, are subject to requirements for Channel Calibration (SR 4.17 .1, item 3), which explicitly requires verification of the bypass removal function. In ITS, the requirement for automatic removal of the ZPM bypass is provided as part of Table 3. 3 .1-1 footnote (c). These treatments are equivalent, with the ITS treatment being more explicit.
Actions: In CTS, the Required Actions for the bypass removal (i.e., wide range nuclear *1 instrument) channels are those provided for inoperable RPS trip, units. If a channel I were bypassed (other than as allowed by the footnotes), it could not perform its I specified function and would be inoperable. Therefore, if a trip channel were ZPM I bypassed, when a bypass removal (wide range NI) channel became inoperable, either I the bypass would have to be removed (which places the plant in a condition outside the I applicability), or the bypassed trip channel(s) would have to be declared inoperable. In I ITS, those actions have been preserved as ITS actions D.1 ~d D.2. These treatments I are equivalent, with the ITS treatment being more explicit. I Palisades Nuclear Plant Page 6of11 05/30/99
M.2 ATTACHMENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.3.1, RPS INSTRUMENTATION CTS 3.17.1.6 requires specific actions be taken when the number of OPERABLE channels is less than specified in the "Minimum OPERABLE Channels" column of Table 3 .17 .1. The actions are to place the reactor in HOT SHUTDOWN within 12 hours; and place the reactor in a condition where the affected equipment is not required, within 48 hours. In the ITS, the actions when the "number of OPERABLE Channels" is less than the CTS minimum required is to enter LCO 3.0.3. The actions of ITS LCO 3.0.3 are to initiate action within 1 hour to place the plant, as applicable, in MODE 3 within 7 hours; in MODE 4 within 31 hours; and in MODE 5 within 37 hours. Since this change will require the plant to be in a lower MODE in a shorter time frame, this change is considered more restrictive. This change continues to assure that a plant shutdown can be achieved in a controlled manner without challenging plant systems. This change is an additional restriction on plant operations and is consistent with NUREG-1432.
M.3 CTS 3.17. l.6b requires the reactor to be in a condition where the affected equipment is not required within 48 hours. ITS 3.3.1, Condition G Actions require the same condition be achieved within 6 hours. Since the plant is required to be in a lower MODE in a shorter time frame, this change is considered more restrictive. With the plant required to be in MODE 3 in 6 hours, de-energizing the clutch power supplies or borating to the REFUELING BORON CONCENTRATION can also be performed within the same time frame. These actions will not challenge plant systems, since the reactor is subcritical in MODE 3. This change is consistent with NUREG-1432. M.4 Not Used M.5 CTS Table 3.17.1 Footnote (c) provides an allowance to change the setpoint of the low power bypass setpoint from 104 % RTP to 10-1 % RTP during LOW POWER PHYSICS TESTING. Since this allowance is not used, it is not included in the ITS. Since this change deletes an allowance and the plant will no longer be able to change this low power setpoint during the ITS PHYSICS TEST (see Discussion of Changes for ITS Section 1.0), this change is considered more restrictive. This change is consistent with NUREG-1432. . .., . M.6 New Conditions Band C and new Required Actions B.1 and C. l have been added for the Loss of Load and High. Startup Rate Functions. These actions require restoration of the inoperable channel(s) prior to re-entering operational conditions during which the
Functions are required to be Operable. CTS Actions applicable to single channel inoperability for other RPS trip Functions are not required for High Startup Rate and Loss of Load. In ITS, some Action must be specified to avoid an LCO 3.0.3 entry upon channel inoperability. These added Actions are unique to Palisades, since, to the best of our knowledge, Palisades is the only CE plant with non-safety grade High Startup Rate and Loss of Load trips which are not credited in the safety analyses.
Palisades Nuclear Plant Page 7of11 05/30/99
M.6 (continued)
ATTACHMENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.3.1, RPS INSTRUMENTATION CTS allow continued operation for an unlimited period of time with one High Startup Rate of Loss of Load Trip inoperable. Therefore, LCO 3.0.4 would not require their restoration prior to re-entry of their applicable conditions, even after an extended shutdown. This open ended allowance is unnecessary and was not intended when the current CTS Actions were written. CTS contain other Actions for inoperable channels of supplementary equipment which allow continued operation for an unlimited period of time, but require restoration prior to the next startup. (See CTS Actions 3 .17. 6. 8, 9, 10, 11, & etc.) The ITS Actions provide this same limitation. The proposed Required Actions and Completion Times are considered to be acceptable because:
1. They are consistent with current practice, and do not impose unacceptable operational restriction,

2. They are more restrictive than the Actions required by CTS,

3. With one channel inoperable, the trip Function is still capable of initiating a reactor trip with a two-out-of-three logic, and

4. These trip Functions are not credited in the plant safety analysis.
This is a More Restrictive change because the ITS require restoration of an inoperable trip channel where the CTS do not. M. 7 Not used . Palisades Nuclear Plant Page 8of11 05/30/99
ATTACHMENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.3.1, RPS INSTRUMENTATION LESS RESTRICTIVE CHANGES - REMOVAL OF DETAILS TO LICENSEE CONTROLLED DOCUMENTS (LA)
LA.1 CTS SR 4.18.2. lb requires that "Individual excore channel measured AO shall be compared to the total core AO measured by the incores. If the difference is greater than 0.02, the excore monitoring system shall be recalibrated." This SR wording has been replaced with the STS wording "Calibrate the power range excore channels using the incore detectors." The adjustment details contained in CTS 4.18.2. lb have been moved to the procedure. These details merely provide the "the necessary range and accuracy to known values of the parameter that the channel monitors" referred to in the Channel Functional Test definition, and are not typically specified in ITS or STS SRs. These details are not necessary in the SR and have been relocated to the Bases. In the ITS, the pertinent requirements are inclu_ded in SR 3.3.1.4. The details related to how the power range detectors are calibrated are more appropriately included in the Bases. Changes to the Bases will be made in accordance with the Bases Control Program as discussed in ITS Chapter 5.0, Administrative Controls. This change maintains consistency with NUREG-1432 . LESS RESTRICTIVE CHANGES (L) L.1 CTS 3.17.1.3 requires entry when two RPS trip units or associated instrument channels are inoperable in one or more Functions. In the ITS, these conditions are addressed in ITS 3. 3 .1, Condition B. The difference is that the Condition B Required Actions include a Note which excludes the applicability of LCO 3.0.4. This provision was added to allow MODE changes even though two channels are inoperable, with one channel in trip. In this condition, with one channel in trip the protection system is in a one-out-of-two logic. Since the probability of a common cause failure affecting both of the OPERABLE channels during the 7 day Completion Time to restore one trip unit to OPERABLE status is remote, and the low probability of occurrence of an event during this interval, this change is considered acceptable. This change is consistent with NUREG-1432 . Palisades Nuclear Plant Page 9of11 05/30/99
L.2 ATTACHMENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.3.1, RPS INSTRUMENTATION The Frequency of the CHANNEL FUNCTIONAL TEST associated with certain RPS Functions CTS Tables 4.17.1is31 days. In ITS SR 3.3.1.5, the proposed Frequency is 92 days. The proposed change revises the CHANNEL FUNCTIONAL TEST Frequency for certain RPS Functions from 31 days to 92 days. Justification for extending the test interval is provided in Combustion Engineering (CE) Topical Report CEN-327-A Supplement 1, "RPS/ESFAS Extended Test Interval Evaluation," dated January 1989. NRC acceptance of the CE topical report was provided in a letter to the CE Owner's Group (CEOG) dated November 6, 1989. As discussed in the NRC acceptance letter, Consumers Energy has reviewed the instrument drift information for each instrument channel involved and has determined that drift occurring in that channel over the extended surveillance interval will not cause the setpoint to exceed it's Allowable Value as calculated by our setpoint methodology. Onsite records of the recorded instrument drift are available for audit. Therefore, it is proposed that the CHANNEL FUNCTIONAL TEST be performed in accordance with ITS SR 3.3.1.5, at a Frequency of 92 days. This change is consistent with NUREG-1432 .

L.3 CTS Table 4.17.1, Footnote (c), requires that the excore channels be calibrated with a test signal every 31 days. ITS SR 3.3.1.6 requires the surveillance to be performed every 92 days. The proposed change revises the calibration Frequency for the excore power range channels from 31 days to 92 days. This t~st leaves the channel adjusted to account for instrument drift between successive calibrations to ensure that the channel remains OPERABLE between calibrations. Other surveillances are performed on the excore power range channels more frequently to account for overall gain and to* ensure th.e upper and lower subchannel amplifiers are calibrated correctly to correspond to the incore detectors. These surveillances are considered sufficient to ensure the excore power range channels are functioning properly. This change is consistent with NUREG-1432.
L.4 CTS Table 4.17.1 Footnote (b) requires calibration of the Variable High Power Function with heat balance when power is > 15 % RTP. ITS 3. 3 .1 will add a Note to SR 3.3.1.3 (heat balance) which states that the SR is not required to be performed until 12 hours after THERMAL POWER is > 15% RTP. The allowance to delay performance of the SR for 12 hours after power is
> 15 % RTP provides time for the plant to achieve stable operating conditions to calibrate the instruments at a Rower at which the heat balance is accurate. This will provide more accurate results, and thereby providing assurance that the RPS Functions will actuate at the required setpoints. The 12 hours interval is within the 24 hour delay allowed for a missed SR. This change is consistent with NUREG-1432.
Palisades Nuclear Plant Page 10of11 05/30/99
ATTACHMENT 3
L.5 DISCUSSION OF CHANGES SPECIFICATION 3.3.1, RPS INSTRUMENTATION CTS 4.18.2. lb requires calibration of the excore monitoring system with the incore monitoring system at least every 31 days of power operation. ITS 3.3.1.4 requires the same surveillance, but is modified by a Note which states that. the SR is not required to be performed until 12 hours after '.THERMAL POWER is ~ 25% RTP. Since the CTS "power operations" is defined as "greater than 2 %, " this change is considered less restrictive.
This SR need only be performed above 25 % RTP since the calibration performed adjusts the Axial Shape Index (indicated by the excore power range) to agree with Axial Offset (indicated by the incore monitoring system). The ITS Axial Shape Index LCO, 3.2.4, has an Applicability of "MODE 1 with THERMAL POWER > 25 %". Below 25 % RTP, Axial Shape Index (ASI) is not limited by Technical Specifications because there is sufficient thermal margin to allow operation with potential axial power imbalances. Since the ASI LCO is not applicable below 25 % RTP, the excore power range channels need not be re-adjusted to precisely indicate ASL Selecting 25 % RTP as the starting point for the 12 hour window provided in SR 3.3.1.4 is consistent with other ITS requirements which utilize the incore monitoring system. The allowance to delay the surveillance for 12 hours after power is ~ 25 % RTP
provides time for the plant to stabilize at a power level sufficiently high enough to accurately calibrate the instruments. The 12 hour interval is acceptable because the need to calibrate the excores with the incores is dependant on potential changes in the core due to power generation. After outages, other than refueling outages, while the 31 day SR interval may have expired, the former calibration would still be valid; following a refueling outage, post-refueling power assention physics testing, completed prior to exceeding 25 % RTP, would uncover any unacceptable disparity between AO (measured by the incores) and ASI (measured by the excores). The 12 hours interval is within the 24 hour delay .allowed for a missed SR. This change is consistent with NUREG-1432 .

Palisades Nuclear Plant Page 11of11 05/30/99

ADMINISTRATIVE CHANGES (A)
ATTACHMENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.3.2, RPS LOGIC AND TRIP INITIATION 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 CTS 3.17.1 Applicability is when there is fuel in the reactor, more than one
CONTROL ROD is capable of being withdrawn, and the PCS is less than REFUELING BORON CONCENTRATION. In the ITS, the proposed Applicability is MODES 1 and 2, and MODES 3, 4 and 5 with more than one control rod capable of being withdrawn, and the PCS boron concentration is less than REFUELING BORON CONCENTRATION. This Applicability is equivalent to the current Applicability, and the CTS has been revised to reflect the ITS wording. In MODES 1 and 2, more than one rod must be capable of being withdrawn, and therefore the phrase "more than one full length control rod is capable of being withdrawn" is only associated with MODES 3, 4 and 5. The term "Control Rod" has been revised to read "full length control rod," in footnote (a) of ITS Table 3.3.1-1 (and in the Bases, where appropriate). This change in wording, from "control rod" to (revised) "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. The part length control rods have no clutches, remain fully withdrawn during operation, and are unaffected by RPS functions. The applicability limitation of "when there is fuel in the reactor" is retained in the ITS by the use of MODES in describing the applicable conditions. The ITS (and STS) definition of MODE includes the limitation "with fuel in the reactor

vessel." This change only.involves a difference
. in presentation, and is considered administrative.
Palisades Nuclear Plant Page 1of4 05/30/99

