Text

Attachment 5: Proposed Technical Specification Bases Changes
	ML18270A326
Person / Time
Site:	Palisades
Issue date:	09/27/2018
From:	; Entergy Nuclear Operations
To:	; Office of Nuclear Reactor Regulation
Shared Package
ML18270A320	List: ML18270A322; ML18270A323; ML18270A324; ML18270A325; ML18270A326; ML18270A327; ML18270A328; PNP 2018-025, License Amendment Request - Application to Revise Technical Specifications to Adopt TSTF-425, Revision 3, Relocate Surveillance Frequencies to Licensee Control- Risk Informed Technical Specification Task Force (RITSTF) Initiative 5b.;
References
	PNP 2018-025
	Download: ML18270A326 (130)
	v • d • e

ATTACHMENT 5 Proposed Technical Specification Bases Changes (for information only)

(showing proposed changes; additions are highlighted and deletions are strikethrough) 129 pages follow

ATTACHMENT 5 PNP 2018-025 Proposed Technical Specification Bases Changes (for information only)

INSERT 3 The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

SDM B 3.1.1 BASES SURVEILLANCE SR 3.1.1.1 (continued)

REQUIREMENTS (continued) f. Isothermal Temperature Coefficient (ITC).

Using the ITC accounts for Doppler reactivity in this calculation because the reactor is subcritical and the fuel temperature will be changing at the same rate as the PCS.

Samarium is not considered in the reactivity analysis since the analysis assumes that the negative reactivity due to Samarium is offset by the positive reactivity of Plutonium built in.

SR 3.1.1.1 requires SDM to be within the limits specified in the COLR.

This SDM value ensures the consequences of an MSLB, will be acceptable as a result of a cooldown of the PCS which adds positive reactivity in the presence of a negative moderator temperature coefficient as well as the other events described in the Applicable Safety Analysis. As such, the requirements of this SR must be met whenever the plant is in MODES 3, 4, and 5.

The Frequency of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months for the verification of SDM is based on the generally slow change in required boron concentration, and also allO'tvs sufficient time for the o~erator to collect the required data, which may inciud&pertorming a boron concentration analysis, and com~leting tiRe ...,

calculation. r--- - - - ' - - - - ,

Insert 3 REFERENCES 1. FSAR, Section 5.1

2. FSAR, Section 14.14

3. FSAR, Section 14.3

4. 10 CFR 50.67

5. FSAR, Section 14.2 Palisades Nuclear Plant B 3.1.1-5 Amendment No. 226 Revised 04/14/2016

Reactivity Balance B 3.1.2 BASES SURVEILLANCE SR 3.1.2.1 REQUIREMENTS Core reactivity is verified by periodic comparisons of measured and predicted PCS boron concentrations. The comparison is made considering that other core conditions are fixed or stable including control rod position, moderator temperature, fuel temperature, fuel depletion, and xenon concentration . The Surveillance is performed prior to entering MODE 1 as an initial check on core conditions and design calculations at BOC. The SR is modified by a Note in the Surveillance column which indicates that if the normalization of predicted core reactivity to the measured value is to occur, it must take place within the first 60 Effective Full Power Days (EFPD) after each refueling . This allows sufficient time for core conditions to reach steady state, but prevents operation for a large fraction of the fuel cycle without establishing a benchmark for the design calculations. The required ~

subsequent Frequency of 31 EFPD following the initial 60 EFPD after entering MODE 1, is acceptable, based on the slow rate of core Insert 3 changes due to fuel depletion and the presence of other indicators (e.g., T"1' etc.) for prompt indication of an imbalance. A second Note, "only required after initial 60 EFPD," is added to the Frequency column to allow this.

REFERENCES 1. FSAR, Section 5.1

2. FSAR, Chapter 14 Palisades Nuclear Plant B 3.1.2-6 Revised 09/09/2003

Control Rod Alignment B 3.1.4 BASES SURVEILLANCE SR 3.1.4.1 REQUIREMENTS Verification that individual control rod positions are within 8 inches of all other control rods in the group at a 12 hOblr ~reElblency allows the operator to detect a control rod that is beginning to deviate from its expected position. The specified ~reElblency takes into accoblnt other control rod position information that is continbloblsly available to the operator in the control room, so that dming control rod movement, ~

deviations can be detected. Also protection can be provided by the control rod deviation alarm. Insert 3 SR 3.1.4.2 OPERABILITY of two control rod position indicator channels is required to determine control rod positions, and thereby ensure compliance with the control rod alignment and insertion limits. Performance of a CHANNEL CHECK on the primary and secondary control rod position indication channels provides confidence in the accuracy of the rod position indication systems. The control rod "full in" and "full out" lights, which correspond to the lower electrical limit and the upper electrical limit respectively, provide an additional means for determining the control rod positions when the control rods are at either their fully inserted or fully withdrawn positions.

The 12 hom ~reElblency takes into consideration other information ~

continbloblsly available to the operator in the control room, so that dblring control rod movement, deviations can be detected, and protection can be Insert 3 provided by the control rod de',iation alarm.

Palisades Nuclear Plant B 3.1.4-11 Revised