A.3 SPECIFICATION 3.3.2, RPS LOGIC AND TRIP INITIATION ATTACHMENT 3 DISCUSSION OF CHANGES CTS 3 .17 .1. 6b requires the reactor to be placed in a condition where the affected equipment is not required. In the ITS, this requirement is presented by specifying the actual actions necessary to satisfy this requirement. ITS 3.3.2 Required Actions E.1 and E.2 require that no more than one control rod be capable of being withdrawn or that the PCS boron concentration be at the REFUELING BORON CONCENTRATION. These actions place the plant in a condition where the affected equipment is not required. Since this change only provides more specific actions to be taken, it is considered administrative.
A.4 CTS Table 3.17.1 contains a "Minimum Channels Op~rable" column. This column is deleted in the ITS because the Actions in the ITS are based on the number of channels inoperable, from the total number of channels, which is specified in LCO 3.3.2. When the number of inoperable channels in a trip Function exceeds that specified in any related Condition associated with the same trip Function, the plant is outside the safety analysis, and therefore, entry into ITS LCO 3.0.3 is requifed. This change only involves a change in presentation format, and is considered administrative. This. change is consistent. with the presentation format in NUREG-1432 .. A.5 CTS 4.17 requires that the surveillance testing of the logic and control channels listed in CTS Table 3 .17 .1 be performed as specified in CTS Table 4. 17. 1. This CTS 4.17 reference is not used in ITS 3.3.2 since the requirements have b~n placed in the LCO and associated SRs. CTS Tables 3.17.1and4.17.1 have been deleted with respect to the RPS Matrix Logic, RPS Initiation and Manual Trip Functions, and therefore this cross reference is not needed in the ITS. Since this change only involves a difference in presentation between the CTS and the ITS, it is considered administrative. This change is consistent with NUREG-1432 . Palisades Nuclear Plant Page 2 of 4 05/30/99
ATTACHMENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.3.2, RPS LOGIC AND TRIP INITIATION TECHNICAL CHANGES - MORE RESTRICTIVE (M)
M.l CTS 3.17.1.5 requires that specific actions be taken when one channel of RPS Trip Initiation Logic is inoperable. ITS 3.3.2 contains the same requirements, as well as Condition D, which requires additional actions be taken when two channels of Initiation Logic affecting the same trip leg are inoperable. The CTS is revised to add ITS Condition D, which requires de-energizing the clutch power supplies immediately under this condition. With both RPS Trip Initiation Logic channels affecting the same trip leg inoperable, the RPS function is lost. Required Action D. l provides assurance that the RPS function is not lost. Additionally, since this change allows an action which is alternative to a shutdown, i.e., deenergizing the clutch power supplies, operation may continue avoiding a potentially unnecessary plant shutdown transient. Since this action requires immediate response, it-is considered more restrictive. This change is an additional restriction on plant operations an*d is consistent with NUREG-1432. M.2 CTS 3.17.1.6 requires.specific Actions be taken when the "number of OPERABLE channels is less than specified in the "Minimum OPERABLE Channels" column of Table 3.17.1," and when "any action required by CTS 3.17.1 is not met AND the associated completion time has expired." These Actions are to place the reactor in HOT SHUTDOWN within 12 hours, and in a condition where the affected equipment is not required within 48 hours. Based on the current Applicability of CTS 3 .17 .1, the condition where the affected equipment is not required is when more than one CONTROL ROD is capable of being withdrawn and the PCS is less than REFUELING BORON CONCENTRATION. In the ITS, the Required Action, when the "Required Action and associated Completion Times are not met" is to enter ITS 3.3.2 Condition E, which requires that the plant exit the Applicability within 6 hours. Since this change will require the plant to be in a lower MODE in a shorter time frame, this change is considered more restrictive. This change continues to assure that a plant shutdown can be achieved in a controlled manner without challenging plant systems. This change is an additional restriction on plant operations and is consistent with NUREG-1432. LESS RESTRICTIVE CH~GES - REMOVAL OF DETAILS TO LICENSEE CONTROLLED DOCUMENTS (LA) There were no "Removal of Details" changes associated with this specification . Palisades Nuclear Plant Page 3 of 4 05/30/99
LESS RESTRICTIVE CHANGES (L)
ATTACHMENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.3.2, RPS LOGIC AND TRIP INITIATION L. l The Frequency of the CHANNEL FUNCTIONAL TEST associated with the RPS Logic (i.e., Matrix Logic and Trip Initiation Logic) in CTS 4. 17. 1 is 31 days. In ITS SR 3.3.2.1, the proposed Frequency is 92 days. The proposed change revises the surveillance Frequency of the RPS Matrix Logic and Initiation Logic CHANNEL FUNCTIONAL TEST from 31 days to 92 days. Justification for extending the test interval is provided in Combustion Engineering (CE) Topical Report CEN-327-A Supplement 1, "RPS/ESFAS Extended Test Interval Evaluation," dated January 1989. NRC acceptance of the CE topical report was provided in a letter to the CE Owner's Group (CEOG) dated November 6, 1989. As discussed in the NRC acceptance letter, Consumers Energy has reviewed the instrument drift information for each instrument channel involved and has determined that drift occurring in that channel over the extended surveillance interval will not cause the setpoint to exceed it's Allowable Value as calculated by our setpoint methodology. Onsite records of the recorded instrument drift are available for audit. Therefore the CHANNEL FUNCTIONAL TEST will performed in accordance with ITS SR 3.3.2.1ata92 day Frequency. This change is consistent with NUREG-1432 . Palisades Nuclear Plant Page 4 of 4 05/30/99
ADMINISTRATIVE CHANGES (A)
ATTACHMENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.3.3, ESF INSTRUMENTATION A.l 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 required number of channels associated with the ESF and Isolation Instrument Functions are specified in CTS Tables 3.17.2 and 3.17.3, respectively; ITS LCO 3.3.3
*specifies the requirement that four instrument channels are required to be OPERABLE.
The "Required Channels" column of CTS Table 3.17.2 and 3.17.3 is not included in ITS Table 3.3.3-1. Since the number of channels in the CTS and ITS are identical this change is editorial in nature and is therefore considered administrative. This change is consistent with NUREG-1432. A.3 CTS 3.17.2 requires the instrumentation channels to be OPERABLE except as allowed by the permissible operational bypasses. Reference to operational bypasses is deleted, because the CTS Table 3.17 .2, Function 1.d (Safety Injection Signal) references to operational bypasses are addressed in lTS LCO 3.3.4. In addition, the overall Applicability of this Specification is when PCS temperature is greater than or equal to 300°F as indicated in the Applicability of CTS 3.17.2. The Applicability of ITS 3.3.3 is "In accordance with Table 3.3.3-1." The ITS Table includes a MODES column where the Applicable conditions are included for each function_. The ITS Applicability for the Functions associated with CTS 3.17.2 is MODES 1, 2 and 3. In the ITS, MODE 3 is whenever PCS temperature is greater than or equal to 300°F as specified in ITS Table 1.1-1. Therefore the Applicability in the CTS and ITS are identical since operations in MODE 1 and 2 will also be with PCS temperature greater than or equal . to 300°F. This change in format to identify the Applicability of the Functions is equivalent to the current requirements therefore this change is considered administrative . Palisades Nuclear Plant Page 1of8 05/30/99
A.4 ATTACHMENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.3.3, ESF INSTRUMENTATION A Note was added to the Actions of CTS 3:17.2 and 3.17.3 which allows separate Condition entry for each ESFAS trip and applicable bypass removal function. The Note in ITS 3.3.3 provides explicit instructions for proper application of the Actions for Technical Specification compliance. In conjunction with the proposed Specification 1.3 - "Completion Times, " this Note provides direction consistent with the intent of the existing Actions for the ESP instrumentation. This change is consistent with NUREG-1432.
A.5 CTS 3.17.3 requires the instrumentation channels to be OPERABLE except as allowed by the permissible operations bypasses. Reference to operational bypasses is deleted,, because the CTS Table 3.17.3, Functions 3.c and 3.d (Steam Generator Low Pressure) references to operational bypasses are addressed in ITS LCO 3.3.4. In addition, the overall Applicability of this Specification is when the PCS is above COLD SHUTDOWN. The Applicability of ITS 3.3.3 is "In accordance with Table 3.3.3-1." ITS Table 3.3.3-1 includes a MODES column where the Applicable conditions are included for each Function. The ITS Applicability associated with CTS Table 3.17.3 Functions 1.b and 1.c (Containment High Pressure), 2.c (Containment High Radiation), and 3.c and 3.d (Steam Generator Low Pressure) is MODES 1, 2, 3 and 4 . The differences in the applicability between the CTS and ITS are negligible. For the CTS COLD SHUTDOWN versus the ITS MODE 5, the temperature requirement is being less than 210°F versus being less than 200°F in the ITS. This difference which is common between the CTS terms HOT SHUTDOWN and COLD SHUTDOWN and the corresponding ITS MODES 3 and 5 is the reactivity condition. The ITS MODE 3 and 5 are defined, as a reference point, by a reactivity condition of Keff < .99. However, in ITS S~ction 3 .1, the equivalent amount of SHUTDOWN MARGIN is required as that specified in the CTS definitions of HOT SHUTDOWN and COLD SHUTDOWN. Therefore, the amount of SHUTDOWN MARGIN is considered to be same when the requirements of proposed ITS 3.1 are considered. These changes with respect to the Applicability are all considered to be administrative in that there is no significant impact to operation of the plant and they reflect the terminology and usage rules of NUREG-1432. A.6 CTS 3.16 includes the requirement that the Engineered Safety Features (ESF) System instrumentation setting limits shall be as stated in Table 3.16. The CTS 3.16 Applicability is when the associated ESF or Isolation Function instrumentation is required to be OPERABLE by Specification 3.17.2 or 3.17.3. CTS 3.16.1 requires that the instrumentation be declared inoperable when the settings are not within the allowable values of Table 3.16 and complete the corrective action as directed by Specification 3.17. ITS 3.3.3 includes all of these requirements in the associated LCO and ACTIONS. Therefore, this cross reference to CTS 3 .17 is not necessary and is deleted. Since this change is simply a change in format it is considered administrative. Palisades Nuclear Plant Page 2 of 8 05/30/99
ATTACHMENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.3.3, ESF INSTRUMENTATION A.7 CTS Table 4.17 .2, Functional Unit 2.c, requires both a CHANNEL FUNCTIONAL TEST and a CHANNEL CALIBRATION _to be performed on the SIRWT Level Switches every 18 months. In the ITS, a CHANNEL CALIBRATION (SR 3.3.3.3) is only required because the definition of CHANNEL CALIBRATION explicitly encompasses the CHANNEL FUNCTIONAL TEST. Since the requirements for CHANNEL CALIBRATION include the CHANNEL FUNCTIONAL TEST, this change is considered administrative. MORE RESTRICTIVE CHANGES (M) M.1 CTS 3.17.2.5 contains Required Actions to be taken when the "number of OPERABLE channels is less than specified in the "Minimum OPERABLE Channels" column of Table 3.17.2. In CTS, these Required Actions are to place the reactor in HOT SHUTDOWN within 12 hours and place the reactor in a condition where the affected equipment is not required, within 48 hours. In ITS, these Required Actions have been replaced by LCO 3.0.3 .
Both CTS 3.17.2 and ITS 3.3.3 contain: Required Actions for conditions where the plant equipment does not meet the LCO, but still is able to provide the required safety function. In CTS, an explicit "shutdown" Required Action is also provided for equipment failures beyond this acceptable level. In ITS, since no condition is provided to address conditions beyond _the acceptable levelof degradation, the shutdown requirement is provided by LCO 3.0.3. Both of these Required Action treatments require the plant to be taken to a condition where the affected LCO is not applicable.
The times specified, to reach similar conditions, in ITS are shorter and therefore more restrictive. The CTS "Operating Condition" definitions differ from the ITS "Mode'; definitions of the same name. CTS Required Action 3.17.2.Sa requires the plant to be in Hot Shutdown (i.e., subcritical) within 12 hours; ITS LCO 3.0.3a requires the plant to be in MODE 3 (subcritical) within 7 hours. When starting from power operation, the entry conditions for CTS Hot Shutdown and ITS MODE 3 are essentially identical; when the plant is subcritical. Since the required action is the same, and the completion time is shortened, this change is considered to be More Restrictive . Palisades Nuclear Plant Page 3 of 8 05/30/99
M. l (continued)
ATTACHMENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.3.3, ESF INSTRUMENTATION CTS 3.17.2 is applicable when the plant is above 300°F. CTS Required Actions 3.17.2.Sb requires the plant to be "in a condition where the affected equipment is not required within 48 hours." For CTS LCO 3 .17. 2, that would be below 300°F. ITS LCO 3.0.3 requires "the plant shall be placed in a MODE or other specified condition in which the LCO is not applicable." ITS LCO allows 31 hours to be in MODE 4 (i.e., below 300°F). Again, since the required action is the same, and the completion time is shortened, this change is considered to be More Restrictive. Therefore the replacement of CTS 3.17.2.5 wording "or if the number of operable channels is less than specified in the "Minimum Operable Channels" and reliance upon the requirements of ITS LCO 3.0.3 will result in a More Restrictive Change. This change is consistent with NUREG-1432. M.2 Both CTS 3.17.2.5 and ITS 3.3.3 Condition D contain Required Actions to be taken when any Required Action is not completed and the associated completion time has expired. In CTS, these Required Actions are to place the reactor in HOT
SHUTDOWN within 12 hours and place the reactor in a condition where the affected equipment is not required, within 48 hours.1 In ITS LCO 3.3.3 these Required Actions are to be in MODE 3 within 6 hours, and to be in MODE 4 within 12 hours (for functions which are addressed by CTS LCO 3 .17 .2).
The CTS "Operating Condition" definitions differ from the ITS "MODE" definitions of the same name. CTS Required Actions 3 .17 .2 .5a requires the plant to be in Hot Shutdown
-(i.e., subcritical) within 12 hours; ITS Required Action 3. 3. 3 D .1 requires the plant to be in MODE 3 (subcritical) within 6 hours. When starting from power operation, the entry conditions for CTS Hot Shutdown and ITS MODE 3 are essentially identical; when the plant is subcritical. Since the required action is the same, and the completion time is shortened, this change is considered to be More Restrictive.
CTS 3.17.2 is applicable when the plant is above 300°F. CTS Required Actions 3.17.2.5b requires the plant to be placed "in a condition where the affected equipment is not required within 48 hours." For CTS LCO 3 .17. 2, that would be below 300°F. ITS Required Action 3.3.3 D.2 requires to be in MODE 4 (i.e., below 300°F) within 30 hours. Again, since the required action is the same, and the completion time is shortened, this change is considered to be More Restrictive.
\
Therefore the replacement of the CTS 3.17.2.5 requirements to be taken "when any Required Action is not completed and the associated completion time has expired" by ITS Required Actions 3.3.3 D.1 and D.2 will result in a More Restrictive Change. This change is consistent with NUREG-1432. Palisades Nuclear Plant Page 4 of 8 05/30/99
M.3 ATTACHMENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.3.3, ESF INSTRUMENTATION CTS 3.17.3.5 contains Required Actions to be taken when the "number of OPERABLE
~hannels is less than specified in the "Minimum OPERABLE Channels" column of Table 3.17.3. In CTS, these Required Actions are to place the reactor in HOT SHUTDOWN within 12 hours and place the reactor in a condition where the affected equipment is not required, within 48 hours. In ITS, these Required Actions have been replaced by LCO 3.0.3.
Both CTS 3 .17. 3 and ITS 3. 3. 3 contain Required Actions for conditions where the plant equipment does not meet the LCO, but still is able to provide the required safety function. In CTS, an explicit "shutdown" Required Action is also provided for equipment failures beyond this acceptable level. In IT_S, since no condition is provided to address conditions beyond the acceptable level of degradation, the shutdown requirement is provided by LCO 3.0.3. Both of these Required Action treatments require the plant to be taken to a condition where the affected LCO is not applicable. The times specified, to reach similar conditions, in ITS are shorter and therefore more restrictive.
The CTS "Operating Condition" definitions differ from the ITS "MODE" definitions of the same name.
CTS Required Action 3.17.3.Sa requires the plant to be in Hot Shutdown (i.e., subcritical) within 12 hours; ITS LCO 3.0.3a requires the plant to be in MODE 3 (subcritical) within 7 hours. When starting .from power operation, the entry conditions for CTS Hot Shutdown and ITS MODE 3 are essentially identical; when the plant is subcritical. Since the required action is the same, and the completion time is shortened, this change is considered to be More Restrictive. CTS 3.17.3 is applicable when the plant is above Cold Shutdown (i.e., above 210°F). CTS Required Actions 3.17.3.Sb requires the plant to be "in a condition where the affected equipment is not required within 48 hours." For CTS LCO 3.17.3, that would be below 210°F. ITS LCO 3.0.3 requires "the plant shall be placed in a MODE or other specified condition in which the LCO is not applicable." ITS LCO allows 37 hours to be in MODE 5 (i.e., below 200°F). Again, since the required action is essentially the same, and the completion time is shortened, this change is considered to be More Restrictive. Therefore the replacement of CTS 3.17.3.5 wording "or if the number of operable channels is less than specified in the "Minimum Operable Channels" and reliance upon the requirements of ITS LCO 3.0.3 will result in a More Restrictive Change. This change is consistent with NUREG-1432. Palisades Nuclear Plant Page 5 of 8 05/30/99
M.4 ATTACHMENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.3.3, ESF INSTRUMENTATION Both CTS 3.17.3.5 and ITS 3.3.3 Condition E contain Required Actions to be taken I when any Required Action is not completed and the associated completion time has I expired. In CTS, these Required Actions are to place the reactor in HOT I SHUTDOWN within 12 hours and place the reactor in a condition where the affected I equipment is not required, within 48 hours. In ITS LCO 3.3.3 these-Required Actions I are to be in MODE 3 within 6 hours, and to be in MODE 5 within 12 hours (for I functions which are addressed by CTS LCO 3.17.3). I I
The CTS "Operating Condition" definitions differ from the ITS "MODE" definitions .I of the same name. I I CTS Required Actions 3.17.3.Sa requires the plant to be in Hot Shutdown I (i.e., subcritical) within 12 hours; ITS Required Action 3.3.3 E.1 requires the plant I
'to be in MODE 3 (subcritical) within 6 hours. When starting from power operation, I the entry conditions for CTS Hot Shutdown and ITS MODE 3 are essentially identical; I when the plant is subcritical. Since the required action is the same, and the completion I.
time is shortened, this change is considered to be More Restrictive .
CTS 3.17.3 is applicable when the plant is above 210°F. CTS Required Actions 3.17.3.Sb requires the plant to be placed "in a condition where the affected equipment is not required within 48 hours." For CTS LCO 3 .17. 3, that would be below 210°F. ITS Required Action 3.3.3 E.2 requires to be in MODE 5 (i.e., below 200°F) within 36 hours. Again, since the required action is essentially the same, and the completion time is shortened, this change is considered to be More Restrictive.
Therefore the replacement of the CTS 3.17.3.5 requirements to be taken "when any Required Action is not completed and the associated completion time has expired" by ITS Required Actions 3.3.3 E.1 and E.2 will ie.sult in a More Restrictive Change. This change is consistent with NUREG-1432. M.5 The Allowable Value for the upper limit on CHP actuation has been reduced from 4.4 psig to 4.3 psig to assure that there is adequate margin for instrument tolerances. During review of the allowed values for the Containment High Pressure (CHP) Engineered Safety Feature (ESF) actuation setpoint for the Improved Technical Specification project, it was determined that the upper limit of 4.4 psig specified in current Technical Specification Table 3.16 Item 2 was not consistent with the assumptions of the FSAR Chapter 14.18 containment response analyses. The containment response analyses assume that the CHP ESF actuation occurs prior to containment pressure exceeding 4.3 psig, when allowance is made for the allowed "as found" calibration tolerances for the actual CHP ESF pressure switches. This change is more restrictive with respect to requiring a more rapid actuation following postulated LOCA or MSLB event. The 4.3 psig setting limit is that currently used by the plant, and is the setting required by the Operating Requirements Manual. Palisades Nuclear Plant Page 6 of 8 05130199 _j
M.5 (Continued)
ATTACHMENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.3.3, ESF INSTRUMENTATION If the CHP pressure switches were to actuate at the 4.4 psig setting as permitted by current Technical Specification Table 3.16 Item 2, CHP actuation could be delayed by up to 0.1 seconds beyond the time assumed. This delay could result in extending the peak calculated containment pressure by up to 0.1 psig. Therefore, it is necessary to place a more restrictive req4irement on the CHP ESF allowable value, reducing the upper limit from 4.4 to 4.3 psig. LESS RESTRICTIVE CHANGES - REMOVAL OF DETAILS TO LICENSEE CONTROLLED DOCUMENTS (LA) There were no "Less Restrictive" changes associated with this specification. ,/ LESS RESTRICTIVE CHANGES (L)
L.1 CTS Actions 3.17.2.3 and 3.17.3.3 are required to be entered when two ESF (except for SIRWT) or isolation instrumentation channels are inoperable in one or more Functions. In the ITS, ~hese conditions are addressed in ITS 3.3.3 Condition B. The difference is that the ITS Required Actions are modified by a Note which excludes the applicability of LCO 3.0.4. This provision was added to allow MODE changes even though two channels are inoperable, and with one channel in trip. In this condition, with one channel in trip the protection system is in a one-out-of-two logic. This change is considered acceptable based on the low probability of a common cause failure affecting both of the OPERABLE channels during the 7 day Completion Time, and the low probability of an event occurring during this time. This change is consistent with NUREG-1432.
L.2 CTS Tables 4.17.2 and 4.17.3 require performance of a CHANNEL FUNCTIONAL TEST at a Frequency of 31 days. ITS SR 3.3.3.2 requires performance of a CHANNEL FUNCTIONAL TEST at a proposed Frequency of 92 days. The proposed change revises the surveillance Frequency of the ESF instrumentation CHANNEL FUNCTIONAL TEST from 31 days to 92 days. Justification for extending the test interval is provided in Combustion
Engineering (CE) Topical Report CEN-327-A Supplement 1, "RPS/ESFAS Extended Test Interval Evaluation," dated January 1989. NRC acceptance of the CE topical report was provided in a letter to the CE Owner's Group (CEOG) dated November 6, 1989.
Palisades Nuclear Plant Page 7 of 8 05/30/99
L.2 (Continued)
ATTACHMENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.3.3, ESF INSTRUMENTATION As discussed in the NRC acceptance letter, Consumers Energy has reviewed the instrument drift information for each instrument channel involved and has determined that drift occurring in that channel over the extended surveillance interval will not cause the setpoint to exceed it's Allowable Value as calculated by our setpoint methodology. Onsite records of the recorded instrument drift are available for audit. Therefore, the CHANNEL FUNCTIONAL TEST will performed in accordance with ITS SR 3.3.3.2 at a Frequency of 92 days. This change is consistent with NUREG-1432.
L.3 CTS 3.17.3 requires the ESF Instrumentation Function of Steam Generator Low Pressure (SGLP) to be OPERABLE "when the PCS is above COLD SHUTDOWN." For ITS, this would be reflected as an Applicability of MODES 1, 2, 3, and 4. However, the purpose of the SGLP Function is to
. I provide automatic isolation of the main steam and main feedwater lines following a Main Steam Line Break (MSLB) or Main Feedwater Line Break (MFLB). CTS 3.5.1.f requires that the Main Steam Isolation Valves (MSIVs) be OPERABLE when PCS is above 300°F, i.e., ITS MODES 1, 2, and :t Also, CTS does not include any specific requirements for OPERABILITY of valves which isolate main feedwater, i.e., the MFIVs. ITS 3.7.2 requires the MSIVs to be OPERABLE during MODE 1, and during MODES 2 and 3 except when all MSIVs are closed and de-activated. ITS 3.7.3 requires the MFIVs to be OPERABLE during MODES 1, 2 and 3 except when all MFIVs are closed and de-activated, or are isolated by closed manu£11 valves. Since the automatic
isolation function will have no purpose if the valve is not required to be OPERABLE, the CTS is revised such that the ITS 3.3.3 Applicability for the SGLP Function is MODES 1, 2, and 3 with MODES 2 and 3 modified by a Note. The Note will indicate that the SGLP is not required when all MSIVs are closed and de-activated, and all MFIVs are closed and de-activated, or are isolated by closed manual valves. This Note makes the automatic isolation function Applicability consistent with the Applicability of the components actuated by the instrumentation Function .
Palisades Nuclear Plant Page 8 of 8 05/30/99
I
ATTACHMENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.3.4, ESF LOGIC AND MANUAL INITIATION 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 CTS 3.17.2 requires the instrumentation channels to be OPERABLE except as allowed by the permissible operational bypasses. Reference to operational bypasses is revised such that they are specifically addressed in the ITS. In addition, the overall
. Applicability of this Specification is when PCS temperature is greater than or equal to 300°F as indicated in the Applicability of CTS 3.17.2. The Applicability of ITS 3.3.4 is "According to Table 3.3.4-1." The ITS Table includes a MODES column where the Applicable conditions are included for each function. The ITS Applicability for the Functions associated with CTS 3.17.2 is MODES 1, 2 and 3. In the ITS, MODE 3 is whenever PCS temperature is greater than or equal to 300°F as specified in ITS Table 1.1-1. Therefore the Applicability in the CTS and ITS are identical since operations in MODE 1 and 2 will also be with PCS temperature greater than or equal to 300°F. This change in format to identify the Applicability of the Functions is equivalent to the current requirements and is considered administrative.
A.3 Not Used. A.4 A Note was added to the Actions of CTS 3.17.2 and 3.17.3 which allows separate Condition entry for each Function. The Note in ITS 3.3.4 provides explicit instructions for proper application of the Actions for Technical Specification compliance. In conjunction with the proposed Specification 1.3 - "Completion Times," this Note provides direction consistent with the intent of the existing Actions for the ESF instrumentation. This change is consistent with NUREG-1432 . Palisades Nuclear Plant Page 1of8 05/30/99
A.5 SPECIFICATION 3.3.4, ESF LOGIC AND MANUAL INITIATION ATTACHMENT 3 DISCUSSION OF CHANGES CTS 3 .17. 3 requires the instrumentation channels to be OPERABLE except as allowed by the permissible operations bypasses. Reference to operational bypasses is revised such that they are specifically addressed in the ITS. In addition, the overall Applicability of this Specification is when the PCS is above COLD SHUTDOWN.
The Applicability of ITS 3. 3. 4 is "According to Table 3. 3 .4-1." ITS Table 3. 3 .4-1 includes a MODES column where the Applicable conditions are included for each Function. The ITS Applicability associated with.CTS Table 3.17.3 is MODES 1, 2, 3, and 4. The differences in the applicability between the CTS and ITS are negligible. For the CTS COLD SHUTDOWN versus the ITS MODE 5, the temperature requirement is being less than 210°F versus being less than 200°F in the ITS. This difference which is common between the CTS terms HOT SHUTDOWN and COLD SHUTDOWN and the corresponding ITS MODES 3 arid 5 is the. reactivity condition. The ITS MODE 3 and 5 are defined, as a reference point, by a reactivity condition of Keff < .99. However, in ITS Section 3.1, the equivalent amount of SHUTDOWN MARGIN is required as that specified in the CTS definitions of HOT SHUTDOWN and COLD SHUTDOWN . Therefore, the amount of SHUTDOWN MARGIN is considered to be same when the requirements of proposed ITS 3.1 are considered. These changes with respect to the Applicability are all considered to be administrative in that there is no significant impact to operation of the plant and they reflect the terminology and usage rules of NUREG-1432. A.6 The requirements of CTS Tables 3.17.2 and 3.17.3 are revised to incorporate Footnote (c), which states that manual initiation may be achieved by individual component controls, and is applicable to Manual Initiation for Function 4.a (SGLP). This footnote is equivalent to the CTS entry requiring "1 set [of controls]/train, and is added to clarify that use of any individual component controls to actuate the SGLP Function is adequate. This change is considered to be administrative, in that it only provides clarification and does not alter any technical requirements. A.7 CTS Table 4.17 .2 has been revised by adding a Note to the 92 day CHANNEL FUNCTIONAL TEST which states that testing of the Actuation Logic shall include verification of the proper operation of each initiation relay. This change is considered to be administrative, in that it only provides clarification and does not alter any technical requirements . Palisades Nuclear Plant Page 2 of 8 05/30/99
ATTACHMENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.3.4, ESF LOGIC AND MANUAL INITIATION MORE RESTRICTIVE CHANGES (M)
M.l CTS 3.17.2.5 contains Required Actions to be taken when the "number of OPERABLE channels is less than specified in the "Minimum OPERABLE Channels" column of Table 3.17.2. In CTS, these Required Actions are to place the reactor in HOT SHUTDOWN within 12 hours and place the reactor in a condition where the affected equipment is nof required, within 48 hours. In ITS, these Required Actions have been replaced by LCO 3.0.3. Both CTS 3.17.2 and ITS 3.3.4 contain Required Actions for conditions where the plant equipment does not meet the LCO, but still is able to provide the required safety function. In CTS, an explicit "shutdown" Required Action is also provided for equipment failures beyond this acceptable level. In ITS, since no condition is provided to address conditions beyond the acceptable level of degradation, the shutdown requirement is provided by LCO 3.0.3. Both of these Required Action treatments require the plant to be taken to a condition where the affected LCO is not applicable. The times specified, to reach similar conditions, in ITS are shorter and therefore more restrictive . The CTS "Operating Condition" definitions differ from the ITS "MODE" definitions 1 of the same name. CTS Required Action 3.17 .2.5a requires the plant to be in Hot Shutdown (i.e., subcritical) within 12 hours; ITS LCO 3.0.3a requires the plant to be in MODE 3 (subcritical) within 7 hours. When starting from power operation, the entry conditions for CTS Hot Shutdown and ITS MODE 3 are essentially identical; when the plant is subcritical. Since the required action is the same, and the completion tinie is shortened, this change is considered to be More Restrictive. CTS 3.17.2 is applicable when the plant is above 300°F. CTS Required Actions 3.17.2.Sb requires the plant to be "in a condition where the affected equipment is not required within 48 hours." For CTS LCO 3.17.2, that would be below 300°F. ITS LCO 3.0.3 requires "the plant shall be placed in a MODE or other specified
condition in which the LCO is not applicable." ITS LCO allows 31 hours to be in MODE 4 (i.e., below 300°F). Again, since the required action is the same, and the completion time is shortened, this change is considered to be More Restrictive.
Therefore the replacement of CTS 3.17.2.5 wording "or if the number of operable channels is less than specified in the "Minimum Operable Channels" and reliance upon the requirements of ITS LCO 3.0.3 will result in a More Restrictive Change. This change is consistent with NUREG-1432. Palisades Nuclear Plant Page 3 of 8 05/30/99
M.2 SPECIFICATION 3.3.4, ESF LOGIC AND MANUAL INITIATION ATTACHMENT 3 DISCUSSION OF CHANGES Both CTS 3.17.2.5 and ITS 3.3.4 Condition B contain Required Actions to be taken when any Required Action is not completed and the associated completion time has expired. In CTS, these Required Actions are to place the reactor in HOT SHUTDOWN within 12 hours and place the reactor in a condition where the affected equipment is not required, within 48 hours. In ITS LCO 3.3.4 these Required Actions are to be in MODE 3 within 6 hours, and to be in MODE 4 within 12 hours (for functions which are addressed by CTS LCO 3.17.2).
The CTS "Operating Condition" definitions differ from the ITS "MODE" definitions of the same name. CTS Required Actions 3.17.2.5a requires the plant to be in Hot Shutdown (i.e., subcritical) within 12 hours; ITS Required Action 3.3.4 B.1 requires the plant to be in MODE 3 (subcritical) within 6 hours. When starting from power operation, the entry conditions for CTS Hot Shutdown and ITS MODE 3 are essentially identical; when the plant is subcritical. Since the required action is the same, and* the completion time is shortened, this change is considered to be More Restrictive .
CTS 3.17.2 is applicable when the plant is above 300°F. CTS Required Actions 3.17.2.5b requires the plant to be placed "in a condition where the affected equipment is not required within 48 hours." For CTS LCO 3 .17. 2, that would be below 300°F. ITS Required Action 3.3.4 B.2 requires to be in MODE 4 (i.e., below 300°F) within 30 hours. Again, since the required action is the same, and the completion*time is shortened, this change is considered to be More Restrictive.
Therefore the replacement of the CTS 3.17 .2.5 requirements to be taken "when any Required Action is not completed and the associated completion time has expired" by ITS Required Actions 3.3.4 B.1 and B.2 will result in a More Restrictive Change. This change is consistent with NUREG-1432 . Palisades Nuclear Plant Page 4 of 8 05/30/99
M.3 SPECIFICATION 3.3.4, ESF LOGIC AND MANUAL INITIATION ATTACHMENT 3 DISCUSSION OF CHANGES CTS 3.17 .3.5 contains Required Actions to be taken when the "number of OPERABLE I channels is less than specified in the "Minimum OPERABLE Channels" column of I Table 3.17.3. In CTS, these Required Actions are to place the reactor in HOT I SHUTDOWN within 12 hours and place the reactor in a condition where the affected I equipment is not required, within 48 hours. In ITS, these Required Actions have been I replaced by LCO 3.0.3. I I
Both CTS 3.17.3 and ITS 3.3.4 contain Required Actions for conditions where the I plant equipment does not meet the LCO, but still is able to provide the required safety I function. In CTS, an explicit "shutdown" Required Action is also provided for I equipment failures beyond this acceptable level. In ITS, since no condition is provided I to address conditions beyond the acceptable level of degradation, the shutdown I requirement is provided by LCO 3.0.3. Both of these Required Action treatments I require the plant to be taken to a condition where the affected LCO is not applicable. I The times specified, to reach similar conditions, in ITS are shorter and therefore more. I restrictive. I I The CTS "Operating Condition" definitions differ from the ITS "MODE" definitions I of the same name. I I CTS Required Action 3.17.3.5a requires the plant to be in Hot Shutdown I (i.e., subcritical) within 12 hours; ITS LCO 3.0.3a requires the plant to be in MODE 3 I (subcritical) within 7 hours. When starting from power operation, -the entry conditions I for CTS Hot Shutdown and ITS MODE 3 are essentially identical; when the plant is I subcritical. Since the required action is the same, and the completion time is I shortened, this change is considered to be More Restrictive. I I CTS 3.17.3 is applicable when the plant is above Cold Shutdown (i.e., above 210°F). I CTS Required Actions 3.17.3.Sb requires the plant to be "in a condition where the I affected equipment is not required within 48 hours." For CTS LCO 3 .17. 3, that would I be below 210°F. ITS LCO 3.0.3 requires "the plant shall be placed in a MODE or I other specified condition in which the LCO is not applicable." ITS LCO allows I 37 hours to be in MODE 5 (i.e., below 200°F). Again, since the required action is I. essentially the same, and the completion time is shortened, this change is considered to I be More Restrictive. I
.I Therefore the replacement of CTS 3.17.3.5 wording "or if the number of operable I channels is less than specified in the "Minimum Operable Channels" and reliance upon I the requirements of ITS LCO 3.0.3 will result in a More Restrictive Change. This I
change is consistent with NUREG-1432. I I
I Palisades Nuclear Plant Page 5 of 8 05/30/99
M.4 SPECIFICATION 3.3.4, ESF LOGIC AND MANUAL INITIATION ATTACHMENT 3 DISCUSSION OF CHANGES Both CTS 3.17.3.5 and ITS 3.3.4 Condition C contain Required Actions to be taken I when any Required Action is 'not completed and the associated completion time has I expired. In CTS, these Required Actions are to place the reactor in HOT I SHUTDOWN within 12 hours and place the reactor in a condition where the affected I equipment is not required, within 48 hours. In ITS LCO 3.3.4 these Required Actions *I are to be in MODE 3 within 6 hours, and to be in MODE 5 within 12 hours (for I functions which are addressed by CTS LCO 3.17.3). I I
The CTS "Operating Condition" definitions differ from the ITS "MODE" definitions I of the same name. I I CTS Required Actions 3.17.3.5a requires the plant to be in Hot Shutdown I (i.e., subcritical) within 12 hours; ITS Required Action 3.3.4 C. l requires the plant to I be in MODE 3 (subcritical) within 6 hours. When starting from power operation, the **I entry conditions for CTS Hot Shutdown and ITS MODE 3 are essentially identical; *I when the plant is subcritical. Since the. required action is the same, and the completion I time is shortened, this change is considered to be More Restrictive. I
CTS 3.17.3 is appiicable when the plant is above 210°F. CTS Required Actions 3.17.3.5b requires the plant to be placed "in a condition where the affected equipment is not required within48 hours." For CTS LCO 3.17.3, that would be below 210°F. ITS Required Action 3.3.4 C.2 requires to be in MODE 5 (i.e., below 200°F) within 36 hours. Again, since the required action is ess.entially the same~ and I
the completion time is shortened, this change is considered to be More Restrictive. Therefore the replacement of the CTS 3.17.3.5 requirements to be taken "when any Required Action is not completed and the associated completion time has expired" by ITS Required Actions 3.3.4 C.1 and C.2 will result in a More Restridive Change. This change is consistent with NUREG-1432 . Palisades Nuclear Plant Page 6 of 8 05/30/99
ATTACHMENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.3.4, ESF LOGIC AND MANUAL INITIATION LESS RESTRICTIVE CHAN~ES - REMOVAL OF DETAILS TO LICENSEE CONTROLLED DOCUMENTS (LA)
LA.1 CTS Tables 3. 17. 2 and 4. 17. 2 Function 1. b and 1. c contain details of the scope
*of logic circuits associated with these Functions. Specifically, the CTS Function 1.b lists "Initiation, Actuation, and low pressure block auto reset,"
while Function 1.c lists "5P Relay Outputs." In the ITS, proposed ITS Table 3. 3 .4-1 Function 1. b requires that the SIS Actuation Logic (Pressurizer Pressure - Low) be Operable while proposed Function 1.c requires that the SIS Actuation Logic (CHP) be Operable. The details of what constitutes an Operable SIS. Actuation
. Logic (Pressure - Low) and SIS Actuation Logic (CHP) are specified in the Bases. The Bases state that "an Actuation Logic Channel consi~ts of all circuitry housed within the actuation logic circuits, including the initiating relays contacts responsible for actuating the ESF equipment." For the .
SIS Actuation logic (Pressurizer Pressure - Low) the automatic bypass removal feature is also required. Removing the. detail.s of the ESF from the CTS and placing them in the Bases of the ITS is acceptable since these details are not pertinent to the actual requirements. . 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 . Palisades Nuclear Plant Page 7 of 8 05/30/99
ATTACHMENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.3.4, ESF LOGIC AND MANUAL INITIATION LESS RESTRICTIVE CHANGES (L)
L.1 CTS 3.17.3 requires the ESF Instrumentation Function of Steam Generator Low Pressure (SGLP) to be OPERABLE "when the PCS is above COLD SHUTDOWN." For ITS, this would be reflected as an Applicability of MODES 1, 2, 3, and 4. However, the purpose of the SGLP Function is to
-provide automatic isolation of the main steam and main feedwater lines following a Main Steam Line Break (MSLB) or Main Feedwater Line Break (MFLB). CTS 3.5.1.f requires that the Main Steam Isolation Valves (MSIVs) be OPERABLE when PCS is above 300°F, i.e., ITS MODES 1, 2, and 3.
Also, CTS does not include any specific requirements for OPERABILITY of valves which isolate main feedwater, i.e., the MFIVs. ITS 3.7.2 requires the MSIVs to be OPERABLE during MODE 1, and during MODES 2 and 3 except when all MSIVs are closed and de-activated. ITS 3.7.3 requires the MFIVs to be OPERABLE during MODES 1, 2 and 3 except when all MFIVs are closed and de-activated, or are isolated by closed manual valves. Since the automatic isolation function will have no purpose if the valve i,s not required to be OPERABLE, the CTS is revised such that the ITS 3.l.3 Applicability for the SGLP Function is MODES 1, 2, and 3 with MODES 2 and 3 modified by a Note. The Note will indicate that the SGLP is not required when all MSIVs are closed and de-activated, and all MFIVs are closed and de-activated, or are isolated by closed manual valves. This Note makes the automatic isolation function Applicability consistent with the Applicability of the components actuated by the instrumentation Function . Palisades Nuclear Plant Page 8 of 8 05/30/99
ADMINISTRATIVE CHANGES (A)
ATTACHMENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.3.5, DG-UNDERVOLTAGE START There were no "Administrative" changes associated with this specification. MORE RESTRICTIVE CHANGES (M) M. l A new specification is proposed to establish consistency with NUREG-1432. Proposed ITS 3.3.5, "Diesel Generator (DG) - Undervoltage Start (UV Start)," contains the Limiting Conditions for Operation, Applicability, Actions, and Surveillance Requirements to ensure the plant is protected following various accidents in conjunction with a loss of voltage or degraded voltage condition on the plant Class lE emergency buses by ensuring OPERABILITY of the loss of voltage and degraded voltage automatic start features of the DGs. Therefore, this change is more restrictive. The DG - UV Start Function is required in MODES 1, 2, 3, and 4 because ESP Functions are designed to provide protection in these MODES. Actuation in MODE 5 or 6 is required whenever the required DG must be OPERABLE, so that it can perform its function on a loss of power or degraded power to the vital bus . LESS RESTRICTIVE CHANGES - REMOVAL OF DETAILS TO LICENSEE CONTROLLED DOCUMENTS (LA) There were no "Less Restrictive - Removal of Details to Licensee" changes associated with this specification. LESS RESTRICTIVE CHANGES (L) There were no "Less Restrictive" changes associated with this specification . Palisades Nuclear Plant Page 1of1 05/30/99
ATTACHMENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.3.6, REFUELING CHR INSTRUMENTATION 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 CTS 3.8 and 3.8.1 applies to operating limitations during refueling operations. CTS Table 3 .17 .6 Function 20 (Containment Refueling Radiation Monitor) is applicable during REFUELING OPERATIONS when irradiated fuel is in the Containment. The Applicability of ITS 3.3.6, "Refueling Containment High Radiation (CHR) Instrumentation," is stated as "during CORE ALTERATIONS" and "during movement of irradiated fuel assemblies within containment." The CTS defines "refueling operations" as "any operation involving movement of core components (except for incore detectors) when the reactor vessel head is untensioned or removed with fuel in the reactor vessel." In the ITS, CORE ALTERATIONS is defined, in part, as "the movement or manipulation of any fuel, sources, or reactivity control components within the reactor vessel with the head removed and fuel in the vessel." Although the wording of the CTS applicability and ITS Applicability are not exact, the intent of both the CTS and ITS is to address the condition where there is a potential for a fuel handling accident. As such, this change in wording is considered to be editorial in nature. A.3 CTS 3.8. lc requires two radiation monitors to be tested and verified to be operable immediately prior to refueling operations. In addition, Table 3.17.6 requires two Containment Refueling Radiation Monitors to be OPERABLE during REFUELING OPERATIONS when irradiated fuel is in the containment. In the ITS, the requirement that two radiation monitors be OPERABLE is placed in ITS LCO 3.3.6. The details of. the CTS Applicability and testing requirements have been relocated to the Applicability and Surveillance
Requirements of ITS 3.3.6. This change is simply a change in format placing the requirements of the CTS in the corresponding ITS LCO, and is therefore considered to be administrative.
Palisades Nuclear Plant Page 1of4 05/30/99
A.4 SPECIFICATION 3.3.6, REFUELING CHR INSTRUMENTATION ATTACHMENT 3 DISCUSSION OF CHANGES The Bases of the current Technical Specifications for this section have been completely replaced by revised Bases that reflect the format and applicable content consistent with NUREG-1432. The revised Bases are shown in the proposed Technical Specification Bases.
A.5 CTS Table 3.17.6, Function 20, Footnote (a), contains an allowance that Specifications 3.0.4 and 4.0.4 are not applicable. Footnote (b) contains an allowance that Specifications 3.0.3, 3.0.4, and 4.0.4 are not applicable. Since the applicability of Specifications 3.0.3, 3.0.4 and 4.0.4 is changed in the ITS to only apply in MODES 1, 2, 3 and 4, these footnotes have been deleted. Since this change does not alter any technical requirements, it is"considered to be administrative. MORE RESTRICTIVE CHANGES (M) M.1 CTS Table 4.17.6, Function 20, requires that a CHANNEL CHECK be performed on the containment refueling radiation monitor at a Frequency of 24 hours. ITS SR 3.3.6.1 requires that the CHANNEL CHECK be performed at a Frequency of 12 hours. This change is a reduction in surveillance Frequency and is considered more restrictive. The 12 hour Frequency is based on operating experience, and minimizes the chance of a loss of a protective function due to failure of redundant channels.* This change is consistent with NUREG-1432. M.2 ITS LCO 3.3.6 contains a requirement for two CHR manual actuation function channels to be operable in addition to t~e two refueling containment radiation monitor channels required by CTS LCOs 3.8 and 3.17. This requirement was added to assure that diverse actuation methods are available if open containment isolation penetrations need to be closed during Core Alterations or movement of irradiated fuel. The requirement for manual closure capability is consistent ~ith NUREG 1432 . Palisades Nuclear Plant Page 2 of 4 05/30/99
ATTACHMENT 3
DISCUSSION OF CHANGES SPECIFICATION 3.3.6, REFUELING CHR INSTRUMENTATION LESS RESTRICTIVE CHANGES - REMOVAL OF DETAILS TO LICENSEE CONTROLLED DOCUMENTS (LA)

LA.1 CTS 3.8.1 requires two radiation monitors that initiate containment vent and purge isolation to be OPERABLE. The requirement also describes the location of the monitors, the associated logic (one-out-of two logic), the monitors must be part of the plant area monitoring system, the equipment is not required during normal *plant operation and a switch shall be provided so that isolation action can be initiated during refueling only. These details are not necessary in the LCO and have been relocated to the Bases. In the ITS, the pertinent requirements are included in LCO 3.3.6. The details on the associated logic

and what the channels consist of are included in the Bases. Changes to the Bases will be made in accordance with the Bases Control Program as discussed in TS Chapter 5.0, Administrative Controls. This change maintains consistency with NUREG-1432. .
LESS RESTRICTIVE CHANGES (L)
L.1 CTS 3.8. lc requires the containment vent and purge system radiation monitors to be tested and verified OPERA_BLE "immediately prior to refueling operations." In addition, CTS Table 4.17.6 requires a CHANNEL CHECK every 24 hours, a CHANNEL FUNCTIONAL TEST every 31 days and a CHANNEL CALIBRATION every 18 months. Proposed ITS 3.3.6 and SR 3.0.4 require the OPERABILITY of each radiation monitor prior to entering the MODE of Applicability (e.g., prior to CORE ALTERATIONS, or the movement of irradiated fuel assemblies within the containment). The CTS.has been revised to delete the "immediately prior to refueling operation" requirement and only require the CTS Table 4.17.6 surveillances. This is acceptable since the ITS provides general rules for the application of SRs in th.e Technical Specifications. SR 3.0.4 establishes the requirement that the applicable SRs must be met before entry into a mode or other specified condition in the Applicability. In addition, the specific time frames and condition necessary for meeting the SRs are specified in the Frequency.
Although the phrase "immediately prior to refueling operations" implies a conditional type frequency, proposed SRs 3.3.6.1, 3.3.6.2, and 3.3.6.3 specify a specific Frequency. The proposed surveillances in these SRs are consistent with similar testing for ESF instrumentation which is considered to be acceptable. This change is consistent with NUREG-1432 .
Palisades Nuclear Plant Page 3 of 4 05/30/99
ATTACHMENT 3
L.2 DISCUSSION OF CHANGES SPECIFICATION 3.3.6, REFUELING CHR INSTRUMENTATION CTS 3.8.2 requires that if any of the.conditions of CTS 3.8.1 are not met, all refueling operations shall cease immediately. For ITS LCO 3.3.6, when one required channel of automatic (or manual) actuation is inoperable, that Required Action has been replaced with Actions which require placing the inoperable channel in trip (i.e., performing its safety function) or removing the plant from the applicable conditions of the LCO. The completion time has been extended from immediately to 4 hours. The extension of the completion time from "immediately" to 4 hours is offset by the additional requirement (see DOC M.2) to have two manual actuation channels Operable. The four hour completion time only applies when at least one channel of manual actuation and one channel of automatic actuation remain Operable.
With two channels inoperable, ITS 3.3.6 retains the CTS action of removing the plant from the applicable conditions of the LCO immediately. CTS 3.8.2 further requires that, "work shall be initiated to satisfy the required conditions and that no operations that may change the reactivity of the core shall be made." These requirements are not appropriate once the Applicability for the LCO has been exited. Therefore, they have been deleted .
The ITS retains the pertinent, and sufficient, requirements in LCO 3.3.6 to assure the OPERABILITY of the Refueling CHR instrumentation when it is needed to support the reduction of dose consequences following a fuel handling accident. Therefore, this deletion can be made with no impact to the health and safety of the public. This change is consistent with NUREG-1432 .

Palisades Nuclear Plant Page 4 of 4 05/30/99
ATTACHMENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.3.7, PAM INSTRUMENTATION ADl\flNISTRATIVE CHANGES (A) A.1 The proposed change will reformat, renumber, and reword the Current Technical Specifications (CTS), with no change of intent, to be consistent with NUREG-1432. As a result, the Improved Technical Specifications (ITS) should be more easily readable and, therefore, understandable by plant operators, as well as other users. During the Palisades Nuclear Plant ITS development, certain wording preferences or conventions were adopted which resulted in no technical changes to the CTS. Additional information may also have been added to more fully describe each LCO and to be consistent with NUREG-1432. However, the additional information does not change the intent of the CTS. The reformatting, ren~mb.ering, and rewording process involves no technical changes to CTS. A.2 CTS 3.17.4 is applicable when the PCS temperature is > 300°F. ITS 3.3.7 is applicable in MODES 1, 2, and 3. *In accordance with ITS.Table 1.1-1 average reactor coolant temperature must be > 300°F while in MODES 1, 2, and 3. As such, this change is administrative. This change is consistent with NUREG-1432.
A.3 A Note was added to CTS 3.17.4 Actions which allows separate Condition entry for
. each function. The Note in ITS 3.3. 7 provides explicit instructions for proper application of the actions for Technical Specification compliance. In conjunction with the proposed Specification 1.3 - "Completion Times," this Note provides direction consistent with the intent of the existing Action~ for the post accident m_onitoring instrumentation. This change is consistent with NUREG-1432.
MORE RESTRICTIVE CHANGES (M) M.1 CTS 3.17.4 provides an.exception to Specification 3.0.3. ITS 3.3.7 does not include an exception to ITS LCO 3.0.3. ITS LCO 3.0.3 delineates the time limits for placing the plant in a safe MODE or other specified condition when operation cannot be maintained within the limits for safe operation as defined by the LCO and its ACTIONS. The addition of this restriction, consistent with NUREG-1432, is an additionalJimitation on plant operation and, therefore, a more restrictive change. This change will not adversely affect plant safety .
Palisades Nuclear Plant Page 1of5 05/30/99
----------------~
A TTACH:MENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.3.7, PAM INSTRUMENTATION M.2 CTS 3 .17.4 provides an exception to Specification 4. 0 .4. ITS does not include an exception to ITS SR 3.0.4. This Specification ensures that system and component OPERABILITY requirements and variable limits are met before entry into MODES or other specified conditions in the Applicability for which these systems and components ensure safe operation of the plant. The addition of this restriction, consistent with NUREG-1432, is an additional limitation on plant operation and, therefore, a more restrictive change. This change will not adversely affect plant safety. M.3 CTS 3.17.4.4 requires the reactor be placed in Hot Shutdown within 12 hours, and the reactor be placed in a condition where the affected equipment is not required, within 48 hours, if any action required by 3.17.4.1through3.17.4.3 is not met and the associated completion time has expired. ITS 3.3. 7 Action F in conjunction with Action E requires the plant be MODE 3 within 6 hours and in MODE 4 within 30 hours if any accident monitoring Function except Reactor Vessel Water Level or High Range Containment Radiation Functions does not have at least one Operable channel after expiration of the Completion Times in ITS 3.3.7 Condition C or D. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems. This restriction, consistent with NUREG-1432, is an
additional limitation on plant operation and, therefore, a more restrictive change. This change will not adversely affect plant safety.
LESS RESTRICTIVE CHANGES - REMOVAL OF DETAILS TO LICENSEE CONTROLLED DOCUMENTS (LA) LA. I CTS Table 4 .17 .4 requires that the Core Exit Thermocouples (CETs) be calibrated by substituting a known voltage for the thermocouple. These details are not necessary in the LCO and have been relocated to the Bases. In the ITS, the pertinent requirements are included in SR 3. 3. 7. 2. The details related to how the CETs are calibrated are more appropriately included in the Bases. Changes to the Bases will be made in accordance with the Bases Control Program as discussed in TS Chapter 5 .0, Administrative Controls. This change maintains consistency with NUREG-1432 .
Palisades Nuclear Plant Page 2 of 5 05/30/99

LESS RESTRICTIVE CHANGES (L)
ATTACHMENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.3.7, PAM INSTRUMENTATION L.1 The ITS Actions contain completion times which are longer than their CTS counterparts; ITS Action A.1 allows 30 days for restoration of a single inoperable channel for one or more Functions; CTS 3.3.4 Actions la, 3a, and Sa allow 7 days for this restoration. The longer ITS completion times are taken from the STS. They are based on the additional operating experience available when the STS were written and take into account the passive nature of the instruments (no automatic action is assumed to occur from these instruments), and the low probability of an event requiring accident monitoring instrumentation during this interval. The 30 day Completion Time for restoration of a single inoperable channel (for one or more Functions) also considers the remaining Operable channel. This change is consistent with NUREG 1432. L.2 The ITS Actions contain completion times which are longer than their CTS counterparts; ITS Action D. l allows 3 days for restoration for one of two inoperable channels of Hydrogen Monitoring; CTS Action 2a allows 2 days for this restoration. The longer ITS completion times are taken from the STS. They are based on the
additional operating experience available when the STS were written and take into account the passive nature of the instruments (no automatic action is .assumed to occur from these instruments), and the low probability of an event requiring accident monitoring instrumentation during this interval. These changes are consistent with NUREG 1432.
L.3 The ITS Actions contain completion times which are longer than their CTS
.counterparts; ITS Action C.1 *allows 7 days for restoration for one of two inoperable channels (for Functions other than Hydrogen Monitoring); CTS Actions 2a and 6a I allow 2 days for this restoration. The longer ITS completion times are taken from the I .. STS. They are based on the additional operating experience available when the STS I were written and take into account the passive nature of the instruments (no automatic . I action is assumed to occur from these instruments), and the low probability of an event I requiring accident monitoring instrumentation during this interval. This change is I consistent with NUREG 1432. I I
Palisades Nuclear Plant Page 3 of 5 05/30/99 j
L.4 ATTACHMENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.3.7, PAM INSTRUMENTATION The ITS Action B.1 allows continued operation, provided a report is filed with the NRC, if a single inoperable channel for one or more functions cannot be restored to operable status within the specified completion time of Action A.1; CTS Actions 4a and 4b require a plant shutdown in this event (for Functions other than Reactor Vessel Water Level and Containment Area Radiation Monitoring).
The requirement to file a report with the NRC (initiate action in accordance with Specification 5.6.6), is appropriate given the likelihood of plant conditions that would require information provided by this instrumentation. Also, alternative Required Actions are identified before a loss of functional capability condition occurs. This change is consistent with NUREG-1432. L.5 CTS 3.17.4. 7b requirement that alternate monitoring be initiated within 48 hours with two Reactor Vessel Water Level channels inoperable is deleted. The instrumentation which is used as an alternative to RVWL channels is also required by CTS 3.17.4 and ITS 3.3.7. No new alternative measurement methods are initiated to implement the requirements of CTS 3.17.4. 7b, rather, the Emergency Operating Procedures list alternative indications to be used in lieu of operable RVWL channels, which provide similar, although not identical, information. The PCS temperature channels and Pressurizer Level instruments provide information to supplement any inoperable RVWL channels. ITS 5.6.6 requires the Accident Monitoring Report to outline the preplanned alternate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels of the Function to Operable . status. These controls assure that the appropriate margin of safety is maintained. This change is consistent with NUREG-1432. L.6 CTS 3.17.4.7d requirement that the inoperable channels for functions 16 through 21 be restored to Operable status prior to startup from the next refueling is deleted. However, the NRC will be informed of the schedule for restoring the system, as required by ITS 5.6.6. This requirement will ensure thatthe inoperable channel(s) are restored to Operable status in a timely manner. Also, good operating practice and management oversight dictate that plant systems be restored as soon as possible. This change is consistent with NUREG-1432. L. 7 CTS 3.17.4.3 provides action requirements for inoperable position indication of containment isolation valves which allows the associated valves to be locked in the closed position. ISTS Table 3.3.11-1 Note (a) also provides this allowance but also allows other options such as isolation of the associated penetration by any automatic valve which is closed and de-activated, a closed manual valve, a blind flange, or a
check valve with flow through the valve secured. Since all of these options provide acceptable isolation of the penetration, ITS 3.3. 7 will adopt the additional allowances as provided in the ISTS. This change is consistent with NUREG-1432.
Palisades Nuclear Plant Page 4 of 5 05/30/99
ATTACHMENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.3.7, PAM INSTRUMENTATION L.8 CTS Table 4.17.4 provides CHANNEL CHECK Surveillance Requirements for post accident monitors with a Frequency of 31 days. ISTS SR 3.3.11.1 also provides CHANNEL CHECK Surveillance Requirements for post accident monitors with a Frequency of 31 days, but requires the CHANNEL CHECK to be performed only if the required instrumentation channel is normally energized. CTS requires the check regardless of the normal energization state. Therefore, this change is less restrictive than CTS. The revised Frequency is acceptable based on industry experience that demonstrates channel failure is rare, and on the use of less formal but more frequent checks of the channels during normal operational use of displays associated with the - required channels. This change is consistent with NUREG-1432 .
Palisades Nuclear Plant Page 5 of 5 05/30/99
ATTACHMENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.3.8, ALTERNATE SHUTDOWN SYSTEM ADMINISTRATIVE CHANGES (A) A. l The proposed change will reformat, renumber, and reword the Current Technical Specifications (CTS), with no change of intent, 'to be consistent with NUREG-1432. As a result, the Improved Technical Specifications (ITS) should be more easily readable and, therefore, understandable by plant operators, as well as other users. During the Palisades Nuclear Plant ITS development, certain wording preferences or conventions were adopted which resulted in no technical changes to the CTS. Additional information may also have been added to more fully describe each LCO and to be consistent with NUREG-1432. However, the additional information does not change the intent of the CTS. The reformatting, renumbering, and rewording process involves no technical changes to CTS. A.2 CTS 3.17.5 is applicable when the PCS temperature is > 300°F. ITS 3.3.8 is applicable in MODES 1, 2, and 3. In accordance with ITS Table 1.1-1 average reactor coolant temperature must be > 300°F while in MODES 1, 2, and 3. As such, this change is administrative. This change is consistent with NUREG-1432. A.3 A Note was added to CTS 3.17.5 Actions which allows separate Condition entry for each function. The Note in ITS 3.3.8 provides explicit instructions for proper application of the Actions for Technical Specification compliance. In conjunction with the proposed Specification 1.3 - "Completion Times," this Note provides direction consistent with the intent of the existing Actions for the Alternate Shutdown System. As such, this change is administrative. This change is consistent with NUREG-1432. A.4 CTS Table 3.17.5 specifies the number of channels required to be Operable for Functions 19 and 20. ITS SR 3.3.8.4 requires verification that each Alternate Shutdown System control circuit and transfer switch is capable of performing its intended function every 18 months. In conjunction with the proposed Specification 1.1 definition of OPERABLE-OPERABILITY, ITS SR 3.3.8.2 provides direction consistent with the intent of the existing Specification for the Alternate Shutdown System. As such, this change is administrative. This change is consistent with NUREG-1432 .
Palisades Nuclear Plant Page 1of5 05/30/99

A.5 SPECIFICATION 3.3.8, ALTERNATE SHUTDOWN SYSTEM CTS Table 4.17.5 requires a CHANNEL FUNCTIONAL TEST and a CHANNEL ATTACHMENT3 DISCUSSION OF CHANGES CALIBRATION be performed every 18 months for functions 13, 14, and 15. CTS Table 4.17.5 also requires a CHANNEL CHECK be performed every 18 months for functions 13 and 14. Since the proposed Specification 1.1 definition of CHANNEL CALIBRATION includes a requirement to perform a CHANNEL FUNCTIONAL TEST, and a CHANNEL CALIBRATION cannot be performed without also conducting a CHANNEL CHECK, the ITS SR which requires the performance of a CHANNEL CALIBRATION every 18 months includes these CTS requirements. As such, this change is administrative. This change is consistent with NUREG-1432.
TECHNICAL CHANGES - MORE RESTRICTIVE (M) M.1 CTS 3.17.5 provides an exception to Specification 3.0.3. ITS 3.3.8 does not include an exception to ITS LCO 3.0.3. ITS LCO 3.0.3 delineates the time limits for placing the plant in a safe MODE or other specified condition when operation cannot be maintained within the limits for safe operation as defined by the LCO and its Actions. The addition of this restriction, consistent with NUREG-1432, is an additional limitation on plant operation and, therefore, a more restrictive change. This change will not adversely affect plant safety. M.2* CTS 3.17.5 provides an exception to Specification 4.0.4. ITS 3.3.8 does not include an exception to ITS SR 3.0.4. This Specification ensures that system and component OPERABILITY requirements and variable limits are met before entry into MODES or other specified conditions in the Applicability for which these systems and components ensure safe operation of the plant. The addition of this restriction, consistent with NUREG-1432, is an additional limitation on plant operation and, therefore, a more restrictive change. This change will not adversely affect plant safety. M.3 CTS 3.17.5.2 requires the reactor be placed in Hot Shutdown within 12 hours, and the reactor be placed in a condition where the affected equipment is not required, within 48 hours, if any action required by 3.17.5.1 is not met and the associated completion time has expired. ITS 3.3.8 Action B requires the plant be in MODE 3 within 6 hours and in MODE 4 within 30 hours if any Alternate Shutdown System Function is inoperable after expiration of the Completion Times in Condition A. The proposed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems . Palisades Nuclear Plant Page 2 of 5 05/30/99
ATTACHMENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.3.8, ALTERNATE SHUTDOWN SYSTEM M.3 (continued) With respect to PCS temperature requirement differences between the CTS and ITS, the proposed ITS MODE 3 is specified as being greater than 300°F, while the CTS HOT SHUTDOWN is specified as being greater than 525°F. While the ITS covers a broader range, for a shutdown the ITS Completion Time to achieve this condition is shorter, and therefore, the;phange is more restrictive. In addition, the current requirement to place the reactor in a condition where the affected equipment is not required (the Applicability is when PCS temperature is greater than or equal to 300°F) within 48 hours has been changed to be in MODE 4 in 30 hours. This change is also more restrictive since the MODE 4 condition begins at the same CTS PCS temperature cutoff of 300°F. The other parameter which is common between the CTS terms HOT SHUTDOWN and COLD SHUTDOWN and the corresponding ITS MODES 3, 4, and 5 is the reactivity condition. ITS MODES 3, 4, and 5 are defined, as a reference point, by a reactivity condition of Keff <0.99. However, in ITS Section 3.1, the equivalent amount of SHUTDOWN MARGIN is required as that specified in the CTS definitions of HOT and COLD SHUTDOWN. Therefore, the amount of SHUTDOWN MARGIN is considered to be same when the requirements of proposed ITS 3.1 are considered . Since this change will require the plant to be in a lower MODE in a shorter time frame, this .change is considered more restrictive. This change is appropriate because it continues to provide adequate time for an orderly plant shutdown without challenging plant systems. This change imposes additional restrictions on plant operation~ and is consistent with NUREG-1432. TECHNICAL CHANGES - MOVEMENT OF INFORMATION TO LICENSEE-CONTROLLED DOCUMENTS (LA) LA. l CTS Tables 3.17 .5 and 4.17 .5 contain details related to Alternate Shutdown System component identification. These details are not retained in the ITS and are relocated to the Bases. These details are not necessary to ensure OPERABILITY of the Alternate Shutdown System. ITS 3.3.8 establishes the necessary requirements to ensure Alternate Shutdown System OPERABILITY, and therefore these details are not required to be in the Technical Specifications to provide adequate protection of the public health and safety. Any changes to these requirements in the Bases will require compliance with the Bases Change Control Program, as described in ITS Section 5.0. This change is a less restrictive movement of details change with no impact on safety. This change is consistent with NUREG-1432 . Palisades Nuclear Plant Page 3 of 5 05/30/99
\
ATTACHMENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.3.8, ALTERNATE SHUTDOWN SYSTEM TECHNICAL CHANGES - LESS RESTRICTIVE (L) L.1 CTS Table 4.17.5 provides CHANNEL CHECK Surveillance Requirements for the Alternate Shutdown System instrumentation. ISTS SR 3.3.12.1 also provides CHANNEL CHECK Surveillance Requirements for Remote Shutdown System instrumentation Functions, but requires the CHANNEL CHECK to be performed only if the required instrumentation channel is normally energized. CTS requires the check regardless of the normal energization state. However, none of the Alternate Shutdown System instrumentation is normally energized. Therefore, the ISTS bracketed SR is not included in the ITS. Since this change eliminates the quarterly CHANNEL CHECKS, the change is less restrictive than CTS. This change is appropriate and acceptable since the CTS CHANNEL CHECK performance requires placing the system in service, i.e., actuating the transfer to the Alternate Shutdown System instrumentation. This performance also removes the controls from the control room during normal operation. Obviously this is not a preferred condition. The 18 month CHANNEL CALIBRATION provides the necessary confirmation of channel OPERABILITY and is considered sufficient based on industry experience that demonstrates channel failure is rare, and on the use of less formal but more frequent checks of portions of the
L.2 instrumentation channels during normal operational use of displays associated with the required channels. This change is consistent with NUREG-1432.
The two CTS 3.17.5 Actions which are required when one or more Alternate Shutdown System Function is i_noperable have been replaced by the corresponding STS Action. The CTS Actions require: a) Provide equivalent shutdown capability within 7 days, and
. b) Restore the inoperable channels to OPERABLE status within 60 _days.
The STS and ITS require: A.1 Restore required Functions to Operable status [within] 30 ~ays. With CTS Actions, equivalent shutdown capability (i.e., functionality) would be restored within 7 days, but restoration to full Operable status would not be required for 60 days. When Action 3.17.5a) was initially added to the CTS, each Function in the table carried an explicit instrument channel designator. Therefore, CTS required a) restoration of the ability to monitor the specified parameters or to provide equivalent control functions quickly, and b) restorat~on of the designated instrument channel to Operable status within 60 days .
Page 4 of 5 05/30/99 Palisades Nuclear Plant
ATTACHMENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.3.8, ALTERNATE SHUTDOWN SYSTEM L.2 (continued) ITS and STS each require restoration of Operability within 30 days, but make no requirement to restore any level of functionality before that. The omission of the requirement to restore functionality with 7 days is a Less Restrictive feature of this change. That omission is considered to be acceptable based on the requirement to more quickly restore full operability, the fact that the subject equipment is not required to perform any immediate, or automatic, safety function. There are no particular design features at Palisades which would make the Completion Times approved for inclusion in STS inappropriate for use in ITS. This change is consistent with NUREG-1432 . Palisades Nuclear Plant Page 5 of 5 05/30/99
ATTACHMENT 3
ADMINISTRATIVE CHANGES (A)
A.l DISCUSSION OF CHANGES SPECIFICATION 3.3.9, NEUTRON FLUX MONITORING CHANNELS All reformatting and renumbering are in accordance with NUREG-1432. As a result, the Technical Specifications (fS) 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 If there are less than two operable channels of nuclear instrumentation, CTS 3.17.6. lb .requires that the plant must be placed in HOT SHUTDOWN or below within 15 minutes. ITS 3.3.9 does not include this requirement. Since
the applicability of LCO 3.3.9 is "MODES 3, 4, and 5" the requirements of CTS Action 3.17.6. lb would already be met. Reguirements for nuclear instrument channels in Modes 1 and 2 are addressed by LCO 3. 3 .1 ~ This change is consistent with NUREG-1432 .

Palisades Nuclear Plant Page 1of3 05/30/99

A.3 SPECIFICATION 3.3.9, NEUTRON FLUX MONITORING CHANNELS ATTACHMENT 3 DISCUSSION OF CHANGES The CTS Applicability in CTS Table 3.17.6 for item 1 (Neutron Flux Monitoring) is "below 104 % RATED POWER with fuel in the reactor." The ITS 3.3.9 Applicability for this Function is MODES 3, 4 and 5. The CTS Applicable Conditions of "below 104 % RATED POWER with fuel in the reactor" is not directly related to ITS Modes, however, with the reactor "below 104 % RATED POWER with fuel in the reactor", the plant must be in either MODE 2, 3, 4, 5, or
6. The Wide Range Neutron Flux Monitors, which provide flux monitoring when below 104 % power, are required to be operable in MODE 2 by ITS LCO 3.3.1 as the instrumentation associated with the High Startup Rate RPS trip function and in MODE 6 by ITS 3.9.2. This LCO, combined with ITS LCOs 3.3.1 and 3.9.2, require the Neutron Flux Monitoring channels to be operable whenever reactor power is below 10-4 % RTP.
Since neutron flux monitoring in the other Modes is addressed by other Specifications within the ITS, this change is considered to only address the MODES 3, 4, and 5 part of the CTS "less than 10-4 %" applicability, this change is considered to be Administrative. A.4 CTS Table 4.17.6 item 1 (Neutron Flux Monitoring) requires a CHANNEL
FUNCTIONAL TEST to be performed in accordance with Footnote (a) which is once within 7 days prior to each reactor startup. This requirement is not included in the ITS 3.3.9 for this Function since it is identified as necessary for "reactor startup" which is MODE 2 in ITS and addressed, i.e., retained, for the High Startup Rate Function in ITS 3.3.1, "Reactor Protective System (RPS)
Instrumentation," which utilizes these wide range instrumentation channels. The Neutron Flux Monitoring Function is required in MODES 3, 4, and 5, with the appropriate accompanying conditions, as discussed in DOC A.3, above. The plant is subcritical in these MODES, therefore performing a CHANNEL FUNCTIONAL TEST within 7 days prior to a reactor startup does not address the surveillance Frequency for these MODES of operation. Since this Function provides an indication of neutron flux levels and is not required to provide an automatic actuation Function, the current 18 month Frequency for CHANNEL CALIBRATION and 12 hour Frequency for CHANNEL CHECK are considered to be sufficient to ensure the Function remains OPERABLE to provide adequate information to the operators to indicate a change of neutron flux levels. Since the CTS CHANNEL FUNCTIONAL TEST Frequency does not address the requirements in MODES 3, 4 and 5, with no more than one control rod capable of being withdrawn, where the Neutron Flux Monitoring instrumentation is required to be OPERABLE, this change is considered to be an administrative change. The proposed Frequencies are consistent with' the Surveillance Frequencies of the Post Accident Monitoring Instrumentation Frequencies in ITS 3.3.7, which also only provide indication only Functions. Therefore this change is considered to be consistent with NUREG-1432 for similar type instrumentation functions. Palisades Nuclear Plant Page 2 of 3 05/30/99
ATTACHMENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.3.9, NEUTRON FLUX MONITORING CHANNELS TECHNICAL CHANGES - MORE RESTRICTIVE (M) There were no "More Restrictive" changes associated with this specification. TECHNICAL CHANGES - MOVEMENT OF INFORMATION TO LICENSEE-CONTROLLED DOCUMENTS (LA) There were no "Removal of Details" changes associated with this specification. LESS RESTRICTIVE CHANGES (L) There were no "Less Restrictive" changes associated with this specification .
Palisades Nuclear Plant Page 3 of 3 05/30/99

ADMINISTRATIVE CHANGES (A)
ATTACHMENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.3.10, ESRV INSTRUMENTATION A.1 A separate specification is proposed to maintain current licensing basis for the Engineered Safeguards (ES) Pump Room Radiation Monitors of CTS 3.16, 3.17, and 4.17. Proposed ITS 3.3.10, "Engineered Safeguards Room Ventilation (ESRV) Instrumentation," contains, the Limiting Conditions for Operation, Applicability, Actions, and Surveillance Requirements to ensure an assumption.of the radiological consequences analysis of the Loss of Coolant Accident (LOCA) is maintained. The analysis results are based on an assumption of automatic isolation of the ES pump rooms upon detection of high radiation levels following initiation of the recirculation phase of operation. All reformatting is 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 rewordin Palisades Nuclear Plant Page 1of3 05/30/99
ATTACHMENT 4
ADMINISTRATIVE CHANGES (A)
NO SIGNIFICANT HAZARDS CONSIDERATION SECTION 3.3, INSTRUMENTATION The Palisades Nuclear Plant is converting to the Improved Technical Specifications (ITS) as outlined in NUREG-1432, "Standard Technical Specifications, Combustion Engineering Plants." Some of the proposed changes involve reformatting, renumbering, and rewording of Technical Specifications. These changes, since they do not involve technical changes to the Technical Specifications, are administrative. This type of change is connected with the movement of requirements within the current requirements, or with the modification of wording which does not affect the technical content of the current Technical Specifications. These changes will also include nontechnical modifications of requirements to conform to the Writer's Guide or provide consistency with the Improved Standard Technical Specifications in NUREG-1432. Administrative changes are not intended to add, delete, or relocate any technical requirements of the current Technical Specifications. In accordance with the criteria set forth in 10 CFR 50.92, Palisades Nuclear Plant staff has evaluated these proposed Technical Specification changes and determined they do not represent a significant hazards consideration. The following is provided in support of this conclusion .
1. Does the change involve a significantincrease in the probability or consequences of an accident previously evaluated?
The proposed changes involve reformatting, renumbering, and rewording of the existing Technical Specification. These modifications involve no technical changes to the existing Technical Specifications. The majority of changes were done in order to be consistent with NUREG-1432. During the development of NUREG-1432, certain wording preferences or English language conventions were adopted. The changes are administrative in nature and do not impact initiators of analyzed events. They also do not impact the assumed mitigation of accidents or transient events. Therefore, the changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.
2. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?
The proposed changes involve reformatting, renumbering, and rewording of the existing Technical Specifications. The changes do not involve a physical alteration of the plant (no new or different type of equipment will be installed) or changes in methods governing normal plant operation. The changes will not impose any new or different requirements or eliminate any existing requirements. Therefore, the changes
do not create the possibility of a new or different kind of accident from any accident previously evaluated.
Palisades Nuclear Plant Page 1of4 05/30/99
3.
NO SIGNIFICANT HAZARDS CONSIDERATION ATTACHMENT 4 SECTION 3.3, INSTRUMENTATION Does this change involve a significant reduction in margin of safety? The proposed changes involve reformatting, renumbering, and rewording of the existing Technical Specifications. The changes are administrative in nature and will not involve any technical changes. The changes will not reduce a margin of safety because it has no impact on any safety analysis assumptions. Also, since these changes are administrative in nature, no question of Safety is involved. Therefore, the changes do not involve a significant reduction in a margin of safety. MORE RESTRICTIVE CHANGES (M) The Palisades Nuclear Plant is converting to the Improved Technical Specifications (ITS) as outlined in NUREG-1432, "Standard Technical Specifications, Combustion Engineering Plants." Some of the proposed changes involve adding more restrictive requirements to the existing Technical Specifications by either making current requirements more stringent or by adding new requirements which currently do not exist.
These changes may include additional requirements that decrease allowed outage time, increase frequency of surveillance, impose additional surveillance, increase the scope of a specification to include additional plant equipment, increase the applicability of a specification, or provide additional actions. These changes are generally made to conform with the NUREG-1432.
In accordance with the criteria set forth in 10 CFR 50.92, Palisades Nuclear Plant staff has evaluated these proposed Technical Specification changes and determined they do not represent a significant hazards consideration. The following is provided in support of this conclusion.
1. Does the change involve a significant increase in the probability or consequences of an accident previously evaluated?
The proposed changes provide more stringent requirements than previously existed in the Technical Specifications. These more stringent requirements do not result in operation that will increase the probability of initiating an analyzed event. If anything, the new requirements may decrease the probability or consequences of an analyzed event by incorporating the more restrictive changes. The changes do not alter assumptions relative to mitigation of an accident or transient event. The more restrictive requirements continue to ensure process variables, structures, systems, and components are maintained consistent with the safety analyses and licensing basis. Therefore, the changes do not.involve a significant increase in the probability or consequences of an accident previously evaluated . Palisades Nuclear Plant Page 2 of 4 05/30/99
2.
ATTACHMENT 4 NO SIGNIFICANT HAZARDS CONSIDERATION SECTION 3.3, INSTRUMENTATION Does the change create the possibility of a new or different kind of accident from any accident previously evaluated? The proposed changes provide more stringent requirements than previously existed in the Technical Specifications. The changes do not alter the plant configuration (no new or different type of equipment will be installed) or make changes in the methods governing normal plant operation. The changes do impose different requirements.* However, these changes are consistent with the assumptions in the safety analyses and licensing basis. Therefore, the changes do not create the possibility of a new or different kind of accident from any accident previously evaluated.
3. Does this change involve a significant reduction in margin of safety?
The proposed changes provide more stringent requirements than previously existed in the Technical Specifications. Adding more restrictive requirements either increases or has no impact on the margin of safety. The changes, by definition, provide additional restrictions to enhance plant safety. The changes maintain requirements within the safety analyses and licensing basis. As such, no question of safety is involved. Therefore, the changes do not involve a significant reduction in a margin of safety . LESS RESTRICTIVE CHANGES - REMOVAL OF DETAILS TO LICENSEE CONTROLLED DOCUMENTS (LA) The Palisades Nuclear Plant is converting to the Improved Technical Specifications (ITS) as outlined in NUREG-1432, "Standard Technical Specifications, Combustion Engineering Plants." Some of the proposed changes involve moving details (engineering, procedural, etc.) out of the Technical Specifications and into a licensee controlled document. This information may be moved to the ITS Bases, FSAR, plant procedures or other programs controlled by the licensee. The removal of this information is considered to be less restrictive because it is no longer controlled by the Technical Specification change process. Typically, the information moved is descriptive in nature and its removal conforms with NUREG-1432 for format and content. In accordance with the criteria set forth in 10 CFR 50.92, Palisades Nuclear Plant staff has evaluated these proposed Technical Specification changes and determined they do not represent a significant hazards consideration. The following is provided in support of this conclusion .
Page 3 of 4 05/30/99 Palisades Nuclear Plant

1.
an accident previously evaluated? NO SIGNIFICANT HAZARDS CONSIDERATION ATTACHMENT 4 SECTION 3.3, INSTRUMENTATION Does the change involve a significant increase in the probability or consequences of Analyzed events are assumed to be initiated by the failure of plant structures, systems or components. Consequences of a previously analyzed event are dependent on the initial conditions assumed for the analysis, and the availability and successful functioning of the equipment assumed to operate in response to the analyzed event. The proposed changes move details from the Technical Specifications to a licensee controlled document. The removal of details from the Technical Specifications is not assumed to be an initiator of any analyzed event. The proposed changes do not reduce the functional requirement or alter the intent of any specification. As such, the consequences of an accident remain unchanged. Therefore, the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.
2. Does the change create the. possibility of a new or different kind of accident from any accident previously evaluated?
The proposed changes move detail from the Technical Specifications to a licensee controlled document. The changes will not alter the plant configuration (no new or different type of equipment will be installed) or make changes in methods governing normal plant operation. The changes will not impose 'different requirements, and adequate control of information will be maintained. The changes will not alter assumptions made in the safety analysis and licensing basis. Therefore, the changes

will not create the possibility of a new or different kind of accident from any accident previously evaluated.
3. Does this change involve a significant reduction in a margin of safety?
Margin of safety is determined by the design and qualification of the plant equipment, the operation of the plant within analyzed limits, and the point at which protective or mitigative actions are initiated. There are no design changes or equipment performance parameter changes associated with this change. No setpoints are affected, and no change is being proposed in the plant operational limits as a result of this change. The proposed changes remove details from the Technical Specifications and place them under licensee control. Removal of these details is acceptable since this information is not directly pertinent to the actual requirement and does not alter the intent of the requirement. Since these details are not necessary to adequately describe the actual regulatory requirement, they can be moved to licensee controlled document without a
significant impact on safety. Therefore, the proposed changes do not involve a significant reduction in a margin of safety.
Page 4 of 4 05/30/99 Palisades Nuclear Plant
ATTAC1'ENT 4 NO SIGNIFICANT HAZARDS CONSIDERATION SECTION 3.3, INSTRUMENTATION RELOCATED CHANGES (R) The Palisades Nuclear Plant is converting to the Improved Technical Specifications (ITS) as outlined in NUREG-1432, "Standard Technical Specifications, Combustion Engineering Plants." Some of the proposed changes involve relocating existing Technical Specification Requirements and Surveillances to licensee controlled documents. The Palisades Nuclear Plant has evaluated the current Technical Specifications using the criteria set forth in 10 CFR 50.36. Specifications identified by this evaluation that did not meet the retention requirements specified in the regulation are not included in the Improved Technical Specifications (ITS) submittal. These specifications have been relocated from the current Technical Specifications to the FSAR or licensee controlled documents referenced in the FSAR. Relocating requirements which do not meet the Technical Specifications criteria to licensee controlled documents allows the Technical Specifications to be reserved only for those conditions or limitations upon reactor operation which are necessary to adequately limit the possibility of an abnormal situation or event giving rise to an immediate threat to the public health and safety, thereby focusing the scope of the Technical Specifications. In accordance with the criteria set forth in 10 CFR 50.92, the Palisades Nuclear Plant has evaluated these proposed Technical Specification changes and determined they do not represent a significant hazards consideration. The following is provided in support of this conclusion.
1. Does the change involve a significant increase in the probability or consequences of an accident previously evaluated?
The proposed change relocates requirements and surveillances for structures, systems, components or variables which did not meet the criteria for inclusion in Technical Specifications. The affected structures, systems, components or variables are not assumed to be initiators of analyzed events and are not assumed to mitigate design basis accident or transient events. The requirements and surveillances for these affected structures, systems, components or variables will be relocated from the Technical Specifications to an appropriate administratively controlled document and maintained pursuant to 10 CFR 50.59. Therefore, this change does not involve a significant increase in the probability or consequences of an accident previously evaluated.
2. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?
The proposed change does not necessitate a physical alteration of the plant (no new or different type of equipment will be installed) or change in parameters governing normal plant operation. The proposed change will not impose any different requirements and
adequate control of information will be maintained. Thus, this change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

Palisades Nuclear Plant Page 1of4 05/30/99

3.
ATTACHMENT 4 NO SIGNIFICANT HAZARDS CONSIDERATION SPECIFICATION 3.3.1, RPS INSTRUMENTATION Does this change involve a significant reduction in a margin of safety? The proposed change relocates requirements and surveillances for structures, systems, components, or variables that do not meet the criteria for inclusion in Technical Specifications. The change will not reduce a margin of safety since it has no impact on any safety analysis assumptions. In addition, the relocated requirements and _surveillances for the affected structure, system, component, or variable remain the same as the existing Technical Specifications. Since any future changes to these requirements or the surveillance will be evaluated pursuant to 10 CFR 50.59, there will be no reduction in a_ margin of safety. Therefore, the change does not involve a significant reduction in the margin of safety. LESS RESTRICTIVE CHANGES (L) The Palisades Nuclear Plant is converting to the Improved Technical Specifications (ITS) as outlined in NUREG-1432, "Standard Technical Specifications, Combustion Engineering Plants." Changes have been proposed which involve making the requirements in the Current Technical Specifications (CTS) less restrictive. A description of the less restrictive change and corresponding No Significant Hazards Consideration are provided for each Specification as applicable. LESS RESTRICTIVE CHANGE L.1 CTS 3.17 .1.3 requires entry when two RPS trip units or associated instrument channels are inoperable in one or more Functions. In the ITS, these conditions are addressed in ITS 3.3.1, Condition B. The difference is that the Condition B Required Actions include a Note which excludes the applicability of LCO 3.0.4. This provision was added to allow MODE changes even though two channels are inoperable, with one channel in trip. In this condition, with one channel in trip the protection system is in a one-out-of-two logic. Since the probability of a common cause failure affecting both of the OPERABLE channels during the 7 day Completion Time to restore one trip unit to OPERABLE status is remote, and the low probability of occurrence of an event during this interval, this change is considered acceptable. This change is consistent with NUREG-1432 .
05/30/99 Palisades Nuclear Plant Page 1of9

1.
ATTAC:aMENT 4 NO SIGNIFICANT HAZARDS CONSIDERATION SPECIFICATION 3.3.1, RPS INSTRUMENTATION Does the change involve a significant increase in the probability or consequence of an accident previously evaluated? Analyzed events are assumed to be initiated by the failure of plant structures, systems or components. The proposed change allows entry into a higher mode of operation with an inoperable equipment as long as one trip unit is in trip. This will ensure the associated RPS Function will operate if an accident were to occur as long as one of two instrument channels operate as designed. Changing modes in this logic configuration is not considered to be an initiator of any evaluated accident. Therefore, allowing the plant to transfer to a higher mode of operation does not involve a significant increase in the probability of an accident previously evaluated. The consequences of a previously analyzed event are dependent on the initial conditions assumed for the analysis, and the availability and successful functioning of the equipment assumed to operate in response to the analyzed event. The proposed change allows the plant to change modes of operation with one or more RPS Functions in a one-out-of two coincidence logic. Since the plant is permitted to operate with this logic configuration at any power level for up to 7 days the consequences of an accident previously analyzed is unchanged. Therefore, the proposed change does not involve a significant increase in the consequences of an accident previously evaluated.
2. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?
The proposed change does not involve a physical alteration of the plant. No new or different type of equipment will be installed or changes made to plant parameters which govern normal plant operation. The proposed change will continue to ensure prompt restoration of an inoperable RPS Function to re-establish compliance with the limiting condition for operation. Thus, this change does not create the possibility of a new or different kind of accident from any accident previously evaluated.
3. Does this change involve a significant reduction in a margin of safety?
The margin of safety is determined by the design and qualification of the plant equipment, the operation of the plant within analyzed limits, and the point at which protective or mitigative actions are initiated. The proposed change allows entry into a higher mode of operation with inoperable instrument channels as long as one trip unit is in trip and at least one channel is rep~red in 7 days. Requiring one channel to be in trip will ensure the associated RPS Function will operate as long as one of two
instrument channels operate as designed. Since the probability of a common cause failure affecting both of the OPERABLE channels is remote during this 7 day period, this change does not involve a significant reduction in a margin of safety.
Palisades Nuclear Plant Page 2 of 9 05/30/99
LESS RESTRICTIVE CHANGE L.2 NO SIGNIFICANT HAZARDS CONSIDERATION ATTACHMENT 4 SPECIFICATION 3.3.1, RPS INSTRUMENTATION The Frequency of the CHANNEL FUNCTIONAL TEST associated with certain RPS Functions CTS Tables 4.17.1and4.17.6 is 31 days. In ITS SR 3.3.1.5, the proposed Frequency is 92 days. The proposed change revises the CHANNEL FUNCTIONAL TEST Frequency for certain RPS Functions from 31 days to 92 days. Justification for extending the test interval is provided in Combustion Engineering (CE) Topical Report CEN-327-A Supplement 1, "RPS/ESFAS Extended Test Interval Evaluation," dated January 1989. NRC acceptance of the CE topical report was provided in a letter to the CE Owner's Group (CEOG) dated November 6, 1989. As discussed in the NRC acceptance letter, Consumers Energy has reviewed the instrument drift information for each instrument channel involved and has determined that drift occurring in that channel over the extended surveillance interval will not cause the setpoint to exceed it's Allowable Value as calculated by our setpoint methodology.
Onsite records of the recorded instrument drift are available for audit. Therefore, it is proposed that the CHANNEL FUNCTIONAL TEST be performed in accordance with ITS SR 3.3.1.5, at a Frequency of 92 days. This change is consistent with NUREG-1432.
1. Does the change involve a significant increase in the probability or consequence of an accident previously evaluated?
Analyzed events are assumed to be initiated by the failure of plant structures, systems or components. Consequences of a previously analyzed event are dependent on the initial conditions assumed for the analysis, and the availability and successful functioning of the equipment assumed to operate in response to the analyzed event. The proposed change revises the surveillance Frequency of the RPS CHANNEL FUNCTIONAL TEST from 31 days to 92 days. This Frequency has been analyzed and shown to be an acceptable Frequency in detecting inoperabilities in the RPS trip units and portions of the associated instrument channels. Since the CHANNEL . FUNCTIONAL TEST Frequency is not considered to be an initiator of any accident previously evaluated and since the proposed Frequency has been determined to be an adequate Frequency in detecting inoperabilities in the associated equipment, this change does not involve a significant increase in the probability or consequence of an accident previously evaluated.
2. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?
The proposed change does not involve a physical alteration of the plant. No new or different type of equipment will be installed or changes made to plant parameters which .
govern normal plant operation. The proposed change continues to ensure the RPS instrumentation is functioning properly. Thus, this change does not create the possibility of a new or different kind of accident from any accident previously evaluated.
Palisades Nuclear Plant Page 3 of 9 05/30/99
3.
NO SIGNIFICANT HAZARDS CONSIDERATION ATTACHMENT 4 SPECIFICATION 3.3.1, RPS INSTRUMENTATION Does t,his change involve a significant reduction in a margin of safety? The margin of safety is determined by the design and qualification of the plant equipment, the operation of the plant within analyzed limits, and the point at which protective or mitigative actions are initiated. There are no design changes or equipment performance parameter changes associated with this change. No setpoints are affected, and no change is being proposed in the plant operational limits as a result of this change. Therefore, this change does not involve a significant reduction in the margin of safety. LESS RESTRICTIVE CHANGE L.3 CTS Table 4.17 .1, Footnote (c), requires that the ex core channels be calibrated with a test signal every 31 days. ITS SR 3.3.1.6 requires the surveillance to be performed every 92 days. The proposed change revises the calibration Frequency for the excore power range channels from 31 days to 92 days. This test leaves the channel adjusted to account for instrument drift between successive calibrations to ensure that the channel remains OPERABLE between calibrations. Other surveillances are performed on the excore power rang~ channels more frequently to account for overall gain and to ensure the upper and lower subchannel amplifiers are calibrated correctly to correspond to the incore detectors. These surveillances are considered sufficient to ensure the excore power range channels are functioning properly. This change is consistent with NUREG-1432.
1. Does the change involve a significant increase in the probability or consequence of an accident previously evaluated?
Analyzed events are assumed to be initiated by the failure of plant structures, systems or components. Consequences of a previously analyzed event are dependent on the initial conditions assumed for the analysis, and the availability and successful functioning of the equipment assumed to operate in response to the analyzed event. The proposed change revises the surveillance Frequency of the power range excore channel calibration from 31 days to 92 days. This Frequency has been analyzed and is considered to be adequate to account for instrument drift between successive calibrations to ensure that the channel remains operational between successive calibrations. Since the calibration frequencies are not considered to be an initiator of any accident previously evaluated and since the proposed Frequency has been determined to be adequate to account for '-instrument drift, this change does not involve a significant increase in the probability or consequence of an accident previously
evaluated .
Palisades Nuclear Plant Page 4of9 05/30/99
2.
ATTACHMENT 4 NO SIGNIFICANT HAZARDS CONSIDERATION SPECIFICATION 3.3.1, RPS INSTRUMENTATION Does the change create the possibility of a new or different kind of accident from any accident previously evaluated? The proposed change does not involve a physical alteration of the plant. No new or different type of equipment will be installed or changes made to plant parameters which govern normal plant operation. The proposed change continues to ensure the RPS instrumentation is functioning properly. Thus, this change does not create the possibility of a new or different kind of accident from any accident previously evaluated.
3. Does this change involve a significant reduction in a margin of safety?
The margin of safety is determined by the design and qualification of the plant equipment, the operation of the plant within analyzed limits, and the point at which protective or mitigative actions are initiated. There are no design changes or equipment performance parameter changes associated with this change. No setpoints are affected, and no change is being proposed in the plant operational limits as a result of this change. Therefore, this change does not involve a significant reduction in the margin of safety. LESS RESTRICTIVE CHANGE L.4 CTS Table 4.17.1 Footnote (b) requires calibration of the Variable High Power Function with heat balance when power is > 15 % RTP. ITS 3.3.1 will add a Note to SR 3.3.1.3 (heat balance) which states that the SR is not required to be performed until 12 hours after THERMAL POWER is > 15% RTP. The allowance to delay performance of the SR for 12 hours after power is > 15 % RTP provides time for the plant to achieve stable operating conditions to calibrate the instruments at a power at which the heat balance is accurate. This will provide more accurate results, and thereby providing assurance that the RPS Functions will actuate at the required setpoints. The 12 hours interval is within the 24 hour delay allowed for a missed SR. This change is consistent with NUREG-1432 . Palisades Nuclear Plant Page 5 of 9 05/30/99
1.
NO SIGNIFICANT HAZARDS CONSIDERATION ATTACHMENT 4 SPECIFICATION 3.3.1, RPS INSTRUMENTATION Does the change involve a significant increase in the probability or consequences of an accident previously evaluated? Analyzed events are assumed to be initiated by the failure of plant structures, systems or components. The RPS Functions are not considered to be the initiator of any accident. The role of this instrumentation is in mitigating and,
. thereby, limiting the consequences of analyzed events. The proposed change effectively extends the initial Surveillance Frequency until 12 hours after THERMAL POWER is ~ 15 % RTP. This allows time after the appropriate conditions are established to perform the Surveillance. The Surveillance is not required to be performed below 15% because it is difficult to accurately determine core THERMAL POWER from a heat balance at these low power levels. In addition, at low power levels, a high degree of accuracy is unnecessary due to the large inherent margin to the thermal limits at these power levels. As a result, the consequences of an accident are not affected by this change. This change will not alter assumptions relative to the mitigation of an accident or transient event. Therefore, delaying the surveillance does not involve a significant. increase in the probability or consequences of an accident previously evaluated.
2. Does the change create the possibility of a new or different kind of accident from any previously evaluated?
The proposed change does not involve a physical alteration of the plant. No new or different type of equipment will be installed or changes made to plant parameters which govern normal plant operation. The proposed change continues to ensure the RPS instrumentation is functioning properly. Thus, this change does not create the possibility of a new or different kind of accident from any accident previously evaluated . Palisades Nuclear Plant Page 6of9 05/30/99
3.
ATTACHMENT 4 NO SIGNIFICANT HAZARDS CONSIDERATION SPECIFICATION 3.3.1, RPS INSTRUMENTATION Does this change involve a significant reduction in margin of safety? The margin of safety is determined by the design and qualification of the plant equipment, the operation of the plant within analyzed limits, and the point at which protective or mitigative actions are initiated. There are no design changes or equipment performance parameter changes associated with this change. No setpoints are affected, and no change is being proposed in the plant operational limits as a result of this change. The margin of safety is not reduced by this change since the proposed change to the Surveillance Frequency provides the necessary assurance that the RPS instrumentation has been accurately calibrated at an early opportunity. This change extends the initial performance of the Surveillance Requirement to within 12 hours after reaching 15% RTP. This is considered acceptable since below 15% RTP a high degree of accuracy between the excore power instrumentation and the actual core THERMAL POWER is unnecessary due to the large inherent margin to the thermal limits at these power levels. In addition, this change provides the benefit of allowing the Surveillance to be postponed until appropriate plant
*conditions exist for performing the Surveillance accurately. The safety analysis assumptions will still be maintained thus no question of safety exists.
Therefore, this change does not involve a significant reduction in a margin of safety. LESS RESTRICTIVE CHANGE L.5 CTS 4.18.2. lb requires calibration of the excore monitoring system with the incore monitoring system at least every 31 days of power operation. ITS 3.3.1.4 requires the same surveillance, but is modified by a Note which states that the SR is not required to be performed until 12 hours after THERMAL POWER is~ 25% RTP. Since the CTS "power operations" is defined as "greater than 2 %, " this change is considered less restrictive. This change is acceptable because it eliminates uncertainties in the excore and incore measurement process at low power levels. The allowance to delay the surveillance for 12 hours after power is
~ 25 % RTP provides time for the plant to stabilize at a power level sufficiently high enough to accurately calibrate the instruments. This will yield more accurate results, and thereby provide additional assurance that the RPS Functions will actuate at the required setpoints. The 12 hours interval is within the 24 hour delay allowed for a missed SR. This change is consistent with NUREG-1432 .
Palisades Nuclear Plant Page 7 of 9 05/30/99
1.
NO SIGNIFICANT HAZARDS CONSIDERATION ATTACHMENT 4 SPECIFICATION 3.3.1, RPS INSTRUMENTATION Does the change involve a significant increase in the probability or consequences of an accident previously evaluated? Analyzed events are assumed to be initiated by the failure of plant structures, systems or components. The RPS Functions are not considered to be the initiator of any accident. The role of this instrumentation is in mitigating and, thereby, limiting the consequences of analyzed events. The proposed change effectively extends the initial Surveillance Frequency until 12 hours after THERMAL POWER is ;,: 25 % RTP. This allows time after the appropriate conditions are established to perform the Surveillance. The Surveillance is not required to be performed below 25 % RTP due to the uncertainties in the excore and incore measurement process at lower power levels. In addition, at low power levels, a high degr~ of accuracy is unnecessary due to the large inherent margin to the thermal limits at these power levels. As a result, the consequences of an accident are not affected by this change. This change will not alter assumptions relative to the mitigation of an accident or transient event. Therefore, delaying the surveillance does not involve a significant increase in the probability or consequences of an accident previously evaluated .
2. Does the change create the possibility of a new or different kind of accident from any previously evaluated?
The proposed change does not involve a physical alteration of the plant. No new or different type of equipment will be installed or changes made to plant parameters which govern normal plant operation. The proposed change continues to ensure the RPS instrumentation is functioning properly. Thus, this change does not create the possibility of a new or different kind of accident from any accident previously evaluated . Palisades Nuclear Plant Page 8 of 9 05/30/99
3.
NO SIGNIFICANT HAZARDS CONSIDERATION ATTACHMENT 4 SPECIFICATION 3.3.1, RPS INSTRUMENTATION Does this change involve a significant reduction in margin of safety? The margin of safety is determined by the design and qualification of the plant equipment, the operation of the plant within analyzed limits, and the point at which protective or mitigative actions are initiated. There are no design changes or equipment performance parameter changes associated with this change. No setpoints are affected, and no change is being proposed in the plant operational limits as a result of this change. The margin of safety is not reduced by this change since the proposed change to the Surveillance Frequency provides the necessary assurance that the RPS instrumentation has been accurately calibrated at an early opportunity. This change extends the initial performance of the Surveillance Requirement to within 12 hours after reaching 25 % RTP. This is considered acceptable since below 25 % RTP a high degree of accuracy between the excore power and incore instrumentation is unnecessary due to the large inherent margin to the-thermal limits at these power levels. In addition, this change provides the benefit of allowing the Surveillance to be postponed until appropriate plant conditions exist for performing the Surveillance accurately. The safety analysis assumptions will still be maintained thus no question of safety exists.
Therefore, this change does not involve a significant reduction in a margin of safety .
Palisades Nuclear Plant Page 9 of 9 05/30/99
LESS RESTRICTIVE CHANGE L.1 NO SIGNIFICANT HAZARDS CONSIDERATION ATTACHMENT 4 SPECIFICATION 3.3.2, RPS LOGIC AND TRIP INITIATION The Frequency of the CHANNEL FUNCTIONAL TEST associated with the RPS Logic (i.e., Matrix Logic and Trip Initiation Logic) in CTS 4 .17. 1 is 31 days. In ITS SR 3. 3. 2. 1, the proposed Frequency is 92 days. The proposed change revises the surveillance Frequency of the RPS Matrix Logic and Initiation Logic CHANNEL FUNCTIONAL TEST from 31 days to 92 days. Justification for extending the test interval is provided in Combustion Engineering (CE) Topical Report CEN-327-A Supplement 1, "RPS/ESFAS Extended Test Interval Evaluation," dated January 1989. NRC acceptance of the CE topical report was provided in a letter to the CE Owner's Group (CEOG) dated November 6, 1989. As discussed in the NRC acceptance letter, Consumers Energy has reviewed the instrument drift information for each instrument channel involved and has determined that drift occurring in that channel over the extended surveillance interval will not cause the setpoint to exceed it's Allowable Value as calculated by our setpoint methodology. Onsite records of the recorded instrument drift are available for audit. Therefore the CHANNEL FUNCTIONAL TEST will performed in accordance with ITS SR 3.3.2.1 at a 92 day Frequency. This change is consistent with NUREG-1432.

1* Does the change involve a significant increase in the probability or consequence of an accident previously evaluated?
Analyzed events are assumed to be initiated by the failure of plant structures, systems or components. Consequences of a previously analyzed event are dependent on the initial conditions assumed for the analysis, and the availability and successful functioning of the equipment assumed to operate in response to the analyzed event. The proposed change revises the surveillance Frequency of the RPS Matrix Logic and Initiation Logic CHANNEL FUNCTIONAL TEST from 31 days to 92 days. This Frequency has been analyzed and shown to be an acceptable Frequency in detecting inoperabilities in the Matrix and Initiation logic channels. Since the CHANNEL FUNCTIONAL TEST Frequency is not considered to be an initiator of any accident previously evaluated and since the proposed Frequency has been determined to be an adequate Frequency in detecting inoperabilities in the associated equipment, this change does not involve a significant increase in the probability or consequence of an accident previously evaluated.
2. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?
The proposed change does not involve a physical alteration of the plant. No new or different type of equipment will be installed or changes made to plant parameters which govern normal plant operation. The proposed change continues to ensure the RPS Matrix and Initiation Logic instrumentation is functioning properly. Thus, this change does not create the possibility of a new or different kind of accident from any accident previously evaluated. Palisades Nuclear Plant Page 1of2 05/30/99
3.
NO SIGNIFICANT HAZARDS CONSIDERATION ATTACHMENT 4 SPECIFICATION 3.3.2, RPS LOGIC AND TRIP INITIATION Does this change involve a significant reduction in a margin of safety? The margin of safety is determined by the design and qualification of the plant equipment, the operation of the plant within analyzed limits, and the point at which
protective or mitigative actions are initiated. There are no design changes or equipment performance parameter changes associated with this change. No setpoints are affected,

and no change is being proposed in the plant operational limits as a result of this change. Therefore, this change does not involve a significant reduction in the margin of safety .
Palisades Nuclear Plant Page 2of2 05/30/99
LESS RESTRICTIVE CHANGE L.1 ATTACHMENT 4 NO SIGNIFICANT HAZARDS CONSIDERATION SPECIFICATION 3.3.3, ESF INSTRUMENTATION CTS Actions 3.17.2.3 and 3.17.3.3 are required to be entered when two ESF (except for SIRWT) or isolation instrumentation channels are inoperable in one or more Functions. In the ITS, these conditions are addressed in ITS 3.3.3 Condition B. The difference is that the ITS Required Actions are modified by a Note which excludes the applicability of LCO 3.0.4. This provision was added to allow MODE changes even though two channels are inoperable, and with one channel in trip. In this condition, with one channel in trip the protection system is in a one-out-of-two logic. This change is considered acceptable based on the low probability of a common cause failure affecting both of the OPERABLE channels during the 7 day Completion Time, and the low probability of an event occurring during this time. This change is consistent with NUREG-1432.
1. Does the change involve a significant increase in the probability or consequence of an accident previously evaluated?
Analyzed events are assumed to be initiated by the failure of plant structures, systems or components .. The proposed change allows entry into a higher MODE of operation with inoperable equipment as long as one channel is in trip. This will ensure the associated ESF Function will operate if an accident were to occur as long as one of two instrument channels operate as designed. Changing MODES in this logic configuration is not considered to be an initiator of any evaluated accident. Therefore, allowing the plant to change to a higher MODE of operation does not involve a significant increase in the probability of an accident previously evaluated. The consequences of a previously analyzed event are dependent on the initial conditions assumed for the analysis, and the availability and successful functioning of the equipment assumed to operate in response to the analyzed event. The proposed change allows the plant to change MODES with one or more ESF Functions in a one-out-of-two coincidence logic. Since the plant is permitted to operate with this logic configuration at any power level for up to 7 days the consequences of an accident previously analyzed is unchanged. Therefore, the proposed change does not involve a significant increase in the consequences of an accident previously evaluated.
2. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?
The proposed change does not involve a physical alteration of the plant. No new or different type of equipment will be installed or changes made to plant parameters which govern normal plant operation. The proposed change will continue to ensure prompt restoration of an inoperable ESF Function to re-establish compliance with the limiting condition for operation. Thus, this change does not create the possibility of a new or different kind of accident from any accident previously evaluated. Palisades Nuclear Plant Page 1of5 05/30/99
3.
ATTACHMENT 4 NO SIGNIFICANT HAZARDS CONSIDERATION SPECIFICATION 3.3.3, ESF INSTRUMENTATION Does this change involve a significant reduction in a margin of safety? The margin of safety is determined by the design and qualification of the plant equipment, the operation of the plant within analyzed limits, and the point at which protective or mitigative actions are initiated. The proposed change allows entry into a higher MODE of operation with inoperable instrument channels as long as one channel is in trip and at least one channel is restored to OPERABLE status in 7 days. Requiring one channel to be in trip will ensure the associated ESF Function will operate as long as one of two instrument channels operate as designed. Since the probability of a common cause failure affecting both of the OPERABLE channels is remote during this 7 day period, this change does not involve a significant reduction in a margin of safety. LESS RESTRICTIVE CHANGE L.2 CTS Tables 4.17.2 and 4.17.3 require performance of a CHANNEL FUNCTIONAL TEST at a Frequency of 31 days. ITS SR 3.3.3.2 requires performance of a CHANNEL FUNCTIONAL TEST at a proposed Frequency of 92 days. The proposed change revises the surveillance Frequency of the ESF instrumentation CHANNEL FUNCTIONAL TEST from 31 days to 92 days. Justification for extending the test interval is provided in Combustion Engineering (CE) Topical Report CEN-327-A Supplement 1, "RPS/ESFAS Extended Test Interval Evaluation," dated January 1989. NRC acceptance of the CE topical report was provided in a letter to the CE Owner's Group (CEOG) dated November 6, 1989. As discussed in the NRC acceptance letter, Consumers Energy has reviewed the instrument drift information for each instrument channel involved and has determined that drift occurring in that channel over the extended surveillance interval will not cause the setpoint to exceed it's Allowable Value as calculated by our setpoint methodology. Onsite records of the recorded instrument drift are available for audit. Therefore, the CHANNEL FUNCTIONAL TEST will performed in accordance with ITS SR 3.3.3.2 at a Frequency of 92 days. This change is consistent with NUREG-1432 . Palisades Nuclear Plant Page 2 of 5 05/30/99
1.
NO SIGNIFICANT HAZARDS CONSIDERATION ATTACHMENT 4 SPECIFICATION 3.3.3, ESF INSTRUMENTATION Does the change involve a significant increase in the probability or consequence of an accident previously evaluated? Analyzed events are assumed to be initiated by the failure of plant structures, systems or components. Consequences of a previously analyzed event are dependent on the initial conditions assumed for the analysis, and the availability and successful functioning of the equipment assumed to operate in response to the analyzed event. The proposed change revises the surveillance Frequency of the ESF CHANNEL FUNCTIONAL TEST from 31 days to 92 days. This Frequency has been analyzed and shown to be an acceptable Frequency in detecting inoperabilities in the ESF trip units and portions of the associated instrument channels. Since the CHANNEL FUNCTIONAL TEST Frequency is not considered to be an initiator of any accident previously evaluated and since the proposed Frequency has been determined to be an adequate Frequency in detecting inoperabilities in the associated equipment, this change does not involve a significant increase in the probability or consequence of an accident previously evaluated.
2* Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?
The proposed change does not involve a physical alteration of the plant. No new or different type of equipment will be installed or changes made to plant parameters which govern normal plant operation. The proposed change continues to ensure the ESF instrumentation is functioning properly. Thus, this change does not create the possibility of a new or different kind of accident from any accident previously evaluated.
3. Does this change involve a significant reduction in a margin of safety?
The margin of safety is determined by the design and qualification of the plant equipment, the operation of the plant within analyzed limits, and the point at which protective or mitigative actions are initiated. There are no design changes or equipment performance parameter changes associated with this change. No setpoints are affected, and no change is being proposed in the plant operational limits as a result of this change. Therefore, this change does not involve a significant reduction in the margin of safety . Palisades Nuclear Plant Page 3 of 5 05/30/99
LESS RESTRICTIVE CHANGE L.3 NO SIGNIFICANT HAZARDS CONSIDERATION ATTACHMENT 4 SPECIFICATION 3.3.3, ESF INSTRUMENTATION CTS 3 .17. 3 requires the ESF Instrumentation Function of Steam Generator Low Pressure (SGLP) to be OPERABLE "when the PCS is above COLD SHUTDOWN." For ITS, this would be reflected as an Applicability of MODES 1, 2, 3, and 4. However, the purpose of the SGLP Function is to provide automatic isolation of the main steam and main feedwater lines following a Main Steam Line Break (MSLB) or Main Feedwater Line Break (MFLB).
CTS 3.5. lf requires that the Main Steam Isolation Valves (MSIVs) be OPERABLE when PCS is above 300°F, i.e., ITS MODES 1, 2, and 3. Also, CTS does not include any specific requirements for OPERABILITY of valves which isolate main feed water, i.e., the MFIVs. ITS 3.7.2 requires the MSIVs to be OPERABLE during MODE 1, and during MODES 2 and 3 except when all MSIVs are closed and de-activated. ITS 3.7.3 requires the MFIVs to be OPERABLE during MODES 1, 2 and 3 except when all MFIVs are closed and de-activated, or are isolated by closed manual valves. Since the automatic isolation function will have no purpose if the valve is not required to be OPERABLE, the CTS is revised such that the ITS 3.3.3 Applicability for the SGLP Function is MODES 1, 2, and 3 with MODES 2 and 3 modified by a Note .
The Note will indicate that the SGLP is not required when all MSIV s are closed and de-activated, and all MFIVs are closed. and de-activated, or are isolated by closed manual valves. This Note makes the automatic isolation function Applicability consistent with the Applicability of the components actuated by the instrumentation Function.
1. Does the change involve a significant increase in the probability or consequence of an accident previously evaluated?
Analyzed events are assumed to be initiated by the failure of plant structures, systems or components. The proposed change removes Applicability of MODE 4 for the SGLP Function of the ESFI, and part of MODE 3 when* all main steam and main feedwater penetrations are properly isolated. This provides consistent Applicability requirements between the instrumentation that provides automatic closure signals and the valves which are closed by the automatic closure signals. Since the isolation valves are not required to be OPERABLE in the omitted MODEs, the SGLP Function of the ESFI has no safety function to perform in these MODES. Therefore, removing these MODES of Applicability for the SGLP Function of the ESFI does not involve a significant increase in the probability of an accident previously evaluated. The consequences of a previously analyzed event are dependent on the initial conditions assumed for the analysis, and the availability and successful functioning of the equipment assumed to operate in response to the analyzed event. The proposed change
removes Applicability of the SGLP Function of the ESFI during conditions when the safety function has already been completed or the steam generator energy is sufficiently low that the safety function is not required. Therefore, the proposed change does not involve a significant increase in the consequences of an accident previously evaluated.
Palisades Nuclear Plant Page 4 of 5 05/30/99
2.
ATTACHMENT 4 NO SIGNIFICANT HAZARDS CONSIDERATION SPECIFICATION 3.3.3, ESF INSTRUMENTATION Does the change create the possibility of a new or different kind of accident from any accident previously evaluated? The proposed change does not involve a physical alteration of the plant. No new or different type of equipment will be installed or changes made to plant parameters which govern normal plant operation. The proposed change will continue to ensure the SGLP
. Function of the ESFI is OPERABLE when needed to actuate isolation of the main steam and main feedwater lines. Thus, this change does not create the possibility of a new or different kind of accident from any accident previously evaluated.
3. Does this change involve a significant reduction in a margin of safety?
The margin of safety is determined by the design and qualification of the plant equipment, the operation of the plant within analyzed limits, and the point at which protective or mitigative actions are initiated. The proposed change maintains the safety function requirements during all conditions under which it may be needed. Therefore, this change does not involve a significant reduction in a margin of safety . Palisades Nuclear Plant Page 5 of 5 05/30/99
NO SIGNIFICANT HAZARDS CONSIDERATION SPECIFICATION 3.3.4, ESF MANUAL AND LOGIC INITIATION LESS RESTRICTIVE CHANGE L.1 ATTACHMENT4 CTS 3 .17. 3 requires the ESF Instrumentation Function of Steam Generator Low Pressure (SGLP) to be OPERABLE "when the PCS is above COLD SHUTDOWN." For ITS, this would be reflected as an Applicability of MODES 1, 2, 3, and 4. However, the purpose of the SGLP Function is to provide automatic isolation of the main. steam and main feed water lines following a Main Steam Line Break (MSLB) or Main Feedwater Line Break (MFLB).
CTS 3.5. lf requires that the Main Steam Isolation Valves (MSIVs) be OPERABLE when PCS is above 300°F, i.e., ITS MODES 1, 2, and 3. Also, CTS does not include any specific requirements for OPERABILITY of valves which isolate main feedwater, i.e., the MFIVs. ITS 3.7.2 requires the MSIVs to be OPERABLE during MODE 1, and during MODES 2 and 3 except when all MSIVs are closed and de-activated. ITS 3.7.3 requires the MFIVs to be OPERABLE during MODES 1, 2 and 3 except when all MFIVs are closed and de-activated, or are isolated by closed manual valves. Sin~e the automatic isolation function will have no purpose if the valve is not required to be OPERABLE, the CTS is revised such that the ITS 3.3.3 Applicability for the SGLP Function is MODES 1, 2, and 3 with MODES 2 and 3 modified by a Note. The Note will indicate that the SGLP is not required when all MSIVs are
closed and de-activated, and all MFIVs are closed and de-activated, or are isolated by closed manual valves. This Note makes the automatic isolation function Applicability consistent with the Applicability of the components actuated by the instrumentation Function.
1. Does the change involve a sig~iflcant increase in the probability or consequence of an accident previously evaluated?
Analyzed events are assumed to be initiated by the failure of plant structures, systems or components. The proposed change removes Applicability of MODE 4 for the SGLP Function of the ESFI, and part of MODE 3 when all main steam and main feedwater penetrations are properly isolated. This provides consistent Applicability requirements between the instrumentation that provides automatic closure signals and the valves which are closed by the automatic closure signals. Since the isolation valves are not required to be OPERABLE in the omitted MODES, the SGLP Function of the ESFI has no safety function to perform in these MODES. Therefore, removing these MODES of Applicability for the SGLP Function of the ESFI does not involve a significant increase in the probability of an accident previously evaluated. The consequences of a previously analyzed event are dependent on the initial conditions* assumed for the analysis, and the availability and successful functioning of the equipment assumed to operate in response to the analyzed event. The proposed change removes Applicability of the SGLP Function of the ESFI during conditions when the safety function has already been completed or the steam generator energy is sufficiently low that the safety function is not required. Therefore, the proposed change does not involve a significant increase in the consequences of an accident previously evaluated. Palisades Nuclear Plant Page 1of2 05/30/99
2.
SPECIFICATION 3.3.4, ESF MANUAL AND LOGIC INITIATION ATTACHMENT 4 NO SIGNIFICANT HAZARDS CONSIDERATION Does the change create the possibility of a new or different kind of accident from any accident previously evaluated? The proposed change does not involve a physical alteration of the plant. No new or different type of equipment will be installed or changes made to plant parameters which govern normal plant operation. The proposed change will continue to ensure the SGLP Function of the ESFI is OPERABLE when needed to actuate isolation of the main steam and main feedwater lines. Thus, this change does not create the possibility of a new or different kind of accident from any accident previously evaluated.
3. Does this change involve a significant reduction in a margin of safety?
The margin of safety is determined by the design and qualification of the plant equipment, the operation of the plant within analyzed limits, and the point at which protective or mitigative actions are initiated. The proposed change maintains the safety function requirements during all conditions under which it may be needed. Therefore, this change does not involve a significant reduction in a margin of safety . Palisades Nuclear Plant Page 2 of 2 05/30/99
LESS RESTRICTIVE CHANGES (L)
NO SIGNIFICANT HAZARDS CONSIDERATION SPECIFICATION 3.3.5, DG UNDERVOLTAGE START ATTACHMENT 4 There were no "Less Restrictive" changes associated with this specification . Palisades Nuclear Plant Page 1of1 05/30/99
ATTACHMENT 4 NO SIGNIFICANT HAZARDS CONSIDERATION SPECIFICATION 3.3.6, REFUELING CHR INSTRUMENTATION LESS RESTRICTIVE CHANGE L.1 CTS 3.8. lc requires the containment vent and purge system radiation monitors to be tested and verified OPERABLE "immediately prior to refueling operations." In addition, CTS Table 4.17.6 requires a CHANNEL CHECK every 24 hours, a CHANNEL FUNCTIONAL TEST every 31 days and a CHANNEL CALIBRATION every 18 months. Proposed ITS 3.3.6 and SR 3.0.4 require the OPERABILITY of each radiation monitor prior to entering the MODE of Applicability (e.g., prior to CORE ALTERATIONS or the movement of irradiated fuel assemblies within the containment). The CTS has been revised to delete the "immediately prior to refueling operation" requirement and only require the CTS Table 4.17.6 surveillances. This is acceptable since the ITS provides general rules for the application of SRs in the Technical Specifications. SR 3.0.4 establishes the requirement that the applicable SRs must be met before entry into a MODE or other specified condition in the Applicability.
In addition, the specific time frames and condition necessary for meeting the SRs are specified in the Frequency. Although the phrase "immediately prior to refueling operations" implies a conditional type frequency, proposed SRs 3.3.6.1, 3.3.6.2, and 3.3.6.3 specify a specific Frequency. The proposed surveillances in these SRs are consistent with similar testing for ESF instrumentation which is considered to be acceptable. This change is consistent with
NUREG-1432.
1. Does the change involve a significant increase in the probability or consequence of an accident previously evaluated?
Analyzed events are assumed to be initiated by the failure of plant structures, systems or components. The propqsed change revises the time frame for demonstrating the OPERABILITY of the Refueling Containment High Radiation (CHR) channels prior to entering the MODE of Applicability. Confirming OPERABILITY of the instrumentation prior to entering the MODE of Applicability does not have a detrimental impact on the integrity of any plant structure, system or component. This relaxation will not alter the operation of any plant equipment, or otherwise increase its failure probability. The probability that equipment failures will result in occurrence of an analyzed event is unrelated to a component which initiates a protective action. As such, the probability of occurrence for a previously analyzed accident is not significantly increased. The consequences of a previously analyzed event are dependent on the initial conditions assumed for the analysis, and the availability and successful functioning of the equipment assumed to operate in response to the analyzed event, and the setpoints at which these actions are initiated. Performance of the surveillance requirement confirming the OPERABILITY of the Refueling CHR channels prior to entering the
MODE of Applicability ensures that the assumptions of the safety analysis are met and does not affect the performance of any credited equipment. As a result, no analyses assumptions are violated. Based on this evaluation, there is no significant increase in the consequences of a previously analyzed event.
Palisades Nuclear Plant Page 1of4 05/30/99
2.
SPECIFICATION 3.3.6, REFUELING CHR INSTRUMENTATION ATTACHMENT 4 NO SIGNIFICANT HAZARDS CONSIDERATION Does the change create the possibility of a new or different kind of accident from any accident previously evaluated? The proposed change does not involve a physical alteration of the plant. No new equipment is being introduced, and no installed equipment is being operated in a new or different manner. There is no alteration to the parameters within which the plant is
normally operated or in the setpoints which initiate protective or mitigative actions. No change is being proposed to the procedures governing normal plant operation or those procedures relied upon to mitigate a design basis event. Relaxing the requirement to confirm the OPERABILITY of the Refueling CHR channels prior to entering the MODE of Applicability does not have a detrimental impact on the manner in which plant equipment operates or *responds to an actuation signal. As such, no new failure modes are being introduced. In addition, the change does not alter assumptions made in the safety analyses and licensing basis. Therefore, the change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

3* Does this change involve a significant reduction in a margin of safety?
The margin of safety is determined by the design and qualification of the plant equipment, the operation of the plant within analyzed limits, and the point at which protective or mitigative actions are initiated. Relaxing the requirement to confirm OPERABILITY of the Refueling CHR channels does not significantly impact these factors. There are no design changes or equipment performance parameter changes associated with this change. No setpoints are affected, and no change is being proposed in the plant operational limits as a result of this change. Therefore, this change does not involve a significant reduction in the margin of safety. LESS RESTRICTIVE CHANGE L.2 CTS 3.8.2 requires that if any of the conditions of CTS 3.8.1 are not met, all refueling operations shall cease immediately. For ITS LCO 3.3.6, when one required channel of automatic (or manual) actuation is inoperable, that Required Action has been replaced with Actions which require placing the inoperable channel in trip (i.e., performing its safety function) or removing the plant from the applicable conditions of the LCO. The completion time has been extended from immediately to 4 hours. The extension of the completion time from "immediately" to 4 hours is offset by the additional requirement to have two manual actuation channels Operable. The four hour completion time only applies when at least one channel of manual actuation and one channel of automatic actuation remain Operable. Palisades Nuclear Plant Page 2 of 4 05/30/99
ATTACHMENT 4 NO SIGNIFICANT HAZARDS CONSIDERATION SPECIFICATION 3.3.6, REFUELING CHR INSTRUMENTATION LESS RESTRICTIVE CHANGE L.2 (continued)
With two channels inoperable, ITS 3.3.6 retains the CTS action of removing the plant from the applicable conditions of the LCO immediately. CTS 3.8.2 further requires that, "work shall be initiated to satisfy the required conditions and that no operations that may change the reactivity of the core shall be made." These requirements are not appropriate once the Applicability for the LCO has been exited. Therefore, they have been deleted. The ITS retains the pertinent, and sufficient, requirements in LCO 3.3.6 to assure the OPERABILITY of the Refueling CHR instrumentation when it is needed to support the reduction of dose consequences following a fuel handling accident. Therefore, this deletion can be made with no impact to the health and safety of the public. This change is consistent with NUREG-1432.
1. Does the change involve a significant increase in the probability or consequence of
an accident previously evaluated?
Analyzed events are assumed to be initiated by the failure of plant structures, systems or components. The proposed change extends the allowed action time when one required automatic actuation c~annel is inoperable, and omits actions that are not necessary to assure safety once the Applicability for the LCO has been exited. The four hour completion time extension only applies when at least one channel of manual actuation and one channel of automatic actuation remain Operable, therefore, there is no potential for loss of isolation capability. Once the CORE ALTERATIO NS and movement of irradiated fuel within the containment have ceased, there is no longer a need for the Refueling CHR Instrumentation to protect against a fuel handling accident. Therefore, additional Required Actions are inappropriate. Therefore, this relaxation will not alter the operation of any plant equipment, or otherwise increase its failure probability. The probability that equipment failures will result in occurrence of an analyzed event is unrelated to a component which initiates a protective action. As such, the probability of occurrence for a previously analyzed accident is not significantly increased . Palisades Nuclear Plant Page 3 of 4 05/30/99
NO SIGNIFICANT HAZARDS CONSIDERATION SPECIFICATION 3.3.6, REFUELING CHR INSTRUMENTATION LESS RESTRICTIVE CHANGE L.2 ATTACHMENT 4
1. (continued)
The consequences of a previously analyzed event are dependent on the initial conditions assumed for the analysis, and the availability and successful functioning of the equipment assumed to operate in response to the analyzed event, and the setpoints at which these actions are initiated. The extended allowed outage time does not affect the consequence of any previously analyzed event, because the ability for both automatic and manual closure of containment penetrations is required for this extension to be used. Performance of actions outside conditions under which the fuel handling accident can occur, i.e., the Applicability, is not required to ensure that the assumptions of the safety analysis are met and does not affect the performance of any credited equipment. As a result, no analyses assumptions are violated. Based on this evaluation, there is no significant increase in the consequences of a previously analyzed event.
2. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?
The proposed change does not involve a physical alteration of the plant. No new equipment is being introduced, and no installed equipment is being operated in a new or different manner. There is no alteration to the parameters within which the plant is normally operated or in the setpoints which initiate protective or mitigative actions. No change is being proposed to the procedures governing normal plant operation or those procedures relied upon to mitigate a design basis event. Extending the completion time for degraded isolation capability or relaxing the action requirements while outside the MODE of Applicability does not have a detrimental impact on the manner in which plant equipment operates or responds to an actuation signal. As such, no new failure MODES are being introduced. In addition, the change does not alter assumptions made in the safety analyses and licensing basis. Therefore, the change does not create the possibility of a new or different kind of accident from any accident previously evaluated.
3. Does this change involve a significant reduction in a margin of safety?
The margin of safety is determined by the design and qualification of the plant equipment, the operation of the plant within analyzed limits, and the point at which protective or mitigative actions are initiated. Extending the completion time for degraded isolation capability or relaxing the action requirements while outside the MODE of Applicability does not significantly impact these factors. There are no design changes or equipment performance parameter changes associated with this change. No setpoints are affected, and no change is being proposed in the plant operational limits as a result of this change. Therefore, this change does not involve a significant reduction in the margin of safety. Palisades Nuclear Plant Page 4 of 4 05/30/99
LESS RESTRICTIVE CHANGE L.1 NO SIGNIFICANT HAZARDS CONSIDERATION ATTACHMENT 4 SPECIFICATION 3.3.7, PAM INSTRUMENTATION The ITS Actions contain completion times which are longer than their CTS counterparts; ITS Action A. l allows 30 days for restoration of a single inoperable channel for one or more Functions; CTS 3.3.4 Actions la, 3a, and 5a allow 7 days for this restoration. The longer ITS completion times are taken from the STS. They are based on the additional operating experience available when the STS were written and take into account the passive nature of the instruments (no automatic action is assumed to occur from these instruments), and the low probability of an event requiring accident monitoring instrumentation during this interval. The 30 day Completion Time for restoration of a single inoperable channel (for one or more Functions) also considers the remaining Operable channel. This change is consistent with NUREG 1432.
1. Does the change involve a significant increase in the probability or consequences of an accident previously evaluated?
The proposed change revises the time available to restore an inoperable PAM instrumentation channel from 7 days to 30 days before Action is required. PAM instruments are not initiators of any analyzed event. The other Operable PAM instrumentation channel is available for indication, and the PAM instruments are only required to provide indication, no active function. The proposed change does not significantly affect initiators or mitigation of analyzed events, and therefore does not involve a significant increase in the probability or consequences of an accident previously evaluated.
2. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?
The proposed change revises the time available to restore an inoperable PAM instrumentation channel from 7 days to 30 days before Action is required. This change will not involve a significant change in the design or operation of the plant. No hardware is being added to the plant as part of the proposed change. The proposed change will not introduce any new accident initiators. Therefore, the change does not create the possibility of a new or different kind of accident from any accident previously evaluated.
3. Does the change involve a significant reduction in a margin of safety?
The proposed change revises the time available to restore an inoperable PAM
instrumentation channel from 7 days to 30 days before Action is required. The other OPERABLE PAM instrument channel is available for indication, and the PAM instruments are only required to provide indication, no active function. Therefore, the change does not involve a significant reduction in the margin of safety.
Palisades Nuclear Plant Page 1of9 05/30/99
LESS RESTRICTIVE CHANGE L.2 ATTACHMENT 4 NO SIGNIFICANT HAZARDS CONSIDERATION SPECIFICATION 3.3.7, PAM INSTRUMENTATION The ITS Actions contain completion times which are longer than their CTS counterparts; ITS Action D.1 allows 3 days for. restoration for one of two inoperable channels of Hydrogen Monitoring; CTS Action 2a allows 2 days for this restoration. The longer ITS completion times are taken from the STS. They are based on the additional operating experience available when the STS were written and take into account the passive nature of the instruments (no automatic action is assumed to occur from these instruments), and the low probability of an event requiring accident monitoring instrumentl:tion during this interval. These changes are consistent with NUREG 1432.
1. Does the change involve a significant increase in the probability or consequences of an accident previously evaluated?
The proposed change revises the allowed outage time from 48 hours to 72 hours when 2 channels of hydrogen monitor instrumentation are inoperable. The hydrogen monitor instrumentation is not an initiator of any analyzed event. The hydrogen monitor instrumentation does not provide any safety function in the mitigation of analyzed
events. This instrument provides information to the operators, who are functioning as a backup to the automatic systems designed to mitigate accidents previously evaluated.
The proposed change does not significantly affect initiators or mitigation of analyzed events, and therefore does not involve a significant increase in the probability or consequences of an accident previously evaluated.
2. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?
The proposed change revises the allowed outage time from 48 hours to 72 hours when 2 channels of hydrogen monitor instrumentation are inoperable. The change will not involve a significant change in design or operation of the plant. No hardware is being added to the plant as part of the proposed change. The proposed change will not introduce any new accident initiators. Therefore, the change does not create the possibility of a new or different kind of accident from any accident previously evaluated.
3. Does the change involve a significant reduction in a margin of safety?
The proposed change revises the allowed outage time from 48 hours to 72 hours when 2 channels of hydrogen monitor instrumentation are inoperable. The hydrogen monitor instrumentation does not provide any safety function in the mitigation of analyzed events. Also, the likelihood is remote that an event would occur which would require
the hydrogen monitor instrumentation. Therefore, the proposed change does not involve a significant reduction in a margin of safety.
Palisades Nuclear Plant Page 2 of 9 05/30/99
LESS RESTRICTIVE CHANGE L.3 ATTACHMENT 4 NO SIGNIFICANT HAZARDS CONSIDERATION SPECIFICATION 3.3.7, PAM INSTRUMENTATION The ITS Actions contain completion times which are longer than their CTS counterparts; ITS Action C.1 allows 7 days for restoration for one of two inoperable channels (for Functions other than Hydrogen Monitoring); CTS Actions 2a and 6a allow 2 days for this restoration.
The longer ITS completion times are taken from the STS. They are based on the additional operating experience available when the STS were written and take into account the passive nature of the instruments (no automatic action is assumed to occur from these instruments), and the low probability of an event requiring accident monitoring instrumentation during this interval. This change is consistent with NUREG 1432.
1. Does the change involve a significant increase in the probability or consequences of an accident previously evaluated?
The proposed change revises the allowed outage time from 48 hours to 7 days when 2 channels of PAM instrumentation are inoperable except for hydrogen monitor instrumentation. The PAM instrumentation is not an initiator of any analyzed event. The PAM instrumentation does not provide any safety function in the mitigation of analyzed events. These instruments provides information to the operators, who are functioning as a backup to the automatic systems designed to mitigate accidents previously evaluated. The proposed change does not significantly affect initiators or mitigation of analyzed events, and therefore does not involve a significant increase in the probability or consequences of an accident previously evaluated. *
2. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?
The proposed change revises the allowed outage time from 48 hours to 7 days when 2 channels of PAM instrumentation are inoperable except for hydrogen monitor instrumentation. The change will not involve a significant change in design or operation of the plant. No hardware is being added to the plant as part of the proposed change. The proposed change will not introduce any new accident initiators. Therefore, the change does not create the possibility of a new or different kind of accident from any accident previously evaluated.
3. Does the change involve a significant reduction in a margin of safety?
The proposed change revises the allowed' outage time from 48 hours to 7 days when 2 channels of PAM instrumentation are inoperable except for hydrogen monitor
instrumentation. The PAM instrumentation does not provide any safety function in the mitigation of analyzed events. Also, the likelihood is remote that an event would occur which would require the PAM instrumentation. Therefore, the proposed change does not involve a significant reduction in a margin of safety.
Palisades Nuclear Plant Page 3 of 9 05/30/99
LESS RESTRICTIVE CHANGE L.4 ATTACHMENT 4 NO SIGNIFICANT HAZARDS CONSIDERATION SPECIFICATION 3.3.7, PAM INSTRUMENTATION The ITS Action B.1 allows continued operation, provided a report is filed with the NRC, if a single inoperable channel for one or more functions cannot be restored to operable status within the specified completion time of Action A.1; CTS Actions 4a and 4b require a plant shutdown in this event (for Functions other than Reactor Vessel Water Level and Containment Area Radiation Monitoring).
The requirement to file a report with the NRC (initiate action in accordance with Specification 5.6.6), is appropriate given the likelihood of plant conditions that would require information provided by this instrumentation. Also, alternative Required Actions are identified before a loss of functional capability condition occurs. This change is consistent with NUREG-1432.
1. Does the change involve a significant increase in the probability or consequences of an accident previously evaluated?
The proposed change replaces the shutdown track statement with a requirement to send a report to the NRC. The report will be issued when one required channel of Post Accident Monitoring (PAM) instrument is inoperable and the Completion Time cannot be met. The PAM instrumentation is not an initiator of any _analyzed event. Post-accident monitoring is still av~lable from the one remaining OPERABLE channel and the pre-planned alternate method of monitoring. Also, the PAM instrumentation does not provide an active safety function. The proposed change does not significantly affect initiators or mitigation of analyzed events, and therefore does not involve a significant increase in the probability or consequences of an accident previously evaluated.
2. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?
The proposed change replaces the shutdown track statement with a requirement to send a report to the NRC. The change will not involve a significant change in design or operation of the plant. No hardware is being added to the plant as part of the proposed change. The proposed change will not introduce any new accident initiators. Therefore, the change does not create the possibility of a new or different kind of accident from any accident previously evaluated . Palisades Nuclear Plant Page 4 of 9 05/30/99
3.
NO SIGNIFICANT HAZARDS CONSIDERATION ATTACHMENT 4 SPECIFICATION 3.3.7, PAM INSTRUMENTATION Does the change involve a significant reduction in a margin of safety? The proposed change replaces the shutdown track statement with a requirement to send a report to the NRC. The report will be issued when one required channel of PAM instrument is inoperable and the Completion Time cannot be met. The remaining required channel or the pre-planned alternate method of monitoring is available to provide the required indication for PAM. The PAM instruments provides no automatic actuation functions. Therefore, the proposed change does not involve a significant reduction in a margin of safety. LESS RESTRICTIVE CHANGE L.5 CTS 3.17.4. 7b requirement that alternate monitoring be initiated within 48 hours with two Reactor Vessel Water Level channels inoperable is deleted. The instrumentation which is used as an alternative to RVWL channels is also requited by CTS 3.17.4 and ITS 3.3.7. No new alternative measurement methods are initiated to implement the requirements of CTS 3.17.4.7b, rather, the Emergency Operating Procedures list alternative indications to be
used in lieu of operable RVWL channels, which provide similar, although not identical, information. The PCS temperature channels and Pressurizer Level instruments provide information to supplement any inoperable RVWL channels. ITS 5.6.6 requires the Accident Monitoring Report to outline the preplanned alternate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels of the Function to Operable status. These controls assure that the appropriate margin of safety is maintained. This change is consistent with NUREG-1432.
1. Does the change involve a significant increase in the probability or consequences of an accident previously evaluated?
The proposed change deletes the requirement that alternate monitoring be initiated within 48 hours with two Reactor Vessel Water Level channels inoperable. The R.eactor Vessel Water Level instrumentation is not an initiator of any analyzed events. The requirement to report the Reactor Vessel Water Level instrumentation preplanned alternate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the instrument channels of the Function to OPERABLE status to the NRC will ensure that alternate monitoring is initiated in a timely manner. Good operating practice and management oversight dictate that alternate monitoring be initiated as soon as possible. The Reactor Vessel Water Level instrumentation does not provide any active safety function in the mitigation of analyzed events. The proposed change does not significantly affect initiators or mitigation of analyzed events, and therefore does not involve a significant increase in the probability or consequences of an accident previously evaluated. Palisades Nuclear Plant Page 5 of 9 05/30/99
2.
NO SIGNIFICANT HAZARDS CONSIDERATION ATTACHMENT 4 SPECIFICATION 3.3.7, PAM INSTRUMENTATION Does the change create the possibility of a new or different kind of accident from any accident previously evaluated? The proposed change deletes the requirement that alternate monitoring be initiated within 48 hours with two Reactor Vessel Water Level channels inoperable. The change will not involve a significant change in design or operation of the plant. No hardware is being added to the plant as part of the proposed change. The proposed change will not introduce any new accident initiators. Therefore, the change does not create the possibility of a new or different kind of accident from any accident previously evaluated.
3. Does the change involve a significant reduction in a margin of safety?
The proposed change deletes the requirement that alternate monitoring be initiated within 48 hours with two Reactor Vessel Water Level channels inoperable. The change will continue to ensure that alternate monitoring is initiated. Also the reactor vessel level monitoring instrumentation does not perform any automatic safety function to mitigate analyzed events. Therefore, the proposed change does not involve a
significant reduction in a margin of safety .
LESS RESTRICTIVE CHANGE L.6 CTS 3.17.4. 7d requirement that the inoperable channels for functions 16 through 21 be restored to Operable status prior to startup from the next refueling is deleted. However, the NRC will be informed of the schedule for restoring the system, as required by ITS 5.6.6. This requirement will ensure that the inoperable channel(s) are restored to Operable status in a timely manner. Also, good operating practice and management oversight dictate that plant systems be restored as soon as possible. This change is consistent with NUREG-1432.
1. Does the change involve a significant increase in the probability or consequences of an accident previously evaluated?
The proposed change deletes the requirement for the inoperable PAM instrumentation to be restored to OPERABLE status prior to startup from the next refueling if any channels are inoperable. The PAM instrumentation is not an initiator of any analyzed events. The requirement to report the PAM instrumentation preplanned alternate method of monitoring, the .cause of the inoperability, and the plans and schedule for restoring the instrument channels of the Function to OPERABLE status to the NRC will ensure that repairs are initiated in a timely manner. Good operating practice and management oversight dictate that repairs be initiated as soon as possible. The PAM instrumentation does not provide any active safety function in the mitigation of
analyzed events. The proposed change does not significantly affect initiators or mitigation of analyzed events, and therefore does not involve a significant increase in the probability or consequences of an accident previously evaluated.
Palisades Nuclear Plant Page 6 of 9 05/30/99
2.
ATTACHMENT 4 NO SIGNIFICANT HAZARDS CONSIDERATION SPECIFICATION 3.3.7, PAM INSTRUMENTATION Does the change create the possibility of a new or different kind of accident from any accident previously evaluated? The proposed change deletes the requirement for the inoperable PAM instrumentation to be restored to OPERABLE status prior to startup from the next refueling if any channels are inoperable. The change will not involve a significant change in design or operation of the plant. No hardware is being added to the plant as part of the proposed change. The proposed change will not introduce any new accident initiators. Therefore, the change does not create the possibility of a new or different kind of accident from any accident previously evaluated.
3. Does the change involve a significant reduction in a margin of safety?
The proposed change deletes the requirement for the inoperable PAM instrumentation to be restored to OPERABLE status prior to startup from the next refueling if any channels are inoperable. The change will continue to ensure that the repairs are performed on the PAM instrumentation as soon as possible. Also the PAM instrumentation does no~ perform any automatic safety function to mitigate analyzed
events. Therefore, the proposed change does not involve a significant reduction in a margin of safety.
LESS RESTRICTIVE CHANGE L. 7 CTS 3.17.4.3 provides action requirements for inoperable position indication of containment isolation valves which allows the associated valves to be locked in the closed position. ISTS Table 3.3.11-1 Note (a) also provides this allowance but also allows other options such as isolation of the associated penetration by any automatic valve which is closed and de-activated, a closed manual valve, a blind flange, or a check valve with flow through the valve secured. Since all of these options provide acceptable isolation of the penetration, ITS 3. 3. 7 will adopt the additional allowances as provided in the ISTS. This change is consistent with NUREG-1432.
1. Does the change involve a significant increase in the probability or consequences of an accident previously evaluated?
The proposed change provides additional acceptable methods to isolate a containment penetration for which the position indication is inoperable. This PAM instrumentation is not an initiator of any analyzed event. The PAM instrumentation does not provide any safety function in the mitigation of analyzed events. These instruments provides information to the operators, who are functioning as a backup to the automatic systems designed to mitigate accidents previously evaluated. Also, the additional methods
identified for containment isolation have been repeatedly recognized as sufficient for**
containment of radiological releases. The proposed change does not significantly affect initiators or mitigation of analyzed events, and therefore does not involve a significant increase in the probability or consequences of an accident previously evaluated. Palisades Nuclear Plant Page 7 of 9 05/30/99
2.
ATTACHMENT 4 NO SIGNIFICANT HAZARDS CONSIDERATION SPECIFICATION 3.3.7, PAM INSTRUMENTATION Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?
The proposed change provides additional acceptable methods to isolate a containment penetration for which the position indication is inoperable. The change will not involve a significant change in design or operation of the plant. No hardware is being added to
-the plant as part of the proposed change. The proposed change will not introduce any new accident initiators. Therefore, the change does not create the possibility of a new or different kind of accident from any accident previously evaluated.
3. Does the change involve a significant reduction in a margin of safety?
The proposed change provides additional acceptable methods to isolate a containment penetration for which the position indication is inoperable. The PAM instrumentation does not provide any safety function in the mitigation of analyzed events. Also, the likelihood is remote that an event would occur which would require the PAM instrumentation. Therefore, the proposed change does not involve a significant reduction in a margin of safety . LESS RESTRICTIVE CHANGE L.8 CTS Table 4.17.4 provides CHANNEL CHECK Surveillance Requirements for post accident monitors with a Frequency of 31 _days. ISTS SR 3. 3 .11.1 also provides CHANNEL CHECK Surveillance Requirements for post accident monitors with a Frequency of 31 days, but requires the CHANNEL CHECK to be performed only if the required instrumentation channel is normally energized. CTS requires the check regardless of the normal energization state. Therefore, this change is less restrictive than CTS. The revised Frequency is acceptable based on industry experience that demonstrates channel failure is rare, and on the use of less formal but more frequent checks of the channels during normal operational use of displays associated with the required channels. This change is consistent with NUREG-1432.
1. Does the change involve a significant increase in the probability or consequences of an accident previously evaluated?
The PAMs are used to support mitigation of the consequences of an accident; however, they are not considered the initiator of any previously analyzed accident. As such, the proposed revision of the Surveillance Frequency of the PAMs does not significantly_ increase the probability of any accident previously evaluated. Since the function of the P AMs continues to be verified, and continues to be required to be OPERABLE, the
change of the Surveillance Frequency will not reduce the capability of required equipment to mitigate the event. Therefore, this change does not involve a significant increase in the consequences of any accident previously evaluated.
Palisades Nuclear Plant Page 8 of 9 05/30/99
2.
NO SIGNIFICANT HAZARDS CONSIDERATION ATTACHMENT 4 SPECIFICATION 3.3.7, PAM INSTRUMENTATION Does the change create the possibility of a new or different kind of accident from any accident previously evaluated? The proposed change does not necessitate a physical alteration of the plant (no new or different type of equipment will be installed) or changes in parameters governing normal plant operation. The proposed change will still ensure proper surveillances are required for the equipment considered in the safety analysis. Thus, this change does not create the possibility of a new or different kind of accident from any accident previously evaluated.
3. Does the change involve a significant reduction in a margin of safety?
The margin of safety for PAMs is based on availability and capability of the instrumentation to provide the required information to the operator. The Frequency is based on unit operating experience that demonstrates channel failure is rare, and on the use of less formal but more frequent checks of channels during normal operational use of the displays associated with the required channels. Therefore, the availability and
capability of the PAMs continues to be assured by the proposed Surveillance Frequency and this change does not involve a significant reduction in a margin of safety .
Palisades Nuclear Plant Page 9 of 9 05/30/99
LESS RESTRICTIVE CHANGE L.1 ATTACHMENT 4 NO SIGNIFICANT HAZARDS CONSIDERATION SPECIFICATION 3.3.8, ALTERNATE SHUTDOWN SYSTEM CTS Table 4.17.5 provides CHANNEL CHECK Surveillance Requirements for the Alternate Shutdown System instrumentation. ISTS SR 3.3.12.1 also provides CHANNEL CHECK Surveillance Requirements for Remote Shutdown System instrumentation Functions, but requires the CHANNEL CHECK to be performed only if the required instrumentation channel is normally energized. CTS requires the check regardless of the normal energization state. However, none of the Alternate Shutdown System instrumentation is normally energized. Therefore, the ISTS bracketed SR is not included in the ITS. Since this change eliminates the quarterly CHANNEL CHECKS, the change is less restrictive than CTS. This change is appropriate and acceptable since the CTS CHANNEL CHECK performance requires placing the system in service, i.e., actuating the transfer to the Alternate Shutdown System instrumentation. This performance also removes the controls from the control room during normal operation. Obviously this is not a preferred condition. The 18 month CHANNEL CALIBRATION provides the necessary confirmation of channel OPERABILITY and is considered sufficient based on industry experience that demonstrates channel failure is rare, and on the use of less formal but more frequent checks of portions of the instrumentation channels during normal operational use of displays associated with the required channels. This change is consistent with NUREG-1432 .
1. Does the change involve a significant increase in the probability or consequences of an accident previously evaluated?
The prqposed change eliminates the quarterly CHANNEL CHECK for the Alternate Shutdown System instrumentation. This instrumentation is used to support mitigation of the consequences of an accident; however, it is not considered the initiator of any previously analyzed accident. As such, the proposed change to eliminate this surveillance for the Alternate Shutdown System instrumentation does not significantly increase the probability of any accident previously evaluated. Since the function of the Alternate Shutdown System instrumentation continues to be verified, and continues to be required to be OPERABLE, the change of the Surveillance Frequency will not reduce the capability of required equipment to mitigate the event. Therefore, the proposed change to eliminate this surveillance for the Alternate Shutdown System instrumentation does not involve a significant increase in the consequences of any accident previously evaluated.
2. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?
The proposed change does not necessitate a physical alteration of the plant (no new or different type of equipment will be installed) or changes in parameters governing normal plant operation. The proposed change will still ensure proper surveillances are required for
the equipment considered in the safety analysis. Thus, this change does not create the possibility of a new or different kind of accident from any accident previously evaluated.
Palisades Nuclear Plant Page 1of3 05/30/99
3.
NO SIGNIFICANT HAZARDS CONSIDERATION SPECIFICATION 3.3.8, ALTERNATE SHUTDOWN SYSTEM Does the change involve a significant reduction in a margin of safety? ATTACHMENT 4 The margin of safety for the Alternate Shutdown System instrumentation is based on availability and capability of the instrumentation to provide the required information to the operator. The acceptability of the elimination of the CHANNEL CHECK Surveillance Requirement is based on unit operating experience that demonstrates channel failure is rare, and on the use of less formal but more frequent checks of portions of the channels during normal operational use of the displays associated with the required channels. Therefore, the availability and capability of the Alternate Shutdown System instrumentation continues to be assured by the remaining Surveillance Requirements and this change does not involve a significant reduction in a margin of safety. LESS RESTRICTIVE CHANGE L.2 The two CTS 3.17.5 Actions which are required when one or more Alternate Shutdown System Function is inoperable have been replaced by the corresponding STS Action. The CTS Actions require: a) Provide equivalent shutdown capability within 7 days, and b) Restore the inoperable channels to OPERABLE status within 60 days. The STS and ITS require: A.1 Restore required Functions to Operable status [within] 30 days. With CTS Actions, equivalent shutdown capability (i.e., functionality) would be restored within 7 days, but restoration to full Operable status would not be required for 60 days. When Action 3.17.5a) was initially added to the CTS, each Function in the table carried an explicit instrument channel designator. Therefore, CTS required a) restoration of the ability to monitor the specified parameters or to provide equivalent control functions quickly, and b) restoration of the designated instrument channel to Operable status within 60 days. ITS and STS each require restoration of Operability within 30 days, but make no requirement to restore any level of functionality before that. The omission of the requirement to restore functionality with 7 days is a Less Restrictive feature of this change. That omission is considered to be acceptable based on the requirement to more quickly restore full operability, the fact that the subject equipment is not required to perform any immediate, or automatic, safety function. There are no particular design features at Palisades which would make the
Completion Times approved for inclusion in STS inappropriate for use in ITS. This change is consistent with NUREG-1432.
Palisades Nuclear Plant Page 2 of3 05/30/99
1.
SPECIFICATION 3.3.8, ALTERNATE SHUTDOWN SYSTEM ATTACHMENT 4 NO SIGNIFICANT HAZARDS CONSIDERATION Does the change involve a significant increase in the probability or consequences of an accident previously evaluated? The proposed change replaces the CTS 3 .17. 5 Action statements requiring restoration of functionality within 7 days and restoration of Operability within 60 days with a single ITS Action statement requirement requiring restoration of Operability within 30 days. The Alternate Shutdown System is not an initiator of any analyzed event. No analyzed accident relies upon functioning of the Alternate Shutdown System. Automatic and manual shutdown capability is still available from the control room. Also, the Alternate Shutdown System does not provide any immediate or automatic safety function. The proposed change does not significantly affect initiators or mitigation of analyzed accidents and therefore does not involve a significant increase in the probability or consequences of an accident previously evaluated.
2. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?
The proposed change replaces the CTS 3.17.5 Action statements requiring restoration of functionality within 7 days and restoration of Operability within 60 days with a single ITS Action statement requirement requiring restoration of Operability within 30 days. The change will not involve a change in design or operation of the plant. No hardware is being added to the plant as part of the proposed change. The proposed change will not introduce any new accident initiators. Therefore, the change does not create the possibility of a new or different kind of accident from any accident previously evaluated.
3. Does the change involve a significant reduction in a margin of safety?
The proposed change replaces the CTS 3.17.5 Action statements requiring restoration of functionality within 7 days and restoration of Operability within 60 days with a single ITS Action statement requirement requiring restoration of Operability within 30 days. The Alternate Shutdown System is not an initiator of any analyzed event. No analyzed accident relies upon functioning of the Alternate Shutdown System. Automatic and manual shutdown capability is still available from the control room. Also, the Alternate Shutdown System does not provide any immediate or automatic safety function. Therefore, the proposed change does not involve a significant reduction in a margin of safety .
I I
I I Palisades Nuclear Plant Page 3 of 3 05/30/99 _J
NO SIGNIFICANT HAZARDS CONSIDERATION SPECIFICATION 3.3.9, NEUTRON FLUX MONITORING CHANNELS LESS RESTRICTIVE CHANGES ATTACHMENT 4 There were no "Less Restrictive" changes associated with this Specification .
Palisades Nuclear Plant Page 1of1 05/30/99
LESS RESTRICTIVE CHANGES (L)
NO SIGNIFICANT HAZARDS CONSIDERATION ATTACHMENT 4 SPECIFICATION 3.3.10, ESRV INSTRUMENTATION There were no "Less Restrictive" changes associated with this Specification . Palisades Nuclear Plant Page 1of1 05/30/99
ATTACHMENT 5 PALISADES NUCLEAR PLANT
SECTION 3e3 - INSTRUMENTATIONS MARKUP OF NUREG-1432 TECHNICAL SPECIFICATIONS AND BASES

RPS 3 .3 INSTRl.MENTATION 3.3.1 Reactor Protective Syste* (RPS) Instru111entationl7'0peft"t1nj"(Alji1og)lj chonne '* ond 09'oc1oted ero ower Mode ( Ml RAI
[3.17.ll LCO 3.3.1 Four RPS trip units.~ usochttd 1nstru111nt ~pus 3.3.1-6 removal channels for each Function in Table 3. -1 shall.be OPERABLE. APPLICABILITY: [MOOE0 a¢' 2.kAccording to Toble 3.3.1-1. >-© RAI 3.3.1-1
-ACT IOHS --------------*----------------------NOTE-------------------------------------
Separate Condition entry is allowed for each ps tifp O?bygx'ss #1119vfH---125' Function. - - ' ~
-~-----------------------~----------------------------------------------------
[3.17.1.2JA.
REQUIRED ACTION COMPLETION TIME I RAI 3.3.1-6
~ ------------NOTE------------
Not opphcoble to High .Stortup Rote, Loss of Lood or ZPM Byposs Removol Functions. ( Insert 1 ) @ Pl1c1 1ffected trip unit 1n trip. ~
~ J (continued) -I
RAI -
3.3.1-1
(Poli sades Nucleo-r Plant)
!/?foG;sts I 3.3-1
SECTION 3.3 INSERT 1 RAJ:
3.~.\-1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME B. One High Startup Rate trip B.1 Restore trip unit and Prior to entering unit or associated associated instrument MODE 2 from instrument channel channel to OPERABLE MODE3 inoperable. status.
c. One Loss of Load trip unit C .1 Restore trip unit and Prior to increasing or associated instrument associated instrument THERMAL POWER to channel inoperable. channel to OPERABLE ~ 17% RTP following status. entry into MODE 3
3.3-1
---------NOTE----------
Not applicable to ZPM Bypass Removal RPS Inst rumen tat i onl?Opej;4't i nj' (Anf og
. 3.
p Function. CONDITION REQUIRED ACTION COMPLETION TIME
------------NOTE-------------
LCO 3~0.4 is not applicable. [3.17.1.3.al 1 hour RAI 3.3.1-1 3.3.1-6
~hours
[3.17.1.2] fuflmmed1otely )
o. One D. l * !Yhou
. [3.17 .1.3]
bypus rMOva inoperable. QR [See DOC M.ll D.21J'.l (continued) RAI
@ I3.3.1-1 < Insert 1)
(Polisodes Nuclear Plant)
@oG/aj's I 3.3-2 !v 1/ 04//01/9~
SECTION 3.3 INSERT 1 G. Required Action and G.1 Be in MODE 3. 6 hours associated Completion Time not met. AND OR G.2.1 Verify no more than 6 hours one full-length Control room ambient air control rod is capable temperature > 90°F. of being withdrawn.
QR G.2.2 Verify PCS boron 6 hours concentration is at REFUELING BORON
CONCENTRATION

3.3-2
, RPS lnstrumentat i onf ONrablng 'tAna i)?g)) t
3. 3 .1
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME D.2.2~ Restore bypass removal channel a fected trip unit
[48] h\ to PERABLE status. OR
.2.2.2Place af ted trip units in t
48 ours J E. One or more unctions ------------NOTE-------------
with two aut atic LCO 3.0.4 is no applicable. bypass removal channels inopera le. E.l Disable bypa channels.
""\ Place one affected trip unit in bypass and place the other in trip for each affected trip unction.
one bypass [48] hours channel nd the associat trip unit to OPERABL status for each af cted trip Function. F. Required F.l 6 hours associated Time not met .
CEOG STS 3.3-3 Rev 1, 04/07/95
RPS Instr*umentat i orE-~era~g ~na ~p
3. .
~.11]
SURVEILLANCE REQUIREMENTS
--------------------------~----------NOTE-------------------------------------
Refer to Table 3.3.1-1 to determine which SR shall be performed for each~ Function. .
~
SURVEILLANCE FREQUENCY [14,11.1] SR 3.3.1.1 12 hours (continued) I: CEOG STS 3.3-4 Rev 1, 04/07/95
SECTION 3.3 INSERT 1 SURVEILLANCE FREQUENCY SR 3.3.1.2 Verify control room temperature is~ 90'F. 12 hours

3.3-4
GD, RPS InstrumentationE:0pera~g ~nai-e~f ~ 3.3 . SURVEILLANCE REQUIREMENTS continued SURVEILLANCE FREQUENCY [T4.11-1] SR Once within 7 days prior to each reactor * (2) startup SR~ rfonn a CH~Lr~~NCTIO L TEST on ea au matic bypa ~val fu tion. Once wi~in 92 days ior to each ".' tor art up SR 3.3.1.8 ~lsQ} months SR 3.~~~~~- Veri~PS RES~NSE TIME ~-within li~s. [1 month~ (@
"' on a JAGGElttD TEST B
CEOG STS 3.3-5 Rev 1, 04/07/95
~
RPS Instrumentatior(-o~tih§..JAn~:i:~
.--* _ _ _ _ __!..._ _ _ _ _ _ _ _T_.C_l_*-:3-.-3.-1--1-(P919--1-o-f..-2->-------------~
IHctor Protec:tiY9 Syst* lrwt~1tion ==\ SUtVEILWCE 'I FUICTICll IECIUllEMElfTS ALLCWAILE VALUE
~~~~~~~~~~~~~__...;---~~~
1. V1rf.Cle High Power Trip SI 3.3.1.1 ! [10JI HP lbaw current II 3.3.1.2
TllEIML ~ but not
'* SI 3.3. 1 .3 < [30] I ITP '*--
SI 3.3.1.4 > [10711 If! '
3.3.1.5
-.. ,_ SI 3.3. 1.II II 3.3.1.9
2. Powr lite of Chqe - Ntg11<*> II '*'1,3. 1. 1 s [2.6] -
II 3;~1.6
3. RelCtor Cool.,t *lllow - LOi.
II sa SI
....~
3.3. 1. 3.3.1.1 ! [95]1
\ 0 II 3.3.1.4 ' I ~
II 3.3.1.7 II 3.3.1.I @
4. Pressurizer PreuuA -N1.h*\,,
\
II II 3.3.1.9 3.:S.1.1 s~ p1f1 I
3.3. 1 .4 II 3.3.1.1
~ .I \,_ ' ', *,
3.3.1.9 paf1~
I [II 3.3.1.1J s [4.0J I
5. tont1f..-'lt p,..._. - Nigh
~ II 3.3. 1.4 Ill 3.3. 1 **
SI 3.3.1.9
> I I
II 3.3.1.1 t [615J p1f1 II 3.3.1.4
3.J.1.7 II 3.3.1.I

3.3.1.9 3.3.1.1 t [24.711 3.3. 1.4 II .3.t.I
* .t.9 t [24.7]1 II (cont i l'Uld) lllall be IUt...tlcall (b) lliwt*Md w..i TllDML .a.a fl < [11*4
lypas lllall be IUt...tf ly rmll'<l'8d llhen @
11*4JI ITP. Durtne teltf"' to LCO 3.4. 17, ICS Loot:ll - taiwm1Miit__,.l11111 51 m. 1ype11111au be 1Ut ttcally . . - - "'"' T (
- CEOG STS 3.3-6 Rev 1, 04/07/95
SECTION 3.3
INSERT 1 Table 3.3.1-1 (page 1 of 2)
Reactor Protective System Instrumentation f?Al: S'.~.1-1 3.~.\-~ APPLICABLE SURVEILLANCE FUNCTION MODES REQUIREMENTS ALLOWABLE VALUE
l. 1. Variable High Power Trip 1,2,3<al , 4cal , 5cal SR 3.3.1.1 .s; 15% RTP above SR 3.3.1.2 current THERMAL SR 3. 3 .1. 3 POWER with a SR 3 .3 .1.4 minimum of ~ 30%
SR 3.3.1.5 RTP and a maximum SR 3 .3 .1.6 of ~ 106.5% RTP SR 3.3.1.8
2. High Startu~ Rate Tri p<bl 1,2 SR 3 .3.1. l NA SR 3. 3 .1. 7 SR 3.3.1.8

3. Low Primary Coolant Syste~ Flow Trip~l 1, 2 *3(a) *4Cai *5(a) SR 3 .3 .1.1 ~ 95%
SR 3. 3.1. 5 SR 3.3.1.8
4. Low Steam Generator A Level Trip 1, 2 3(a) 4(a) 5(a) 1 1 1 SR 3. 3 .1.1 ~ 25.9% narrow
*SR 3.3.1.5 range SR 3.3.1.8
5. Low Steam Generator B Level Trip 1,2,3<al ,4<al ,5<al SR 3. 3 .1.1 ~ 25.9% narrow SR 3.3.1.5 range SR 3.3.1.8

6. Low Steam Generator A Pressure Trip<c> 1, 2 3(a) 4(a) 5(a) 1 1 1 SR 3. 3 .1.1 ~ 500 psia SR 3.3.1.5 SR 3.3.1.8

7. Low Steam Generator B Pressure Tri p<c> 1, 2 3(a) 4(a) *5(a) 1 1 SR 3.3.1.1 ~ 500 psia SR 3.3.1.5 SR 3.3.1.8

8. High Pressurizer Pressure Trip 1, 2 3(a) 4(a) 5(a) 1 1 1 SR 3.3.1.1 ~ 2255 psia I SR 3.3.1.5 SR 3.3.1.8 (a) With more than one full-length control rod capable of being withdrawn and PCS boron concentration less than REFUELING BORON CONCENTRATION .
(b) Trip may be bypassed when Wide Range Power is < lE-4% RTP or when THERMAL POWER is
> 13% RTP.
(c) Trips may be bypassed when Wide Range Power is < lE-4% RTP. Bypass shall be automatically removed when Wide Range Power is ~ lE-4% RTP. 3.3-6
RPS Instrumentat i on~Op'e9t i n3t---fAna}ggi) 3 .3 .1
~/ Table 3.3.1*1 Cpege 2 of 2> *11 ..ctor Protective Syst* lnstrUllef!Ution //==:...__~~~~~~~~~~~~~~~-==========:::::-
UVEILWCE FUllCTIOll REQUIREMEllTS ALLCWAiLE VALUE
a. Axi1l Powr Distribution - Migh(d) SR 3.3.1.1 Figure 3.3. 1*3 SI 3.3.1.Z sa 3.3.1.3
., sa 3.3.1.4 sa 3.3.1.5 \ SI 3.3.1.7 "* SI Sit 3.3.1.1 3.3.1.9
91. Therml Mlirgin/LI* Prn1ur1 (TIC/LP)Cb)
'"* \ '\
SI 3.3.1.1 SI 3.3.1.Z
\ SI 3.3.1.3 \SI 3.3.1.4 SI 3.3.1.5 ~ 3.3.1.7
[SI 3.3.1.Sl
~:, ,._ -....~ * - o;tt....,<*> : ~~-1-:
SI SI 3.3.1. 3.3.1.1 s [135] psid 10. oir*~><d>
'LO.a of Lad (turbine 1tap wl~control SI SI 3.3. 1 .9 3.3. 1 .6 3.3.1.7 !: [IOO] psi; SI 3.3.1.I \ ~tn 1E*4]1. lypeu wll be Mlt. .tf lly ~ Wien \
(b) Trips *Y be THEllML POWEi is bela.. 51 RTP. TllEIMI. POWEi la [1E-4JI RTJI. Dlrfl'll tntl"I wl l be Mlt. .ttcelly t to LCO 3.4.17, trips ..., a.lien TllEIML POYEI fa t 51 ITP.
~ @ \
(d) Trip _., be ~ Wltn TllEIML 1!CM1 11 c [15]1 ITP. TH£1ML PQWEl is !: [15]1 ITP. lypna Wll be Mlt. .ttcally r-w.cl Wien CEOG STS 3.3-7 Rev 1, 04/07/95
SECTION 3.3
INSERT 1 Table 3.3.1-1 (page 2 of 2)
Reactor.Protective System Instrumentation APPLICABLE SURVEILLANCE EU. 3.~.1-1
~. '3. \-lP FUNCTION MODES REQUIREMENTS ALLOWABLE VALUE
9. Thermal Margin/
Low Pressure Trip~> 1, 2, 3(a), 4(a) , 5(a) SR 3.3.1.1 Table 3.3.1-2 SR 3.3.1.2 SR 3.3.1.3 SR 3.3.1.4 SR 3.3.1.5 SR 3 .3 .1.6 SR 3.3.1.8
10. Loss of Load Trip 1(d) SR 3.3.1. 7 NA SR 3.3.1.8

11. Containment High Pressure Trip 1, 2, 3Cal, 4(a), 5Ca) SR 3.3.1.5 ~ 3.70 psig SR 3.3.1.8

12. Zero Power Mode Bypass Automatic
Removal 1,*2,3<al ,4Cal ,s<al SR 3.3.1.8 NA (a) - With more than one full-length control rod capable of being withdrawn and PCS boron concentration less than REFUELING BORON CONCENTRATION.
(c) Trips may be bypassed when Wide Range Power is < lE-4% RTP. Bypass shall be automatically removed when Wide Range Power is <!: lE-4% RTP. (d) When THERMAL POWER is <!: 17% RTP . I 3.3-7
RPS Instrumentat i onEap.e.ra'NJ1g 'f.Ana 1-Qgp 3.3 . QoNI =A1 x QR1
TR~= 2892x(A1) x (QR1) + 17.
0.1 0.2 0.3 0.4 0.6 ASI
\ Figure 3.3.1*1(page1 o )
Thermal Margin/Low Pressure Trip Setpoint: ASI vs A1
CEOG STS 3.3-8 Rev 1, 04/07/95
(
SECTION 3.3
INSERT 1 Table 3.3.1-2 {page 1 of 1)
Thermal Margin/Low Pressure Trip Function Allowable Value The Allowable Value for the Thermal Margin/Low Pressure Trip, Ptrip' is the higher of two values, Pmin and Pvar' both in psia: Pmin = 1750 . Pvar = 2012(QA){QR1) + 17 .O(T;n) - 9493 Where: QA = - 0.720(ASI) + 1.028; when - 0.628 ~ ASI < - 0.100 QA = - 0.333(ASI) + 1.067; when - 0.100 ~ ASI < + 0.200 QA = + 0.375(ASI) + 0.925; when + 0.200 ~ ASI ~ + 0.565 ASI = Measured ASI when Q ~ 0.0625 ASI = 0.0 when Q < 0.0625 QR1 = 0.412(Q) + 0.588; when Q ~ 1.0 QR1 = Q; when Q > 1.0 Q = THERMAL POWER/RATED THERMAL POWER T;n = Maximum primary coolant inlet temperature, in °F ASI, T1n, and Q are the existing values as measured by the assoctated instrument channel
3.3-8
RPS Instrumentation~OO"eratlftg (A~logp 3.3 . (1.2,1.2 FOR IUUSTRATION ONLY. NOT USE FOR OPERATION. 1.00 1---+-1'~-+---+--+---i--+--~-+---+---+--=£..J-t 0.90 1--+--~--+--+---i--+----+/-:;;.~~---+---i---'--+-........-t 0.5 0.8 0.7 0.8 1.1 1.2 RATED THERMAL POWER Figure 3.3.1-2 (page 1 of 1 n/Low Pressure Trip Setpoint Fraction of RTP vs QR 1 CEOG STS 3.3-9 Rev 1, 04/07/95
,. RPS InstrumentationE;:~at"kl.g_ (~o~j)
.3 .1 (0,1.20) 1.20 UNACCEPTAB OPERATION 1.10 REGION Il a:: I (0.2,1.00) ~
(-0.2,1.00) 1.00 0 a.
...J <C I
I I I a:: w ACCEPTABLE
J: OPERATION I- REGION Cl 0.70 w
I-
~ I
u. 0.60 \
0 \ z 0.50 \\ 0
~ \
(.)
~ 0.40 (-0.I, 0.40) * (0.6, 0.40) \
u. l I 0.30
- 0.4
0.2 0.0 0.2 0.4 O.& o.a PERIPHERAL AXIAL SHAPE IND ure 3.3.1-3 (page 1 of 1)
Peripheral Axi hape Index, Y1 vs Fraction of RTP
CEOG STS 3.3-10 Rev 1, 04/07/95
(
.,,..-..-. ---- RPS Instrumentation-Shutdown (Ana-log /
3.3.2
J. 3 iLco 3.3.2 INSTRUMENTATION
/ / / /
Instrumenta ti on-Shutdow_~/(Ana log) our Power Rate of Change-High RPS trip u ts and associated instrument channels shall be OPERABLE, wi an Allowable Value of s [2.6] dp~. i
*i!
f MODES 3, 4, and S, with any reacto trip circuit breakers (RTCBs) closed and any contr element assembly capable of being withdrawn.
---------------------------- TE----------------------------
Trip may be bypassed when ERMAL POWER is < [lE-4]% RTP. Bypass shall be automati lly reraoved when THERMAL POWER is
~ [lE-4]1 RTP.
I l
\
ACTIONS REQUIRED ACTION COMPLETION TIME
A. One Power Rate of Change-High trip uni or associated instrument channel Place affected trip unit in bypass or trip.
1 hour inoperable. A.2.1 Restore channel to OPERABLE status. A.2.2 Place affected trip unit in trip. I
~----/__ _____ / {continued) l
- CEOG STS 3.3-11
-\\.,o ~ /.,._
Rev 1, 04/07/95 3,1-1~
r; i~:ll'fl~<:.J
3.3 INSTRUMENTATION RPS Logic and Trip Initiation
~
(~Jaji) 3.3.~
*1\(J- . (;,
f ~~e (RPS) Logic and Trip Initiation ({:ffiij:l 0§)) APPLICABILITY: MODES l and 2, MODES 3 4, and 5, e &taen se
hdra tw\ ~ 'm~lJ owe-. fil\.\.- ~i'lo\ COM'T11t>\. tcoD CA.PAl!U: o f ~ e I ""-~ w i r-a..W VI a."\.J
?riW'\'4r':f.. C.0.01'1.i.. t 5y~+e 1'4 [Pl:S) \ao ro>'- CC°i'lt: e 111.~ ro.+ 1'"""° ACTIONS 1-4~ lCFIAEL-IN6 Bo~AJ C.CAJU.IJ"ft.loT1tH),
CONDITION REQUIRED ACTION COMPLETION TIME A. l Restore channel to 4_8 hours OPERABLE status
One Matrix logic channel inoperable.
(continued)
CEOG STS 3.3-15 Rev 1, 04/07/95

C. ,-5 ACTIONS continued CONDITION RPS Logic and Trip REQUIRED ACTION Initiation~
-3.~
COMPLETION TIME 1 hour (3. Ii. I. 5"] On~~el o~ Trip Bs'A Initiation ~
~~rable D C.l
Re.ca+o('e Of'~A-f31_t:
c,..i...""-1111e t+a. fi..is. I +o D. Two channels of...,.., ....... D.1 !mediately
~ Initiat-ion Logic affecting the same.
trip leg inoperable. (continued) CEOG STS 3.3-16 Rev 1, 04/07/95
2."T5 ACTIONS continued RPS Logic and Trip Initiation tl~l1il1D 3.3.U@
CONDITION REQUIRED ACTION COMPLETION TIME E. E. l Be in MODE 3. 6 hours MH2
~ 6 hours V~vi r'1 '11o ~ore -+V.4.\.1 o"1 e. !='t.JLL- Li=t.161'1-'
COAJTt'OC... R.C>P *S Cti.P6'.gL~ oF- ED 8ErtN£=, Wln40£A~ ve.v 1+'i f't:~ bo r-ci"'
~1'Y1c...2"'-.fva.h<l,,.. I :5 a.t ReF<.J E"Li t-.l C. 5 o e.a ~
C.Dl..J CEJJTt ttTi O u .
SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY Perform a CHANNEL FUNCTIONAL TEST on each ~92! days CD RPS Logic channel and
,Mo..friit ~"- 1'-ip channel. ~1'-f-1'4..f;"""- L~*c.
8 [n.. 11. I (1')") SR 3.3g:2 Perform a CHANNEL FUNCTIONAL TEST on each Once within RPS Manual Trip channel. 7 days prior to each reactor startup
~j Pe orm a CHANN FUNCTIONAL T ,
inclu g separate rification o he G undervo e and shun rips, on eac RTCB *
CEOG STS 3.3-17 Rev 1, 04/07/95
~
3.3 INSTRUMENTATION ESF~ Instrumentation (AnaT°'ib 3.3.~ Enqinee(ed Safety Features~ctiatfbfl Sy'tteili) (ESF~) Instrumentation 3.Ll'r-~ f: (;_I
) .. ((<<7, ~) )~ _fAfta l'o9 JJ~'D }., r(b'ts+~\LS)
LCD 3.3.~ Four ESF@frij;uitt:U and associated instrument-~=""'-"'~~
<t'."eiliCri"in channels for each Function in Table 3.3 1---G OPERABLE. ,~
APPLICABILITY: 'MClD 1:"2., ~ 3. ACTIONS
-------------------------------------NOTE-------------------------------------
Separate Condition entry is allowed for each IESPAS° tiftj ot'§Vpa$,.S:"!e111Qlfi'J) ~ Function. CONDITION REQUIRED ACTION COMPLETION TIME
One Containment Spray
( tuation Si~al
) trip un*t or as so ated inst inoper le.
nt
~ ~1t1n~
Place affected trip
~ \ \ ,,,_.._8'*
i.-:r-'-..V I
\
I 6'1STA
,... i ~
i ~. 1 1.1...1._; \0 One or more Functions
~l Place affected~
IJ. with one ES~ J.17.:.?..~* ~or
)...-- , -,
assoei:t ~ trip. in (Jijljla'S.$ '01'
- * " instrument channel ,~fuijp~D inopera: l._----------1--------,
B.2.1'-. Restore channel to ,{48] ho~ "\ o,E~LE stat~~:
~ ~] - - }Jot-e. - - - - -
ic+- ~ r I lu..b le +o KAS: na *- '* ""*, \
- - - -~ '""
2.~""Pl ace affected trip
~it in tri)>.
48ho~ j (continued) CEOG STS 3.3-18 Rev 1, 04/07/95
j]) (cJ/ f') "-1 t!S) ES~ Instrumentation (:An-atoi]~ 3.3.~~ ACTIONS continued CONDITION REQUIRED ACTION COMPLETION TIME
-- -- N071:~- - - - -
Not- c..prltc.b-ble. +o Rb..S. !HQ D. e or more Fu~ctions 0.1 Disable bypass ~- 1 hour '* ( wi one automatic channel. \
\ bypa removal channel ',\
\"'---....,_i noper le.
\\ - - - - , \, .,",\ 0.2.1 \
QB
""-\,, 'AN,D Place affe~ted trip units in byp~ss or trip. *. ,
1 ho~ I D. 2. 2. l\~estore bypass removal channel and
*,*,,_ [48] hours l "'~""
aff!!cted trip units \
\\ ,t~~.:s~::*:::~
1 \ L u'n~ts in tri"P-.. 48 hours (continued)
CEOG STS 3.3-19 Rev 1, 04/07/95
SECTION 3.3
CONDITION INSERT 1 REQUIRED ACTION COMPLETION TIME C

One RAS bistable or associated C .1 Bypass affected 8 hours instrument channel inoperable. Bistable.
Mill C.2 Restore bistable and associated instrument 7 days channel to OPERABLE status .
3.3-19

ACTIONS continued CONDITION ESF~
REQUIRED ACTION Instrumentation ~
~
COMPLETION TIME r---. e or more Fun~ns ------------NOTE---~--------- wi two automatic LCO 3.0.4 is not apprtcable.
~\ bypa removal ------------------------~~~-- \ channe s inoperable. ~ ................ \ Disable bypass l hour channels.
QB E.2.1 Plac one affected trip u t in bypass and plac the other in trip fo each
\
affected ESF
'* Function.
E.2.2 Res re one bypass chann and the associa trip unit to OPERAS~ status for each af ted trip Function.
~Required Action and associated Completion Be in MOOE 3. 6 hours Time not met k Fv~c...+i.011s ~
Be in MOOE 4. ~hours I '2 1'3,"i oe1 [ik.Jsi?eT ~
-----------------------------------------------------==-==--=-
CEOG STS 3.3-20 Rev 1, 04/07/95
SECTION 3.3
CONDITION INSERT 1 REQUIRED ACTION COMPLETION T'E .
E. Required Action and E.l Be in MODE 3. 6 hours associated Completion Time not _met for Functions 5 or 6. AND
E.2 Be in MODE 5 . 36 hours 3.3-20

SURVEILLANCE REQUIREMENTS/' @
SURVEILLANCE FREQUENCY (:D SR 3 .3 .'1.1 Perfonn a CHANNEL CHECK (prjijcNSFA~ 12 hours @ ( rts.tN,me'hCcharm!J). SR 3.3.tr Perf onn il CHANNEL FUN~ONAL TEST ~ (SN;$ inffi"umee.t d\an .
~9@days CD@
SR ~ 4. 3 Pkform a CHANt4EL FUNCTIO TEST on ea~ \ Once."-..tithin~
~ autO'ib&tic bypa~emoval fun 'on. 92 day~rior o eac~ ~cto s rtup SR
CEOG STS 3.3-21 Rev 1, 04/07/95

r.bl* 3.3.4-1 cpa En;irwered safety F*tur* Actumti ESF~ Instrumentation *~ .., _
1 of 2> SVSt* lrwt~tltion \
*~-~ -~
QES
-VEILLAllC:
llEQUI IEMEllTS
'\ ALLINAllLE VALl.E '\\
1. S.fety Injection Actu.tion Sf11Ml CSIASJ
** CClntli,_t PrlUUl'I- Nlgft 1,2,3 st 3.3.4. 1 s [19.()] psi*
Sl 'J.3.4.2 Sl 3~3.4.4 Sl 3.3~~-5 1,2,3
SI 3.3.4. t--.
3.3.4.2 Sl 3.3.4.3 sa 3.3.4.4 SI 3.3.4.5
2. cont1i,_,t Sprrf Actumtfon Sf11Ml Cb>
** CClnt*i..-it Prasure - Nigh 1,2,3 SI 3.3.4.1 ~ [19.0] pai*
Sl 3.3.4.2 SI 3.3.4.4 Sl 3.3.4.5
3. Cont*f,_,t lsol*tfon *'-tfon Sf11Nl
. \ ..
CClnt*i-' ll'reuurt-lltlh 1,2,3
.. 3.3.4.1 s [19.()] psi* .... 3.3.4.2 \, * 'J.3.4.4 Sl 3'~.4.5
[ . ,b. contaf~ bdl*tfon -111111 J 1,2,3 3.3.4.1 3.3.4.2. SI 3.3.4.4 \
*"-,, SI 3.3.4.5 \ \\\.
4. Mllfn s~ Jsol*tfon Sfpl
** st-*Gene,..t:or PrtslUl"e - Lw<c> ~, 2 <d> 1 3Cd> Sl 3.3.4.1 !: 'ff!95l 'r-11 sa 3.3.4.2 *, \~ ',*, "'\'\ SI 3.3.4.3 SI 3.3.4.4 sa 3.3.4.5 \ \ \ \
5. lecfrcul*tfon Actumtf i' \ \
\
i ** lefuelf,.. W.ter Tn l-Lw 1,2,3 \ [SI 3.3.4.1] SI 3.3.4.2 s
~ 24 'ncll* 8ftd I \ 30l Inell*
sa 3.3.4.4 8bolle tri bottm SI 3.3.4.5
\ <*> [1IOO] psi*. Th*
CEOG STS 3.3-22 Rev 1. 04/07/95
SECTION 3.3
INSERT 1 Table 3.3.3-1 (page 1 of 2)
Engineered Safet~ Features Instrumentation APPLICABLE SURVEILLANCE eA't:. 3~.:i-1 ALLOWABLE FUNCTION MODES REQUIREMENTS VALUE
1. Safety Injection Signal (SIS)

a. Pressurizer Low Pressure 1,2,3 SR 3.3.3.1 2 1593 psi a SR 3.3.3.2 SR 3.3.3.3

2. Steam Generator Low Pressure Signal (SGLP)

a. Steam Generator A Low l,2(a) ,3(a) SR 3.3.3.1 2 500 psi a Pressure SR 3.3.3.2 SR 3.3.3.3

b. Steam Generator B Low l , 2(a) , 3(a) SR 3.3.3.1 2 500 psia Pressure SR 3.3.3.2 SR 3.3.3.3
3. Recirculation Actuation Signal
* (RAS)
a. . SIRWT Low Leve 1 1,2,3 SR 3.3.3.3 2 21 inches and
~ 27 inches above tank bottom
4. Auxiliary Feedwater Actuation Signal (AFAS)

a. Stearn Generator A Low Level 1,2,3 SR 3.3.3.1 2 25.9%
SR 3.3.3.2 narrow range SR 3.3.3.3
b. Steam Generator B Low Level 1,2,3 SR 3.3.3.1 2 25.9%
SR 3.3.3.2 narrow range SR 3.3.3.3 (a) Not required to be OPERABLE when all Main Steam Isolation Valves (MSIVs) are closed and deactivated, and all Main Feedwater Regulating Valves (MFRVs) and MFRV bypass valves are either closed and deactivated, or isolated by closed manual valves . 3.3-22
SECTION 3.3
INSERT 1 (continued)
Table 3.3.3-1 (page 2 of 2) Engineered Safety Features Instrumentation APPLICABLE SURVEILLANCE ALLOWABLE FUNCTION MODES REQUIREMENTS VALUE
5. Containment High Pressure (CHP)

a. Containment High Pressure 1,2,3,4 SR 3.3.3.2 ~ 3.7 psig
- Left Train SR 3.3.3.3 and ~ 4.34 psig
b. Containment High Pressure 1,2,3,4 SR 3.3.3.2 ~ 3.7 psig
- Right Train SR 3.3.3.3 and ~ 4.34 psig
6. Containment High Radiation Signal (CHR)

a. Containment High Radiation 1,2,3,4 SR 3.3.3.1 ~ 20 R/hour SR 3.3.3.2 SR 3.3.3.3

7. Automatic Bypass Removals

a. Pressurizer Low- Pressure 1,2,3 -SR :3.3.3.3 > 1700 psi a Bypass l,2Ca) ,3Ca) SR 3.3.3.3 > 565 psi a

b. Steam Generator A Low Pressure Bypass l ,2(a) ,3Ca) SR 3.3.3.3 >565 psi a

c. Steam Generator B Low Pressure Bypass (a) Not required to be OPERABLE when all Main Steam Isolation Valves (MSIVs) are closed and deactivated, and all Main Feedwater Regulating Valves (MFRVs) and MFRV bypass valves are either closed and deactivated, or isolated by closed manual valves .
3.3-22
Table 3.3.4*1 (pmge 2 of 2)
EngiMered S*ftty FHtures Actuation Syst* ES~ Instrumentation <(Asa~ lnat~ntion , 3.3;~--2.J
~
SURVE I Ll.AJICE ALLOWAllLE F\JllCT I Oii aes lECILI I IEMEllTS VALUE
6. Auxi l i*ry F.-..ter Actuation Si p l CAF.U)
** S t - Gener*tor A L..,.l.- Low 1,2,3 .SI 3.3.4. 1 !! [45.7] ll\
3.3.4.2 8'.3.3.4.4

3..3.4.5
b. S t - Gener*tor I L..,.l - Low 1,2,3
3.3;~
SI 3.3.4.
!! [45. Tl ll\ '-. *. SI 3.3.4.4 ~
II 3.3.4.5
~ I II
c. S t - Gftrator Pressure 't;2,3 SI 3.3.4.1 5~3] psid iI Difference :...Nigh SI 3.3.4.2 J
CA > B> or ct-..~ A> ., SI 3.3.4.4
'* SI 3.3.4.5
CEOG STS 3.3-23 Rev 1, 04/07/95

c.. r :s 3.3 INSTRUMENTATION
~u.w~~S::afety Features <[c"tU4,tiCfu:Syste~ (ES~) Logic and Manual ~
r3.11.z...i i)'.17* 3j Arla.lo -:C-.... \ *h c._..+, 'o " Two ESF@ Manual and two ES~ Actuation Logic channels shall be OPERABLE for each ESF~ Function specified in J Table 3.3 1. ('a-Hd t:i.JS"!JC: ia fee{"') 4- 6'(,0*HJ ~~<' : / r--<§) APPLICABILITY: According to Table 3.3. 1. C.h.#1-m.e/.f _ _ ) ACTIONS
-------------------------------------NOTE-------------------------------------
Separate Condition entry is allowed for each Function. CONDITION REQUIRED ACTION COMPLETION TIME
[J.1?. '2._, 1]
\).11.J. 0 A.
IV\ 1'+-1 *a.ri~ .. ) A.I Restore channel to OPERABLE status. One o c- Mot"~ F"'v..t.i'~~s w\~ 48 hours f-w" .¥10..... w:.I J:..ti*Hc..+-~, Eyp :.s i(e...,.ov..1> ar- Ac:f'iutL.t;O"K Le~< f>,w~ss R_~Ml>\JA...() c htl"-"I! 1-.s l!UfenJ..\a.. +o i FIA. .... tf:p~S I)?. .11 &I"~' l(A,~
'5.';.'3* \
I
~
[!. 11. z .sJ B. Required Action and B.1 Be in HOOE 3. 6 hours @ associated Completion Time of Condition A Arm
~s-G) not met B.2 Be in HOOE 4. R~!. -6'( func.,+io~s 1)2.) J or- Y. '3:S:3- I \: R~ channe~
OPE LE status. (continued)
CEOG STS 3.3-24 Rev 1, 04/07/95
c'.'.Jr1t 0 r ~crit r1.1..-'l ~;O'\{$ 1V1~ 4-wo t'\o. If. VJ..\ :t..._; +:IL+;""" I or ,Ad1U1..-t:~ lo~~L - R~:I l~"--'els \v..cf~r,,.'41..o(l.. .(.., r- - ~l'\1.h.:-t-1;,,.;) J.!.l* 1
F1.1.~L~;~.$. 5 O'r lD

continued CONDITION ESF~Log1c and Manual~(~
REQUIRED ACTION
*~
COMPLETION TIHE [J, 17,J, s-J ~Required Action and ~ Be in MODE 3. 6 hours associated Completion Time of Condition~ AHl2 not met! W ,~ Be in MODE 5. 36 hours {])-- ~ hiiAc..+to~ 5 or- ~ SURVEILLANCE REQUIREMENTS RAJ: 3,;,y-2 I SURVEILLANCE FREQUENCY
-+
SR f>AJ:
~~;4*(p
2.-... e ays assoc ate w p ant eguipment
- --~hat canno_t be operit~d during)~
operation ar-.(! only req~tred to be tes'ted during"-e~ch HOOE S---e~try
\ exceea'l(lg 24 hour~ unless te~~d ~uring the prevjou~
I
~
Perform a CHANNEL FUNCTIONAL TEST on each ~92.@days ~ A AS logic channel. <ffiAZ
~?.'/-~
SR 3.3~~ lSFAS Perform a CH~_NNEL FUNCTIONAL TEST dlfiSiiSIV ~lSJ months~ MaoiLil, Ic:i.p cba1m*). * ©
?er~on"" ~lA"-t-\-;~j{_\ +1's..\:- D~ 'E:-b.CI-\- ~1S q-z. t!c..vs .ACiUATlol-..l CUAt-Jt-l"E:L lk:>'f:.MA.L 1>.tJt::> STA>-ll)BY "PoWF:J::: FuJJC.ltONS.
CEOG STS 3.3-25 Rev 1. 04/07/95
,*~. ~ J.1<11-ru .. J,. l ES~ Logic and Manual 1~ ~ _ . 3.3@--(f) r I
T.tile 3.3.5*1 (peee 1 of 1) E119frwered safety Fntur* Actumtion Syst* ActU1tion Logic: rd ~l Ch-.l Applic:abi
\ \ \ \ FUICTICll ' APPLI MODES I
1. Safety lnjectf 1,2,3, [4]
Contal-t Spray a.tlon Signal 1,2,3, (4J I
3. ontal-t Isolation 1,2,3,4 :~~
\ (§_)
4. Kai St... Isolation SI 1,2,3,4

5. lec:ircu tfon Actuation Sfgna 1,2,3,4

6. Auitll fary ter Ac:tumtlon Si 1,2,3
CEOG STS 3.3-26 Rev 1. 04/07/95
SECTION 3.3
INSERT 1 Table 3.3.4-1 {page 1 of 1)
R'AL 3.3.'1-Z 3, !.~* '! Engineered Safety Features Actuation Logic and Manual Initiation APPLICABLE FUNCTION MODES
1. Safety Injection Signal (SIS) (al 1,2,3

2. Steam Generator Low Pressure Signal (SGLP) (bl( cl

3. Recirculation Actuation Signal (RAS) 1,2,3

4. Auxiliary Feedwater Actuation Signal 1,2,3 (AFAS)

5. Containment High Pressure Signal 1,2,3,4 (CHP) (el
6. Containment High Radiation Signal (CHR)
(a) 1,2,3,4 SIS actuation by Pressurizer Low Pressure may be manually bypassed when pressurizer pressure is ~ 1700 psia. The bypass shall be automatically removed whenever pressurizer pressure is > 1700 psia. (b) SGLP actuation may be manually bypassed when SG pressure is ~ 565 psia. The bypass shall be automatically removed whenever steam generator pressure is
> 565 psia.
(c) Manual Initiation may be achieved by individual component controls. (d) Not required to be OPERABLE when all Main Steam Isolation Valves (MSIVs) are closed and deactivated, and all Main Feedwater Regulating Valves (MFRVs) and MFRV bypass valves are either closed and deactivated, or isolated by closed manual valves. (e) Manual Initiation channels not required
3.3-26

3.3 INSTRUMENTATION 3.3i Diesel Generator (DG) L o Start
~ ~ ~tee*)
LCO 3.3..- ~_cfiai!n~"is of Loss of Voltage Function and channels of Degraded Voltage Function auto-initiation instrumentation DG shall be OPERABLE.
-'t.lt:> -A.S4'0'-IAT'E-P L.OblC.. c,+4AtJ~~ ~or: E:'lli.C.~
I APPLICABILITY: CMoDES I>--2. )'b an<N) r~ When ~soci ated DG is reg5ed to be OPERABLfulSNC0-...3. 8iD 0 At>--SDurc'h-..5 hutswn . ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME
A. One or more Functions with on channel per DG inope le.
Place channel in bypass or trip'. I hour~
]
(continued)
CEOG STS 3.3-27 Rev 1, 04/07/95
_/\j..; r/
...S-+-.1,.. ~ _.-/
DG-~~-- 3.3~@ ACTIONS continued CONDITION REQUIRED ACTION COMPLETION TIME
8. One or more Functions B.l Enter applicable 1 hour with tWQ channels per Conditions and DG i nope'""b-,1e. Require Actions for the assoc ted DG
" made inoper le by DG-LOVS instrumentation. --------NOTE---------
LCO 3.0.4 is not applicable. Pla one channel in 1 hour bypas and the other channel n trip *
C. One o re Functions C.l estore all but two
[48] hours with mor than two nnels to OPERABLE channels in erable. sta s.
'I). Enter applicable l11111ediately Conditions and Required Actions for the associated DG One or VMorc. fu~*O\l\'S made inoperable by Wi#\ 61'1& c"-o..11\~e.\ DG-f!R_.--......_ ?et' bl?' il'\.otjjlr.A.b\e.. instrumentation.
I.JI/ Sifl.~
CEOG STS 3.3-28 Rev 1, 04/07/95

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY J
s.1 SR Perform CHANNEL FUNCTIONAL TEST(Ou E:AQ\) ( PGi-U'J' '!)TAei L.a';IC.. CH~..ikia) {j) SR OIJ e=A.c.K ~ o'F-VDL..IA(;;,l:t- Aklt> ~f1P'EP Yoi.TAb'fi" C:~~tJNE\...
CEOG STS 3.3-29 Rev 1, 04/07/95
(Ref "ehng CHR lnstr,mentotrnn ~~
3.3 3.3.~~~i* [3.8.1.c] LCO ~ Two Refueling CHR Automatic Actuation Function channels and two CHR Manual Actuation Function channels APP LI CAB I l ITV: During CORE ALTERATIONS, During .ove.ent of irradiated fuel assellblies within cont1in111nt. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME
One or more Functions with one channel A.1 Place the affected channel in trip.
4 hours RAI 3.3.6-01 inoperable. A.2.1 Suspend CORE Al TERATIONS. A.2.2 Suspend 110vement of irradiated fuel asslllblies within conta1Ment. One or more Qmmediately) Functions with two channels inoperable * (continued)
( Pahsodes Nuclear Plont) ldoG ifs I 3.3-30 !Rev 17 04//91 /9$1

nstrumentot1on IS~
© ACTIONS COHO IT ION REQUIRED ACTION COMPLETION TIME
SURVEILLAHCE REQUIREMENTS SURVEILLANCE FREQUENCY
~:f ~~
[T 4.17.6(20)] SR 3.3~ Perform 1 CHANNEL CHECK on 11ch 12 hours
~monitor channel. @ID . RAI 3.3.6-01
[T 4.17 .6(20ll SR 3.3~ ~d1ys (continued)
(Pohsodes Nucleor Plont) ij'oG ifs I 3.3-31 IAfv 1/ 04@/9~
(Refuehng CAR ln>U-umentot.on~~al @
3.3. 6 SURVEILLANCE RE UIREHENTS continued ill SURVEILLANCE FREQUENCY [T4.17.6<2~~ ~tm
~8~ 11anths ffi RAI 3.3.6-01 ~*
SR~ ~~ 11anths ffi
~
(Pahsodes Nucleor Plant)
[ij'oG ifs I 3.3-32
CRIS
~~
I 3.3 INSTRUME:NTAiIOH ( I 3.3.8 Control LCO 3.3.8 tion Signal (CRIS) (Analog) 0 e CRIS channel shall be OPERABLE. MODES 1, 2, 3, 4[, 5, and 6], During CORE ALTERATIOHS, During movement of irradiated fuel
"\
CONDITION COMPLETION TIME A. CRIS Manual Trip, A.l TE--------- Actuation Logic, or Place ontrol RoOll [one or rwore required Emer ncy Air Cleanup channels of Sy e11 (CREACS) in particulate/1odino or t ic gas protection gaseous] radiation
if aut0111tic 1110nitors inoperable in transfer to toxic gas MOOE 1, 2, 3, or 4. protection llOde inoperable.
Place one CREACS 1 hour train in 8118rgency radiation protection
.ode.
B. B.l Be in MOOE 3. ;
\l MD ' \
B.2 Be in MOOE 5. I (continued) J CEOG STS 3.3-33 Rev 1, 04/07/95
~r-o~k
1,.3-JS'
CS Isolation Signal (Analog) 3.3.9
~-----~~-----------~-----
0
. i/ ----........ __ _ / 3. 3 INSTRUMENTATION ~=::::::::::-----~ ~---
3 . 3. 9 Ch e111 i ci l LCO 3.3.9 Four chinne of West Penetration Ro04l/Letdown Heat Exchanger oom pressure sensing and two Actuation Logic channel shill be OPERABLE. APPLICABILITY: ACTIONS REQUIRED ACTION One Actuation Logic A. l 48 hours channel inoperable. B. One CVCS isolation B.l in l hour instrument channel inoperable.
Atta Restore the channel to OPERABLE status.
48 hours Place the channel in 48 hours trip. C. Two CVCS isola on C.l Place one channel in inst.,...t ch nels bypass and place the 1noperaal1. other channel in trip. (continued) CEOG STS 3.3-36 Rev 1, 04/07/95 fl,.. rev-q k.
3.3- 3~

3.3 INSTRUMENTATIOH 3.3.10 Shield LCO 3.3.10 Building~
/
tration Actuation Signal Tw channels of SBFAS aut01a&tic and two channels of ip shall be OPERABLE. MODES 1, 2, 3, and 4. CC>>4PLETION TIME A. One Manual Trip or A.1 Restore the annel 48 hours Actuation logic to OPERABL status. channel inoperable. B. Required Action and B. 1 6 hours associated COl!f)letion Ti* not Mt. Altl2
Be in MOOE 5. 36 hours Ptrfora 1 CHANNEL FUHCTIOHAL TEST on each SlfAS auta.atic actuation channel.
P1rfon1 1 CHANNEL FUNCTIONAL TEST on e [18] 110nths SBFAS Manual Trip channel. CEOG STS 3.3-39 Rev 1, 04/07/95
_...,..-G);. 11 ~ PAM Instrumentation ~
~(/O)oJ\ ~~
c..-n 3.3 INSTRUMENTATION
- 3.3.~t Accident Monitoring (PAM) Instrumentation aAfY(O,>> \j.11. 4] LCO 3.3.'1) The PAM instrumentation for each Function in Table 3.3~
shall be OPERABLE. APPLICABILITY: MODES 1, 2, and 3. ACTIONS
-----------------------------------~-NOTES------------------------------------
1. LCO 3.0.4 is not applicable~

2. Separate Condition entry is allowed for each Function.
CONDITION REQUIRED ACTION COMPLETION TIME .
**. i}.11.4-. l]
\_3.1-1. +. 3]
l!.11. +.s] A. One or more Functions with one required channel inoperable. A.l Restore required channel to OPERABLE status. 30 days
8. Required Action and 8.1
Initiate action i~ lnmediately associated Completion accordance with ~
Time of Condition A Specification 5.6 . not 111et. [J.n. 4. 2.] c. ---------NOTE--------- C.l Restore one channel to OPERABLE status. 7 days Not applicable to [i. 11.4-, L.] hydrogen 110n1tor ' channels. One-..or 110r1 Funct 1ans with ttlMI required channels inoperable. (continued) (
CEOG STS 3.3-40 Rev 1, 04/07/95

ACTIONS (continued)
CONDITION PAM Instrumentation REQUIRED ACTION
~ ~
COMPLETION TIME {§ac.. L.i. J D. Two hydrogen monitor 0.1 Restore one hydrogen 72 hours channels inoperable. monitor channel to OPERABLE status. [!.11.4.4-] E. Required Action and E.1 Enter the Condition !mediately associated Completion. referenced in (1.1;,4,1] Time of Condition C Table 3.3.~r or D not met. the channel. 1 [3, 11, 4. 4.,J F. As required by F.1 Be in MODE 3. 6 hours Required Action E.l
~* 11. 4-1 7. y and referenced in Arm Table 3.3.~l.
F.2 Be in MODE 4. ~s
a. 11. 4.7. i:.J [As required by Required Action E.l and referenced in Table 3.3.~.
G.l Initiate action i~ accordance with Specification 5.6. ; I lnnediately ]
CEOG STS 3.3-41 Rev 1, 04/07/95

SURVEILLANCE REQUIREMENTS PAM Instrumentation ~.JI&-G) rh;;;-5R;-;~~;;-t~-;;~h-PAH-1~;t;~;;~~~i~~~-F~~~ti~~-1~-1;b;;-3:3:~-------
SURVEILLANCE FREQUENCY 31 days Eb
--------------------NOTE-------------------
Neutron detectors are excluded from CHANNEL CALIBRATION. Perform CHANNEL CALIBRATION
i1~ months
(
CEOG STS 3.3-42 Rev 1, 04/07/95
. Table 3.3.1.cpage 1 of 1)
PAM Post Accident Mor11toring Instrumentation Instrumentation~
*~
COllDITIONS ~ REFERENCED FR~,.-& C*o.11.4-J FUllCTIOll REQUIRED CKAlllleLS REQUIRED ACTION@ 1
~- 2 6)~ 2@ .. l1'!)'L ©-(). 2 (Jlii' (Q!i) e_. ~ Coolant Syst* Pressure (11ide r.,..) 2 ~ Ruc:tor Vess~~itt~ 00 2 @----$ Ccntair'lll!nt Wa r eve\ (11ide range) ~ Ccntairmmt Pressure (11ide range) ~- Ccntai,_,t lsol1ticn V1lve Position
((!)-d. tlr-o. F F F -<RV [3.11.4.~~ (1) llot r..,ired for isolation valves lllllo9e aaoc:iated penetration ii laolated by at lHlt one clOMd Ind de* activated .rt-tic valve, clOHd ..n.L valve, bl ird flanae, or check valve 11itll flov thl'QUgil the valve secured. rfb> ~one poa~on can~l r- irdica'-fan cltamel !~ired for ~tratton fl"" paths 11itfi only one ;rs;:ullecf t~tion cll'lrwiel.
In the ~s legu~ry
© S". Si..>bcoci\t>J ~d.rg i 't'I Mo't'li+or F I?, 'S"-te4.IM 4 H~rA'1PY'A 'fr~s S' vre F @ \'t. 45;(1~ 6~~ 8 Pr<!'SfVtt' ,:.
CEOG STS 3.3-43 Rev 1, 04/07/95
~
ArterV1c..,1e ) _.._..,..___ Shutdown System ~ °"~ 3.3 INSTRUMENTATION ~ 3.3.~ (}. 11.5] LCO ~ The Shutdown System Functions in Table be OPERABLE. 3.3.~ shall APPLICABILITY: MODES 1, 2, and 3. ACTIONS
---------------------~--------------NOTES------------------------------------
1. LCO 3.0.4 is not applicable.

2. Separate Condition entry is allowed for each Function.
------------------------------~---------------------------------~-------------
CONDITION REQUIRED ACTION COMPLETION TIME A. One or more required A.I Restore required 30 days RA.I. Functions inoperable. Functions to OPERABLE 3,3.s-oi status. 3,3.8*0"7 [3.11.~.z.J B. Required Action and B.1 Be in MOOE 3. 6 hours associated Completion Time not 11et.
~
AHl2 B.2 Be in MOOE 4. ours
~~
3.'3.B-o'l.
!.~.s-01
CEOG STS 3.3-44 Rev 1, 04/07/95

SURVEILLANCE REQUIREMENTS
~.-....- Shutdown System {J\iiiThnL 3.3.5'"
SURVEILLANCE FREQUENCY SR 3.3. Verify each required control circuit and @la@ months transfer switch is capable of perfonning the intended function. SR 3.3.~-----------------NOTE-------------------
~Neutron detectors are excluded from the CHANNEL CALIBRATION. .
Perform CHANNEL CALIBRATION for each ~laj months required instrumentation channel
ED

CEOG STS 3.3-45 Rev 1, 04/07/95
l
~ 'i .~
SECTION 3.3 SURVEILLANCE INSERT 1 FREQUENCY SR 3.3.8.1 Perform CHANNEL FUNCTIONAL TEST of the Once within 7 days Source Range Neutron Flux Function. prior to each reactor startup

3.3-45
~

Tlble 3.3.12-1 (peve 1 of 1)
~~ SllutdoWI Syst* lrwt~tation rd Controls F\JICTIOll/IMSTltt.IEllT IECIUllED Cl allTICl l'AUMETEI ~ OF DIYISIOllS
1. leec:tivi ty Control

b. Source lr.ge Neutron Flux

c. l119Ctor Trip Circuit lrealt*r Position

d. ....._.l leec:tor Trip

2. ll:eec:tor COOl.,,t Sylt* Preuure Control
** P~fzer P....-e m or .....
ICS wi'ae
*,,1.,... l'raaure
Preuuri z~~ Oper*ted
b. [1, control* - t be for powr lelfef Y*lve ~trol Ind opereted relief valve rd llock Y*lw ~l lock Vlllws on . - lfneJ

3. Decay Hut 1_.1 vi*~,~ Genenton
** leector Cool.,,t Not Leg Temperature
b. leector Cool...t [1 per loop]
told L119 Tmperetwe
c. Aux! l i*ry F..-ter c:ontrol1 m
[1 per 1t... generator]
e. St. . . liener*tor Level [1 per 1t. . . generator]
liary F..-Cer FLOM
f. m
"* il. [1]
b. m CEOG STS 3.3-46 Rev 1, 04/07/95
SECTION 3.3
INSERT 1 Table 3.3.8-1 (page 1 of 1)
Alternate Shutdown System Instrumentation and Controls FUNCTION REQUIRED INSTRUMENT OR CONTROL PARAMETER CHANNELS
1. Source Range Neutron Flux 1

2. Pressurizer Pressure 1

3. Pressurizer Level 1

4. Primary Coolant System (PCS) #1 Hot Leg Temperature 1

5. PCS #2 Hot Leg Temperature 1

6. PCS #1 Cold Leg Temperature 1

7. PCS #2 Cold Leg Temperature 1

8. Steam Generator (SG) A Pressure 1
9. SG B Pressure 1
10. SG A Wide Range Level 1

11. SG B Wide Range Level 1

12. Safety Injection Refueling Water (SIRW) Tank Level 1

13. Auxiliary Feedwater (AFW) Flow Indication to SG A 1

14. AFW Flow Indication to SG B 1

15. AFW Low Suction Pressure Alann (P-8B) 1

16. AFW Pump P-8B Steam Supply Valve Control 1

17. AFW Flow Control to SG A 1 18 . AFW Flow Control to SG B 1 3.3-46
C.T 'S
3. 3 INSTRUMENTATI ON t-Je.v+v o>-\ \='l u 'ii. ~
3.3.ql ~ ith
P'bwe Monitoring Channels ~l'Oi)
~ ~eu+Yo~ Flui'-J
[3./7. "J LCO 3.3.£! Two channels of[...,~0-cm:...,,..,,.1.,..,tn~lfik,..,,..,,...]~p'l:":&Wi~r:--""fiVil>C".'.~ iiM monitoring r:. i instrumentation shall be OPERABLE. 1.J3.n.t.(1'l..i Q. 11. t.J APPLICABILITY: fr3. ,., .(. (1)] ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME [3.11. lo.I] A. One or more required A. l Suspend all Inned1ately channel(s) inoperable. operations involving . positive reactivity additions *
AND A.2 Perfona SOM verification in accordance with SR 3.1.1.1 4 hours Once per 12 hours thereafter SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3* *1' Perform CHANNEL CHECK. 12 hours
-*~
(continued)
CEOG STS 3.3-47 Rev 1, 04/07/95
-C...T~
SURVEILLANCE REDUIREMENTS (continued) SURVEILLANCE FREQUENCY [J-4.tl,t.,] SR 3.3.~~~~----------------NOTE-------------------- Neutron detectors are excluded from CHANNEL CALIBRATION. Perform CHANNEL CALIBRATION . iia{>> months 3.3.JO N e IU "Pa.~ e
r.tJ51L T i Spec.i-'-\~A.""~ 3.3.IOl 11
© '\SR Vt.t\-t:\a..+'~c:M.. 'Ivd~.\-r~Me.~+A.*h~
(
CEOG STS 3.3-48 Rev 1, 04/07/95

c-r~
3.3 INSTRUMENTATION SECTION 3.3 INSERT 1
3. 3 .10 Engineered Safeguards Roo:Ql (ESR) Ventilation Instrumentation
[ 3./7.3:-J LCO 3.3.10 Two channels of ESR Ventilation Instrumentation shall be 3 L!'Z*_ J OPERABLE. [s.17.3] APPLICABILITY: MODES 1, 2, 3, and4. ACTIONS
-------------------------------------NOTE-------------------------------------
[ neu:iJ Separate Condition entry is allowed for each channel. CONDITION REQUIRED ACTION COMPLETION 7 TIME
[3.t1.3,tl] A . One or more channels inoperable .
A. l Initiate action to isolate the ESR Ventilation System. Immediately
. SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.3.10.1 Perform a CHANNEL CHECK. 12 hours Perform a CHANNEL FUNCTIONAL TEST. 31 days SR 3.3.10.3 Perform a CHANNEL CALIBRATION. 18 months Verify high radiation setpoint on each ESR Ventilation Instrumentation radiation monitoring channel is s: 2.2E+5 cpm .
3.3.10-1
ATTACHMENT 6 JUSTIFICATION FOR DEVIATIONS SPECIFICATION 3.3.1, RPS INSTRUMENTATION 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 word Analog was removed from the title of each Specification in Section 3.3.
Palisades is an Analog plant and specifically listing this in the title is unnecessary. Analog was placed in NUREG-1432 to distinguish between Analog and Digital Specifications.
7. The word "Operating" has been deleted from the title of ISTS 3. 3. 1, "Reactor Protective System (RPS) Instrumentation - Operating," since ISTS 3.3.2, "Reactor Protective System (RPS) Instrumentation - Shutdown" is not being proposed in the ITS. The ISTS "Power Rate of Change Function" is not credited in the Palisades safety analysis therefore, ISTS 3.3.2 does not apply. The requirements of RPS, both during operating and shutdown modes are included in ITS 3.3.1, "RPS Instrumentation." Reference 3.3.2 has been deleted throughout the BASES ISTS .
Palisades Nuclear Plant Page I of 4 05/30/99

Change Discussion ATTACHMENT 6 JUSTIFICATION FOR DEVIATIONS SPECIFICATION 3.3.1, RPS INSTRUMENTATION
8. The Applicability has been revised to be consistent with the current licensing basis.
TSTF-85, Rev. 1, is incorporated, but revised as necessary to reflect the current licensing basis. The CTS applicability reflects the Palisades safety analyses assumptions with respect to RPS operation. The CTS RPS LCO applicability was issued as part of Amendment 162. The broader ITS applicability encompasses the applicable conditions of STS LCOs 3.3.1 and 3.3.2. The ITS (and CTS) wording "when more than one [full length] control rod is capable of being withdrawn" replaces the STS wording "with any RTCB's closed and any control element assemblies capable of being withdrawn" .
.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. The part length control
*rods have no clutches, remain fully withdrawn during operation, and are unaffected by RPS functions.
Palisades is not equipped with RTCBs; power to the CRDM clutches is interrupted, to initiate a scram, by de-energization of normally energized contactors. These contactors are addressed in ITS LCO 3.3.2. Their functioning is explained in the Bases.
9. The Palisades RPS design does not include Reactor Trip Circuit Breakers (RTCBs),
therefore any reference to these in ITS 3.3.1 has been deleted.
10. ISTS 3.3.1 Required Action A. l and A.2.1 have been deleted since the remaining OPERABLE channels can provide the required trip even with another channel failure.
The requirement to place the channel in trip within 7 days is considered adequate. These allowance were approved in Amendment 162 of the Palisades Operating license.
11. The Palisades RPS design does not include Core Protection Calculators (CPCs),
therefore any reference to these in ITS 3.3.1 has been deleted.
12. The requirement to place one trip unit in bypass has been deleted since this action does not restore the trip capability of the affected Function.
\
13 . The Palisades RPS design does not include Axial Power Distribution (APD) Trips, therefore any reference to these in ITS 3.3.1 has been deleted. Palisades Nuclear Plant Page 2 of 4 05/30/99
Change Discussion ATTACHMENT 6 JUSTIFICATION FOR DEVIATIONS SPECIFICATION 3.3.1, RPS INSTRUMENTATION
14. ISTS SR 3.3.1.7 has been deleted since the 18 month CHANNEL CALIBRATION surveillance (ITS SR 3.3.1.8) is considered adequate to ensure the bypass removal channels are functioning properly. This change is consistent with the plant current licensing basis. Subsequent surveillances have been renumbered, where applicable.

15. ISTS SR 3.3.1.9, the RPS RESPONSE TIME test is not included in the ITS. This test is not required by the current licensing basis since the conclusions of NUREG-082, "Integrated Plant Safety Assessment Systematic Program Palisades Plant" - Final Report USNRC 10/82, concluded that backfitting this testing into the current program was not required.

16. ISTS 3.4.17, "RCS Loops -Test Exceptions" is not being proposed therefore reference to this specification is deleted.

17. Figures 3.3.1-1, 3.3.1-2, and 3.3.1-3 do not apply to Palisades. Table 3.3.1-2 has been included which provides the required relationship and allowable values at the Thermal Margin/Low Pressure (TM/LP) Function.

18. Table 3.3.1-1 Footnote (c) does not'apply, subsequent Footnotes have been renumbered, where applicable.

19. The specific wording which discusses other plants is deleted; the bases are specific to Palisades and will contain, where possible, only Palisades specific information.

20. The bracketed Reviewer's Note has been deleted since it is not meant to be maintained in the plant specific ITS.

21. ISTS Figure B 3.3.1-1 has been deleted.since similar diagrams are already included in FSAR Figure 1 and 7-2.

22. Note 2 of ISTS SR 3.3.1.2 has been deleted since the PHYSICS TESTS are performed below 2 % RTP and therefore, the allowance in Note 1 suffices.

23. Not Used

24. Requirement to verify constants associated with the thermal margin monitors is added, consistent with CTS Table 4.17.1, Item 15 .
Palisades Nuclear Plant Page 3 of 4 05/30/99
Change ATTACHMENT 6 JUSTIFICATION FOR DEVIATIONS SPECIFICATION 3.3.1, RPS INSTRUMENTATION Discussion
25. TSTF-178 is incorporated to omit "trip or bypass removal" from the ACTIONS Note.
The RPS Functions listed in Table 3. 3. 1-1 include trip and bypass removal features where appropriate. Referring to the trip or bypass removal features as separate Functions is incorrect and confusing. Removing the words "trip_ or bypass removal" satisfies the intent of the Note and eliminates the error. This change is also consistent With the CEOG Digital LCO.
26. TSTF-179 is incorporated to revise NUREG Required Action C.2 text from "70% of the maximum allowed THERMAL POWER level" to "70 % RTP." Replacing the undefined phrase"maximum allowed THERMAL POWER level" with the defined phrase "RTP" eliminates possible misinterpretation of the Action and is consistent with
. the conventions in the NUREGs. This change is also consistent with the CEOG Digital Actions.
27. TSTF-189 is incorporated to remove thediscussion of the allowable value and it's
28.
uncertainty from the Containment Pressure - High Bases. References to instrument uncertainty in the Bases are inconsistent with the ITS conventions and not given in other Specifications. The change is consistent with the CEOG Digital Bases. TSTF-80, Rev. 1, is not incorporated, rather the current licensing basis is retained as an added Action for the Loss of Load trip function.
29. The following alterations were made to STS LCO 3.3.1 to implement the CTS Operability, Action, and Surveillance requirements for the instrument channels which provide automatic removal of the Zero Power Mode (ZPM) bypass of certain RPS trip
. functions.
The LCO wording and Table 3.3.1-1 have been changed to explicitly call out the ZPM bypass removal channels as a required Function. Condition "D" and the associated Actions have been revised to reflect the CTS requirements for an inoperable ZPM bypass removal channel. SR 3.3.1.8 has been revised to omit the reference to bypass removal channels because Table 3.3.1-1 explicitly requires that SR for the ZPM bypass removal Function. The ZPM bypass removal channels are the only bypass removal channels required to be Operable in the CTS and the only ones assumed to function in the accident analyses. This is a plant specific change. Palisades Nuclear Plant Page 4 of 4 05/30/99
Change SPECIFICATION 3.3.2, RPS INSTRUMENTATION - SHUTDOWN Discussion ATTACHMENT 6 JUSTIFICATION FOR DEVIATIONS 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.
4.
The following requirements have been renumbered, where applicable, to reflect this deletion. 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. ISTS LCO 3.3.2, "RPS Instrumentation - Shutdown," is not used. The ISTS I LCO 3.3.2 addresses the Power Rate of Change - High trip. The Palisades design I does not credit the comparable trip, High Startup Rate, in any of the Safety Analyses I and that trip is not required to be Operable by the CTS. This lack of reliance is part of I the plant design; the High Startup Rate trips are not installed as safety grade I components. Their design is discussed in some detail in the Bases for LCO 3.3.1. I Therefore, this specification is not applicable to the Palisades Nuclear Plant. Instead, I the CTS LCO 3.17.1 and ITS LCO 3.3.1 utilize an applicability includes all 1 ..
conditions when the Palisades analyses credit functioning of the RPS to trip the reactor. I That applicability was approved as part of Amendment 162 to the Palisades operating I license. TSTF-82, Rev. 1, and TSTF-1 ~O are also not incorporated on the same basis. I Palisades Nuclear Plant Page 1of1 05/30/99
Change SPECIFICATION 3.3.3, RPS LOGIC AND TRIP INITIATION Discussion ATTACHMENT 6 JUSTIFICATION FOR DEVIATIONS 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.
4.
The following requirements have been renumbered, where applicable, to reflect this deletion. Changes have been made (additio11:s, 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 word Analog was removed from the title of each Specification in Section 3.3.
Palisades is an Analog plant and specifically listing this in the title is unnecessary. Analog was placed in NUREG-1432 to distinguish between Analog and Digital Specifications.
7. ISTS 3.3.2 is not included in the ITS, therefore subsequent Specifications have been renumbered accordingly.

8. The Palisades RPS design does not include Reactor Trip Circuit Breakers (RTCBs),
therefore any reference to these in ITS 3.3.2 has been deleted. TSTF-79, Rev. 1, and TSTF-192 are also not incorporated on the same basis.
\
9. The Applicability has been revised to be consistent with the current licensing basis.
Palisades Nuclear Plant Page 1of2 05/30/99
Change ATTACHMENT 6 JUSTIFICATION FOR DEVIATIONS SPECIFICA'PION 3.3.3, RPS LOGIC AND TRIP INITIATION Discussion
10. ITS 3.3.2, Condition A "Note" has been deleted because it does not apply to Palisades.
A failure of a common power source will not de-energize the matrix relays. TSTF-83 is also not incorporated on the same basis.
11. ITS 3.3.2, Condition B has been revised to reflect the current licensing basis. This change is necessary due to design differences between the Palisades RPS Logic and the reference plant that NUREG-1432 is based upon. TSTF-170 is also not incorporated on the same basis.

12. For ITS 3.3.2, Condition Band C, the "Note" has been deleted. The test allowance is already provided by LCO 3.0.5. Therefore, this allowance is not necessary. This change is consistent with TSTF-181.

13. The bracketed "Reviewer's Note" has been deleted as inappropriate to be retained in the plant specific ITS .
14. TSTF-182 which would add Manual Trip to the Condition for two channels of RPS Logic and Trip Initiation affecting a single leg inoperable. Manual Trip is provided by only two channels for this plant; either of which affects both trip legs. The generic STS is based on four chan_nels of Manual Trips; each of which affects only one trip leg.
I I Since having two channels of Manual Trip inoperable at Palisades would be a complete 1* loss of all Manual Trip capability, and the associated Required Actions of TSTF 182 I would trip the plant, this change is not appropriate and is not incorporated. Rather, I two inoperable Manual Trip channels will result in entry into Condition E, with Required Actions to exit the applicable conditions of the Specification. These actions are appropriate for the condition, and consistent with the CTS . Palisades Nuclear Plant Page 2 of 2 05/30/99
Change SPECIFICATION 3.3.4, ESFAS INSTRUMENTATION Discussion ATTACHMENT 6 JUSTIFICATION FOR DEVIATIONS 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.
4.
The following requirements have been renumbered, where applicable, to reflect this deletion. Changes have been made (additions, deletions, and/or changes to the NUREG) to reflect the facility specific nom~nclature, number, reference, system description, or analysis description. '
5. This change reflects the current licensing basis/technical specification.

6. The word Analog was removed from the title of each Specification in Section 3.3.
Palisades is an Analog plant and specifically listing this in the title is unnecessary. Analog was placed in NUREG-1432 to distinguish between Analog and Digital Specifications.
7. ISTS 3.3.2, is not included in the ITS, therefore the followi~g specifications have been renumbered, where applicable to reflect this deletion.

8. ISTS 3.3.4 Action A has been deleted since it does not apply. Therefore, the following Actions have been renumbered where applicable to reflect this deletion.

9. The Applicability has been revised since certain Functions are applicable in MODES 1, 2, and 3, while others are applicable in MODES 1, 2, 3, and 4. The Applicability has been changed to "As specified in Table 3.3.3-1," and a "MODES" column has been included in Table 3.3.3-1. The proposed Applicability is consistent with the current licensing basis.
Palisades Nuclear Plant Page 1of2 05/30/99
Change ATTACHMENT 6 JUSTIFICATION FOR DEVIATIONS SPECIFICATION 3.3.4, ESFAS INSTRUMENTATION Discussion
10. Bypass removal channels are addressed in both LCO 3.3.3 and LCO 3.3.4; the bistables and instrument channels in 3.3.3, and the Logic channels in 3.3.4. The bypass removal bistables and their instrument channels are subject to a Channel Functional Test as part of the Channel Calibration required by ITS SR 3.3.3.3; the bypass removal Logic channels are subject to a Channel Functional Test by ITS SR 3.3.4.3. and most ESP Functions do not have "trip units". Due to these design differences:
The LCO wording "trip units and associated instrument and bypass removal channels," has been changed to "bistables and associated instrument channels." The related ISTS SR 3.3.4.3, i.e., the CHANNEL FUNCTIONAL TEST on each automatic bypass removal function, is not included in the ITS. The Completion Time for placing a channel in trip or bypass has been changed
11.
to 8 hours, consistent with the current licensing basis/Technical Specifications, as approved in Amendment 162 of the Palisades*operating license. ITS Table 3.3.3-1 identifies the applicable SRs for each ESP Instrumentation Function. The appropriate generic SR Note is included to address this format and the wording of the SRs has also been modified to reflect this change.
12. ISTS SR 3.3.4.5, the ESP RESPONSE TIME test, is not included in the ITS. This test is not within the current licensing basis since the conclusions of NUREG-0820, "Integrated Plant Safety Assessment Systematic Program, Palisades Plant" - Final Report USNRC 10/82, concluded that backfitting this testing into the current program was not required.

13. The specific wording which discusses other plants is being deleted. The Bases are specific to Palisades and will contain, where possible, only Palisades specific information.

14. The text in the Bases concerning the justification for surveillance frequencies with the use of topical reports has been deleted since it was not intended to be included in the ITS .
Palisades Nuclear Plant Page 2 of 2 05/30/99
Change SPECIFICATION 3.3.5, ESFAS LOGIC AND MANUAL TRIP Discussion ATTACHMENT 6 JUSTIFICATION FOR DE VIA TIO NS 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.
4.
The following requirements have been renumbered, where applicable, to reflect this deletion. 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 word Analog was removed from the title of each Specification in Section 3.3.
Palisades is an Analog plant and specifically listing this in the title is unnecessary. Analog was placed in NUREG-1432 to distinguish between Analog and Digital Specifications.
7. ISTS 3.3.2, is not included in the ITS, therefore the following specifications have been renumbered, where applicable to reflect this deletion.

8. The SRs have been modified to reflect the current licensing basis. Proposed SRs 3.3.4.1 and 3.3.4.2 represent explicit CTS functional test requirements.
SR 3.3.4.3, the Channel Functional Test requirement, applies to all channels specified in the LCO. *
9. Bypass removal logic channels are addressed in LCO 3.3.4, rather than LCO 3.3.3, because they bypass the logic channels rather than individual instrument channels, and are tested along with the ESF logic, consistent with current design and licensing basis.
Palisades Nuclear Plant Page 1of2 05/30/99
Change ATTACHMENT 6 JUSTIFICATION FOR DE VIA TIO NS SPECIFICATION 3.3.5, ESFAS LOGIC AND MANUAL TRIP Discussion
10. ISTS SR 3.3.5.1 Notes have has been deleted. Circumstances described by the Notes are not applicable to Palisades.

11. Not used.

12. TSTF-187 is incorporated to add a Condition for two inoperable Actuation Logic Channels to ESF Logic and Manual Trip. There is currently no Condition for two inoperable Actuation Logic channels. This change adds the Condition of two Actuation Logic channels inoperable with the Required Action to shutdown the plant. This change puts the appropriate Actions in the Specification and eliminates any confusion that may arise from not addressing the AFAS actuation logic and manual trip functions in the Actions .
Palisades Nuclear Plant Page 2 of 2 05/30/99
Change Discussion ATTACHMENT 6 JUSTIFICATION FOR DEVIATIONS SPECIFICATION 3.3.6, DG - LOVS 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.
4.
The following requirements have been renumbered, where applicable, to reflect this deletion. 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. NUREG-1432 Specification 3.3.6, Actions Note, Actions A, B, and C have been deleted. The Palisades Nuclear Plant instrumentation associated with Diesel Generator (DG) - Loss of Voltage Start (LOVS) is not designed to facilitate bypassing individual channels or placing individual channels in trip.

7. NUREG-1432, Specification 3.3.6, Condition D has been revised. The Palisades Nuclear Plant instrumentation associated with Diesel Generator (DG) - Loss of Voltage Start (LOVS) does not have readily available channel bypass or trip capability.
Therefore, with one or more Functions with one or more channels per DG inoperable the only means of addressing channel inoperabilities is to declare the associated DG(s) inoperable. Palisades Nuclear Plant Page 1of2 05/30/99
Change Discussion ATTACHMENT 6 JUSTIFICATION FOR DE VIA TIO NS SPECIFICATION 3.3.6, DG - LOVS
8. NUREG-1432, Specification 3.3.6, SR 3.3.6.1 has been deleted. The Palisades Nuclear Plant instrumentation associated with Diesel Generator (DG) - Loss of Voltage Start (LOVS)design does not include metering in the control room associated with the undervoltage sensing channels. The remaining ITS SRs have been editorially revised to indicate that the Channel Functional Test applies to the logic channels and the Channel Calibration, to the undervoltage sensors.

9. Applicability of LCO 3.3.5 is revised to read, "When associated DG is required to be OPERABLE." This terminology is consistent with NUREG-1432 Specification 3.8.3, "Diesel Fuel Oil, Lube Oil, and Starting Air."

10. The word Analog was removed from the title of each Spedfication in Section 3. 3.
Palisades is an Analog plant and specifically listing this in the title is unnecessary. Analog was placed in NUREG-1432 to distinguish between Analog and Digital Specifications .
11. ISTS 3.3.2 is not included in the ITS, therefore subsequent Specifications have been renumbered accordingly .
Palisades Nuclear Plant Page 2 of 2 05/30/99
Change SPECIFICATION 3.3.7, CONTAINMENT PURGE ISOLATION SIGNAL (CPIS)
Discussion ATTACHMENT 6 JUSTIFICATION FOR DE VIA TIO NS 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 word Analog was removed from the title of each specification in Section 3.3.
Palisades is an Analog plant and specifically listing this in the title is unnecessary. Analog was place in NUREG-1432 to distinguish between Analog and Digital Specifications.
7. ISTS 3.3.2, is not included in the ITS, therefore the following specifications have been renumbered, where applicable to reflect this deletion. \

8. Palisades includes two Refueling Containment High Radiation (CHR) instrument channels which are required to be OPERABLE during CORE ALTERATIONS, and during movement of irradiated fuel assemblies within containment. The purpose of
*these channels are to close the containment penetrations which have direct access from the containment atmosphere to the outside atmosphere and to initiate the Control Room Ventilation System in the emergency mo~e of operation. Since these channels perform a dual function which provide the protection as required by ISTS 3.3. 7 and ISTS 3.3.8, the requirements have been combined and included into ITS 3.3.6, "Refueling Containment High Radiation (CHR) Instrumentation.
Palisades Nuclear Plant Page 1of2 05/30/99
Change SPECIFICATION 3.3.7, CONTAINMENT PURGE ISOLATION SIGNAL (CPIS)
Discussion ATTACHMENT 6 JUSTIFICATION FOR DE VIA TIO NS
9. ISTS LCO 3.3.7 has been changed to reflect the Palisades equipment (ITS LCO 3.3.6). Palisades utilizes two refueling radiation monitors, each of which directly actuates one train of the isolation functions actuated by the Containment High Radiation (CHR) signal used during plant operation, where the ISTS reflects use of one train four monitor in a two-out-of-four logic. ITS LCO 3.3.6 also requires two CHR manual actuation channels to be operable, where the ISTS requires only one. ITS Condition A provides four hours for restoration when the LCO is not met, but at least one channel of automatic actuation and one channel of manual actuation are available. If equipment, inoperabilities reduce the actuation capability beyond that level, Condition B requires immediate action to place the plant in a condition outside the applicability of the LCO. There are two Refueling CHR channels in which each is directly connected (one-out-of-one logic) to actuate its associated CHR output relays. Thus the term "automatic Actuation Logic" has been deleted throughout the Specification .
10. The verification of the Allowable Value has been removed from the CHANNEL CALIBRATION surveillance since for consistency with the current licensing basis.
This radiation monitor setpoint is not a specific assumption in the safety analysis and as such, an Allowable Value cannot be developed. The CHANNEL CALIBRATION will continue to check the setpoint, but changes to the setpoint will not require a License amendment.* TSTF-186 is also not incorporated on the same basis.
11. Details related to location, associated logic, system ties, and other operational considerations associated with the containment radiation monitors is added, consistent with details relocated from CTS 3.8.1 (see DOC LA. I).

12. Details related to ancillary requirements associated with inoperable containment radiation monitors is added, consistent with details relocated from CTS 3.8.2 (see DOC LA.2).

13. ISTS SR 3.3.7.6, RESPONSE TIME test is not included in the ITS. This test is not I required by the current licensing basis since the conclusions of NUREG-082, I "Integrated Plant Safety Assessment Systematic Program Palisades Plant" - Final I Report USNRC 10/82, concluded that backfitting this testing into the current program I was not required . I
Page 2 of 2 05/30/99 Palisades Nuclear Plant

Change SPECIFICATION 3.3.8, CONTROL ROOM ISOLATION SIGNAL (CRIS)
Discussion ATTACHMENT 6 JUSTIFICATION FOR DEVIATIONS 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 th~oughout 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 fo this fadlity.
4.
The following requirements have been renumbered, where applicable, to reflect this deletion. 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. ISTS LCO 3.3.8, "Control Room Isolation Signal (CRIS)," is not used. Palisades design does not include a Control Room Isolation Signal (CRIS). Therefore, this specification is not applicable to the Palisades Nuclear Plant.
Palisades Nuclear Plant Page 1of1 05/30/99
Change ATTACHMENT 6 JUSTIFICATION FOR DEVIATIONS SPECIFICATION 3.3.9, eves ISOLATION SIGNAL 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.
4.
The following requirements have been renumbered, where applicable, to reflect this deletion. 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. ISTS LCO 3.3.9, CVCS Isolation Signal," is not used. Palisades design does not include a CVCS Isolation Signal. Therefore, this specification is not applicable to the Palisades Nuclear Plant. TSTF-84 and TSTF-187 are also not incorporated on the same basis .
Palisades Nuclear Plant Page 1of1 05/30/99
SPECIFICATION 3.3.10, SHIELD BUILDING FILTRATION ACTUATION SIGNAL Change Discussion ATTACHMENT 6 JUSTIFICATION FOR DEVIATIONS 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. ISTS LCO 3.3.10, "Shield Building Filtration Actuation Signal," is not used.
Palisades design does not include a Shield Building Filtration Actuation Signal. Therefore, this specification is not applicable to the Palisades Nuclear Plant. Palisades Nuclear Plant Page 1 of 1 05/30/99 i
Change SPECIFICATION 3.3.11, PAM INSTRUMENTATION Discussion ATTACHMENT 6 JUSTIFICATION FOR DEVIATIONS 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. Not used.

7. Containment Isolation Valve (CIV) required channels is modified to reflect plant design, which includes only one position indication channel per valve, and not all penetrations are equipped with CIVs of a type that is equipped with position indication, consistent with current design and licensing basis.

8. Details related to method of calibrating core exit thermocouples are added, consistent with details relocated from CTS Table 4.17.4, Footnote (a) (see DOC LA. l).

9. The word Analog was removed from the title of each specification in Section 3.3.
Palisades is an Analog plant and specificaily listing this in the title is unnecessary. The terms, "Analog" and "digital," were placed in NUREG-1432 to distinguish between the Specifications for the two types of instrumentation. Palisades Nuclear Plant Page 1of2 05/30/99
Change SPECIFICATION 3.3.11, PAM INSTRUMENTATION Discussion ATTACHMENT 6 JUSTIFICATION FOR DEVIATIONS
10. ISTS 3.3.2, 3.3.8, 3.3.9 and 3.3.10 are not included in the ITS, therefore ISTS 3.3.11 has been renumbered as 3.3.7. The Specification, LCO and Surveillances have been renumbered, where applicable, to reflect these deletions.

11. In the proposed ITS LCO 3.0.3b, NUREG 1432 lists a time to reach MODE 4 of 13 hours. This item is "bracketed" in NUREG-1432 since some plant designs require a different number. In the proposed Palisades ITS, this time is increased to 31 hours.
Increasing the time allowed to reach MODE 4 allows for more complete degassing of the Primary Coolant System (PCS). The PCS is degassed by venting the pressurizer gas space to the Vacuum Degasifier. The efficiency of this method is maximized by maintaining PCS temperature as high as practical, the subcooling as low as practical, and operating all pressurizer heaters. This results in a net increase in the rate of hydrogen removal from the PCS since increased spray flow and lower PCS pressure offset the lower degas flow rate through the vent path. While the total time to reach MODE 4 is increased, the time to reach MODE 5 is the same in the proposed ITS as
12.
specified in NUREG-1432 . TSTF-188 is incorporated to remove footnote (c) from CST level in Table 3.3.11-1. Footnote (c) deals with core exit thermocouples and applies to Functions 14-17. It was inappropriately applied to Function 13, Condensate Storage Tank. This corrects the error . Palisades Nuclear Plant Page 2 of 2 05/30/99
Change SPECIFICATION 3.3.12, REMOTE SHUTDOWN SYSTEM Discussion ATTACHMENT6 JUSTIFICATION FOR DEVIATIONS 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.
4.
The following requirements have been renumbered, where applicable, to reflect this deletion.
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 bracketed SR for CHANNEL CHECK of the normally energized Alternate (Remote) Shutdown System is omitted since no Alternate Shutdown System Functions are normally energized.

7. Not used.

8. The Applicable Safety Analyses section of the Bases has been modified to reference 10 CFR 50.36(c)(2), consistent with the Bases for other Specifications.

9. Alternate Shutdown System transfer ~witches and their location are added, consistent with details relocated from CTS Tables 3.17.5 and 4.17.5 (see DOC LA.1).
~
Palisades Nuclear Plant Page 1of2 05/30/99
Change SPECIFICATION 3.3.12, REMOTE SHUTDOWN SYSTEM Discussion ATTACHMENT 6 JUSTIFICATION FOR DEVIATIONS
10. ISTS 3.3.2, 3.3.8, 3.3.9 and 3.3.10 are not included in the ITS, therefore ISTS 3.3.12 has been renumbered as 3.3.8. The Specification, LCO and Surveillances have been renumbered, where applicable, to reflect these deletions.

11. . The word Analog was removed from the title of each specification in Section 3. 3.
Palisades is an Analog plant and specifically listing this in the title is unnecessary. The terms, "Analog" and "digital," were placed in NUREG-1432 to distinguish between the Specifications for the two types of instrumentation . Palisades Nuclear Plant Page 2 of 2 05/30/99
Change SPECIFICATION 3.3.13, POWER MONITORING CHANNELS Discussion ATTACHMENT 6 JUSTIFICATION FOR DEVIATIONS 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
4.
deletion . 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 word Analog was removed from the title of each specification in Section.3.3.
Palisades is an Analog plant and specifically listing this in the title is unnecessary. The terms, "Analog" and "digital," were placed in NUREG-1432 to distinguish between the Specifications for the two types of instrumentation.
7. ISTS 3.3.2, 3.3.8, 3.3.9 and 3.3.10 are not included in the ITS, therefore ISTS 3.3.13 has been renumbered as 3.3.9. The Specification, LCO and Surveillances have been renumbered, where applicable, to reflect these deletions.

8. The Applicability of the Specification has been modified to be consistent with the current Technical Specifications to ensure the neutron flux indicators are available whenever the plant is shutdown but not in MODE 6, MODES 3, 4, and 5 (MODE 6 is covered by ITS LCO 3.9.2). The Palisades design does not include reactor trip circuit breakers. While the requirements of ITS. LCO 3.3.9 will somewhat overlap those of ITS LCO 3.3.1, many instrument channels fall under the operability requirements of
several LCOs. Specifying the LCO 3.3.9 applicability as MODES 3, 4, and 5 assures that these instruments will be operable when required.
Palisades Nuclear Plant* Page 1of2 05/30/99
Change SPECIFICATION 3.3.13, POWER MONITORING CHANNELS Discussion ATTACHMENT 6 JUSTIFICATION FOR DEVIATIONS
9. ISTS SR 3.3.13.2 for performance of a CHANNEL FUNCTIONAL TEST is not being proposed since the CHANNEL CHECK and CHANNEL CALIBRATION Frequencies are considered adequate to ensure the OPERABILITY of the equipment. The proposed Frequencies are consistent with the Surveillance Frequencies of the Post Accident Monitoring Instrumentation Frequencies in ITS 3.3. 7, which provides indication-only Functions. Therefore, this change is considered to be consistent with NUREG-1432 for similar type instrumentation functions. This change is also supported by the current licensing basis since no periodic CHANNEL FUNCTIONAL TEST Surveillance is provided in the CTS. A CHANNEL FUNCTIONAL TEST is required just prior to each startup but its purpose is to verify monitoring capability for the startup, not for monitoring during the shutdown conditions. Therefore, the CTS CHANNEL FUNCTIONAL TEST is appropriately addressed in ITS 3.3.1.

10. Not used.
11.
12. Text referring to* other plant designs has been deleted. to make the ITS specific to Palisades.
TSTF-136 is incorporated to reflect the combination of LCO 3.1.1, Shutdown Margin
- Tavg > 200 F and LCO 3.1.2, Shutdown Margin - Tavg < = 200 F (see Section 3.1) .
Palisades Nuclear Plant Page 2 of 2 05/30/99
Change ATTACHMENT6 JUSTIFICATION FOR DEVIATIONS SPECIFICATION 3.3.[10], ESRV INSTRUMENTATION 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.
4.
The following requirements have been renumbered, where applicable, to reflect this deletion. Changes have been made (additions, deletions, and/or changes to the NUREG) to reflect the facility speeific nomenclature, number, reference, system description, or analysis description.
5. This change reflects the current licensing basis/technical specification.

6. An additional Specification is included that is not in NUREG-1432. Proposed ITS 3.3.10, "Engineered Safeguards Room Ventilation (ESRV) Instrumentation,"
contains the CTS 3.16, 3.17, and 4.17 requirements to ensure an assumption of the radiological consequences analysis of the Loss of Coolant Accident (LOCA) is maintained. The analysis results are based on an assumption of automatic isolation of the engineered safeguards pump rooms upon detection of high radiation levels following initiation of the recirculation phase of operation. This is consistent with the current licensing basis as approved in License Amendment No. 31 . Palisades Nuclear Plant Page 1of1 05/30/99
ENCLOSURE 5 CONSUMERS ENERGY COMPANY PALISADES PLANT DOCKET 50-255
CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO JANUARY 6, 1999 NRC COMMENTS CONCERNING ITS SECTION 3.3 SECTION 3.4, REPLACEMENT PAGES
CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO JANUARY 6, 1998 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 No page changes. ATTACHMENT 2 TO ITS CONVERSION SUBMITTAL 83.4.12-12 B 3.4.12-12 05/30/99 Definitions RAI B 3.4.15-5 B 3.4.15-5 05/30/99 Definitions RAI B 3.4.15-6 B 3.4.15-6 05/30/99 editorial ATTACHMENT 3 TO ITS CONVERSION SUBMITTAL
No page changes.

ATTACHMENT 4 TO ITS CONVERSION SUBMITTAL No page changes.
ATTACHMENT 5 TO ITS CONVERSION SUBMITTAL No page changes. ATTACHMENT 6 TO ITS CONVERSION SUBMITTAL No page changes .
LTOP System B 3.4.12
BASES SURVEILLANCE REQUIREMENTS SR 3.4.12.3 (continued)
The 72 hour Frequency considers operating experience with accidental movement of valves having remote control and position indication capabilities available where easily monitored. These considerations include the administrative controls over main control room access and equipment control. SR 3.4.12.4 Perfonnance of a CHANNEL FUNCTIONAL TEST is required every 31 days. A successful test of the required contact(s) of a channel relay may be perfonned by the verification of the change of state of a single contact of the relay. This clarifies what is an acceptable CHANNEL FUNCTIONAL TEST -Of a relay. This is acceptable because all of the other required contacts of the relay are verified by other Technical Specifications and non-Technical Specifications tests at least once per refueling interval with applicable extensions. PORV actuation could depressurize the PCS and is not required. The 31 day Frequency considers
experience with equipment reliability .
A Note has been added indicating this SR is required to be perfonned 12 hours after decreasing any PCS cold leg temperature to< 430°F. This Note allows a discrete period of time to perfonn the required test without delaying entry into the MODE of Applicability for LTOP. This option may be exercised in cases where an unplanned shutdown below 430°F is necessary as a result of a Required Action specifying a plant shutdown, or other plant evolutions requiring an expedited cooldown of the plant. The test must be perfonned within 12 hours after entering the LTOP MODES. SR 3.4.12.5 Perfonnance of a CHANNEL CALIBRATION on each required PORV actuation channel is required every 18 months to adjust the entire channel so that it responds and the valve opens within the required LTOP range and with accuracy to known input. The 18 month Frequency considers operating experience with equipment reliability and is consistent with the typical refueling outage schedule . Palisades Nuclear Plant B 3.4.12-12 05/30/99
PCS Leakage Detection Instrumentation B 3.4.15
BASES ACTIONS (continued)
C.1 If all required monitors are inoperable, no automatic means of monitoring leakage are available and immediate plant shutdown in accordance with LCO 3.0.3 is required. SURVEILLANCE SR 3.4.15.1. SR 3.4.15.2. and SR 3.4.15.3 REQUIREMENTS These SRs require the performance of a CHANNEL CHECK for each required containment sump level indicator, containment atmosphere gaseous activity monitor, and containment atmosphere humidity monitor. The check gives reasonabl~ confidence the channel is operating properly. The Frequency of 12 hours is based on instrument reliability and is reasonable for detecting off normal conditions. SR 3.4.15.4
SR 3.4.15.4 requires the performance of a CHANNEL FUNCTIONAL TEST of the required containment air cooler condensate level switch. Since this instrumentation does not include control room indication of flow rate, a CHANNEL CHECK is not possible. The test ensures that the level switch can perform its function in the desired manner. A successful test of the required contact(s) of a channel relay may be performed by the verification of the change of state of a single contact of the relay. This clarifies what is an acceptable CHANNEL FUNCTIONAL TEST of a relay. This is acceptable because all of the other required contacts of the relay are verified by other Technical Specifications and non-Technical Specifications tests at least once per refueling interval with applicable extensions. The Frequency of 18 months is a typical refueling cycle (performance of the test is only practical during a plant outage) and considers instrument reliability. Operating experience has shown this Frequency is acceptable for detecting degradation .

Palisades Nuclear Plant B 3.4.15-5 05/30/99
PCS Leakage Detection Instrumentation B 3.4.15
BASES SURVEILLANCE REQUIREMENTS SR 3.4.15.5. SR 3.4.15.6. and SR 3.4.15.7 (continued) These SRs require the performance of a CHANNEL CALIBRATION for each required containment sump level, containment atmosphere gaseous activity, and containment atmosphere humidity channel. The calibration verifies the accuracy of the instrument string, including the instruments located inside containment. The Frequency of 18 months is a typical refueling-cycle and considers channel reliability. Operating experience has shown this Frequency is acceptable.
REFERENCES 1. I FSAR, Section 5 .1. 5
2. FSAR, Sections 4.7 and 6.3
Palisades Nuclear Plant B 3.4.15-6 05/30/99
ENCLOSURE 6 CONSUMERS ENERGY COMPANY PALISADES PLANT DOCKET 50-255
CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO JANUARY 6, 1999 NRC COMMENTS CONCERNING ITS SECTION 3.3 SECTION 3.9, REPLACEMENT PAGES
CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS
RESPONSE TO JANUARY 6, 1998 REQUEST FOR ADDITIONAL INFORMATION REVISED PAGES FOR SECTION 3.9
. 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 3.9.2, pg 3.9.2-2 ITS 3.9.2, pg 3.9.2-2 05/30/99 Definitions RAI ATTACHMENT 2 TO ITS CONVERSION SUBMITTAL B 3.9.2-4 B 3.9.2-4 05/30/99 Definitions RAI ATTACHMENT 3 TO ITS CONVERSION SUBMITTAL No page changes.
ATTACHMENT 4 TO ITS CONVERSION SUBMITTAL No page changes.
ATTACHMENT 5 TO ITS CONVERSION SUBMITTAL NUREG 3.9, pg 3.9.3 NUREG 3.9, pg 3.9.3 05/30/99 Definitions RAI ATTACHMENT 6 TO ITS CONVERSION SUBMITTAL JFD 3.9.2, pg 1 of 2 JFD 3.9.2, pg 1 of 2 05/30/99 Definitions RAI
Nuclear Instrumentation 3.9.2
SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.9.2.1 Perform CHANNEL CHECK. 12 hours SR 3.9.2.2 -------------------NOTE--------------------
Neutron detectors are excluded from the CHANNEL CALIBRATION. Perform CHANNEL CALIBRATION. 18 months
Palisades Nuclear Plant 3.9.2-2 Amendment No. 05/30/99 I
J
Nuclear Instrumentation B 3.9.2
BASES SURVEILLANCE REQUIREMENTS (continued) SR 3.9.2.2 is the performance of a CHANNEL CALIBRATION every 18 months. This SR is modified by a Note which states that it is not necessary to calibrate neutron detectors because of the difficulty of simulating a meaningful signal. Wide range and source ranee nuclear instrument channels are not calibrated to indicate the actual power level or the flux in the detector location. The circuitry is adjusted so that wide range and source range readings may be used to determine the approximate reactor flux level for comparitive purposes.
The 18 month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage. Operating experience has shown these components usually pass the Surveillance when performed on the 18 month Frequency.
REFERENCES 1. FSAR, Section 7.6
2. FSAR, Sec ti on 14. 3
Palisades Nuclear Plant B 3.9.2-4 05/30/99
Nuclear Instrumentation 3.9.2
SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.9.2.1 Perform CHANNEL CHECK. 12 hours SR 3.9.2.2 -------------------NOTE--------------------
Neutron detectors are excluded from CHANNEL CALIBRATION. Perform CHANNEL CALIBRATION. j1~months (t) CEOG STS 3.9-3 Rev 1, 04/07/95
Change ATTACHMENT 6 JUSTIFICATION FOR DEVIATIONS SPECIFICATION 3.9.2, NUCLEAR INSTRUMENTATION 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 lmproved 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
4.
facility. - The following requirements have been renumbered, where applicable, to reflect this deletion. 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 initial performance of SR 3. 9 .1.1 within 4 hours of entry into Condition B has been deleted. The accelerated performance of this SR is not warranted based on routine performances of this SR (every 72 hours), and knowledge of
. stable conditions prior to the loss of the source range channel. Secondarily, PCS dilution events are recognizable through other means such as uncontrolled increases in pool water level. This change is consistent with NUREG-1432 as modified by TSTF-96.
7. Not used.
Palisades Nuclear Plant Page 1of2 05/30/99
ENCLOSURE 7 CONSUMERS ENERGY COMPANY PALISADES PLANT DOCKET 50-255
CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO JANUARY 6, 1999 NRC COMMENTS CONCERNING ITS SECTION 3.3 SECTION 3.3, INSTRUMENTATION INSTRUMENT CHANNEL DRIFT MEASUREMENTS
INSTRUMENT CHANNEL DRIFT MEASUREMENTS
As part of the implementation of Improved Technical Specifications (ITS), Palisades is seeking approval to extend the Functional Test interval for the Reactor Protection System (RPS) and Engineered Safety Feature Actuation System (ESFAS) channels from monthly to quarterly. The basis for this extension is provided in the approved NRC topical report CEN-157, "RPS/ESFAS Extended Test Interval Evaluation," as supplemented. This topical report was approved by NRC Safety Evaluation Report (SER) dated November 6, 1989. The SER approved extending the test interval with the following caveat:
"The licensees must confirm that they have reviewed the instrument drift information for each instrument channel involved and have determined that drift occurring in that channel over the period of extended STI will not cause the setpoint value to exceed the allowable .
value as calculated for that channel by their setpoint methodology." The purpose of this letter is to describe the method Palisades used to perform this review and to show that the drift will not cause the setpoint valve to exceed the allowable value based on an extended test interval. Determination of setpoints for RPS/ESFAS at Palisades is based on the methodology provided in ISA Standard S67.04, "Setpoints for Nuclear Safety Related Instrumentation." Starting with the Allowable Value for a trip setting, as listed in the Technical Specifications, instrument loop components accuracy and drift values are statistically combined to establish a trip setpoint which is used to set the trip value in the channel bistables. These setpoints assure that if the channel components function within their accuracy and drift specifications, the channel will trip prior to exceeding the TS Allowable Value. Performance of monthly functional tests and/or calibrations of the instrument loops performed at refueling outage intervals provides continuing assurance that
the instrument loop components operate within specifications during the appropriate interval. As
. some of the bistable components for the RPS and ESFAS differ in their operating principal, different methods of confirming that drift is acceptable is used for each of these systems. The sections below detail the method used to confirm that changing from monthly to quarterly functional testing will not exceed the drift component specified in the setpoint determination calculation.
Reactor Protection System For the RPS, all the analog Bistable Trip Units (BTU) and the same and are interchangeable. The bistables are electronic comparators which utilize a reference voltage as the trip setpoint. A process voltage exceeding the trip setpoint in the appropriate direction trips the bistable deenergizing the output relay. During the monthly functional test, the reference voltage is measured. If the reference voltage as-found is greater than +/-0.025% (+/-0.001V) from the nominal value, it is adjusted back to within this range. Additionally, during the present monthly testing, the process signal is increased or decreased to a value to trip the bistable. As part of this testing, the voltage at which the process trips the bistable is recorded. The methodology for this test uses a built-in test power supply to aid or buck the existing process signal. This method reflects Combustion Engineering's design philosophy that the process signal is never isolated from the RPS during testing. As a result of the methodology, process noise during measurement of the trip point voltage causes difficulty in obtaining an accurate measurement.
1
Bistable drift data for the last 2 years for twelve typical RPS bistables monitoring three trip parameters is shown in the Table 1 attached. The average monthly drift data for the two year period is also shown extrapolated to a quarterly interval (92 days) plus a 25% grace factor or 115 days total. The maximum drift calculated for a 115 day interval based on the measured data was
-0.136%. The Bistable Drift assumed in the RPS setpoint analysis is +/-0.147% Span for the 115 day period. Measured equipment performance during a quarterly test interval is thus shown to be bounded by the drift value assumed in the setpoint calculation.
Figures 1,2 and 3 show data collected for the last two years for the process value which trips the RPS bistable. Figures 1 and 2, for the pressurizer pressure and steam generator pressure bistables re'spectively, show data for channels which have very little process noise. As can be seen by these figures, there is relatively little drift associated with the trip value at which the process input trips the bistable. Figure 3 for the steam generator level channels shows the data obtained for a process signal containing significant process noise. Although the measured trip value fluctuates significantly, the average value for the period can be seen to be flat indicating little channel drift. Engineered Safety Features Actuation System Figure 4 shows the monthly trip data taken for the low primary coolant system pressure channels which provide one of the engineered feature actuations. For these instrument loops, the bistable is an electro/mechanical device. The device consists of an LED and photosensitive driver. The position of a moving arm is changed based on the input signal. When the arm covers the LED, the photosensitive device turns off, changing the state of the output relays. As a result of this design, there is no reference voltage which can be measured and adjusted. Thus, no adjustments are made to the bistable between calibrations. The current/voltage at which the bistable trips is measured and recorded monthly.
The trip voltage data is shown plotted on Figure 4. The data for this channel is taken in a similar manner to that described above for the RPS bistables. A test signal is inserted in parallel with the existing loop signal and the test signal is raised or lowered as appropriate to drive the bistable input toward the trip setting. This method results in the process noise riding on the signal resulting in some of the data scatter. Additional data scatter is results from the repeatability (accounted for in the setpoint methodology) of the electro/mechanical bistable. Review of this data, however, shows that the setpoint remains well above the minimum allowable value derived by our setpoint methodology and that there has been no obvious long term drift of these bistables over the last two and one-half years.
Additional data showing the low drift of these bistables is provided in Figure 5. This data shows the trip setpoint data taken during the last 3 calibration intervals. The data on the time line labeled (F) is the final data whereas the data labeled (NF) is the as-found data. From this data it can be seen that the as-found data taken was always within the allowable final tolerance. Although this calibration data shows the combined drift of both the pressure transmitter and bistable, it shows that the combined drift of all loop components is low during the 18 month calibration interval. Similar data for each of the Engineered Safety Feature Actuation System instrument loops was. reviewed with results similar to that shown for the pressurizer pressure loop. The data as discussed above shows that increasing the functipnal test interval for the Reactor Protection and Engineered Safety Feature Systems from monthly to quarterly will not cause the setpoint value to exceed the allowable value as calculate by the setpoint methodology. 2
TABLE 1 P~S..!...P...!::!a~ra....,m...,,e....,te=r---~ Actual Drift per Month
.wR...... Percent of Span per 115 day Quarter Stm Gen #1 Low Level LA-0751A -0.000136V -0.0128%
LA-07518 -0.000727V -0.0682% LA-0751C -0.00145V -0.136% LA-07510 -0.000364V -0.0341% Stm Gen #2 Low Press PA-0752A O.OV 0.0% PA-07528 -0.000318V -0.0298% PA-0752C -0.000095V -0.0089% PA-07520 0.000227V 0.0213% Pressurizer High Press PA-0102AH 0.000136V 0.0128% PA-0102BH. -0.000045V -0.0043% PA-0102CH 0.000318V 0.0298% PA-0102DH 0.000682V 0.0639% 3
r* -*------------* ------**--*----------*------*----*----c- ----. FIGURE 1 PRESSURIZER PRESSURE 3.98 - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - ------------------- -- --* 0 0 -0 0 0 0 0 0 0- 0 0 0 --0----0 0 --0-----0---0---0-----0----0---0-- -0 3.96
PA-0102AH Q) ~ PA-0102BH O>
cu
._ PA-0102CH ~ 3.94
a. \=:] PA-0102DH I- -0 UPPER LIMIT
-A LOWER LIMIT 3.92 3.9 ~~~-~-~~-~~-~~~--~~___ L__l_______l_~ _L___j___J_ ___ l___ __L_ ____J_ ____ L_ ___ J__ ___ _
APR'97 JUN'97 AUG'97 OCT'97 DEC'97 FEB'98 APR'98 JUN'98 AUG'98 OCT'98 DEC'98 FEB'99 MAY'97 JUL'97 SEP'97 NOV'97 JAN'98 MAR'98 MAY'98 JUL'98 SEP'98 NOV'98 JAN'99 MAR'99 I PA-0102AH Setpoint Adjusted on 10/97, 5/98, and 8/98 : PA-0102BH Setpoint Adjusted on 10/97, 5/98,-8/98-and Tfl98 - PA-0102CH SetpointAdjusted on 4/97, 10/97, 5/98, and 7/98: PA-0102DH SetpointAdjusted on 10/97, 5/98, 7/98 and 1/99
. --*Fl G-OJ~E-*2----- ---* -
STM GEN #2 LOW PRESSURE 2.76 2.74 0----~-~ 0 0 0 0 0-----0--0---~-(T---0~0--~-~---0---0----0
PA-0752A Q) 2.72 ~ PA-07528 O>
ro
._ PA-0752C 0
>c. f~::I PA-07520 I- 2.7 -0 UPPER LIMIT
-8. LOWER LIMIT 2.66 APR'97 JUN'97 AUG'97 NOV'97 JAN'98 MAR'98 JUN'98 AUG'98 OCT'98 DEC'98 MAR'99 MAY'97 JUL'97 SEP'97 DEC'97 FEB'98 APR'98 JUL'98 SEP'98 NOV'98 FEB'99 PA-0752A Setpoint Adjusted on 10/97, 11/97, and 5/98: PA-07528 Setpoint Adjusted on 10/97, 11/97, 9/98, ancf10/98 -----*--- - - --
PA-0752C Setpoint Adjusted on 4/97, 7/97, 10/97, 11/97, 5/98, 9/98, 10/98, 1/99, and 2/99: PA-07520 Setpoint Adjusted on 10/97, 11/97, 5/98, 9/98, an
----------------- *--- *--------------~----------------- -- --- --- *-- --*-
FIGURE 3 STM GEN #1 LOW LVL 2.12 o o o o o o o o o o o o-~--0--~~-4-0-----0---0---0--~-0----0 2.1
LA-0751A Q)

LA-07518 CJ)
~ m 2.08 _. LA-0751C
a. E1 LA-07510
i:::::
I- -0 UPPER LIMIT ls LOWER LIMIT 2.06 2.04 ~~~-~~~-~-1-~-~~ _ _l_ _ _.l_____ L__J_~~~-j__ ___L___j_ _____ L_ __ J_ __ ___J _ _ _ J__ __ _J _ _ _J APR'97 JUN'97 AUG'97 OCT'97 DEC'97 FEB'98 APR'98 JUN'98 AUG'98 OCT'98 DEC'98 FEB'99 MAY'97 JUL'97 SEP'97 NOV'97 JAN'98 MAR'98 MAY'98 JUL'98 SEP'98 NOV'98 JAN'99 MAR'99 LA-0751A Setpoint Adjusted on 10/97, 5/98, and 7/98 : LA-0751 B Setpoint Adjusted on 10/97, 5/98, 7/98, 979~(an-d1079-8 LA-0751C SetpointAdjusted on 10/97, 5/98, 9/98, and 10/98: LA-07510 SetpointAdjusted on 10/97, 2/98, 5/98, and 7/98
VOLTAGE (mV) 10/9/96 11/26/96 12/26/96 en 0
-..J 0
CX> 0 1/21/97 2/14/97 3/11/97 5/13/97 6/11/97 7/17/97 8/12/97 s:: 9/9/97 I 0 10/16/97 r-a; 11/11/97
~ -f c 12/12/97 > 5 fTI. 1/13/98 z c
2/11/98 > 3/11/98 ~ .,,
"'CJ G5 4/14/98 :::cc m :::c 5/28/98 en m ~~
6/17/98
c 7/14/98 N m
8/11/98
c r-0 9/16/98 ~*
10/13/98 "'CJ
c m
11/11/98 en en 12/10/98 c
c 1/12/99 .!!!
2/10/99 3/9/99 i+ t+++ I~ '"CJ lJ lJ lJ 1z ~ 0~ ~ ~ I~
c
~
0 0
~ ~ ~
0 0
~
I\) I\) I\) ! :5:: r0 (') r OJ r r
)> r r r r i5

~
I OJ
m i<

)>
!C 'm ~--------
l FIGURE 5 Rl-3 CALIBRATION DATA(PRESSURIZER LOW PRESSURE) AS FOUND I FINAL 14.6 --~
--+-PIA-0102ALL -PIA-0102BLL PIA-0102CLL ---*"""PIA-0102DLL 14.4 ~Upper Limit
__.._Lower Limit 14.2 I-z w 0::
0::
J 0
14 13.8 13.6 13.4+------------+-------------+-----~-------+-------------,____________-+------------~ 6/21 /1993(F)
6/22/1995(NF) 7/4/1995(F) 11 /13/1996(NF) 11 /13/1996(F) 5/6/1998(NF) .5/6/1998(F)
DATE I ...... _J}}

ML18066A499: Difference between revisions

Revision as of 14:57, 23 February 2020

Text

SUMMARY

3.3.8 APPLICABILITY

Navigation menu

ML18066A499: Difference between revisions

Revision as of 14:57, 23 February 2020

Text

SUMMARY

3.3.8 APPLICABILITY

Navigation menu

Search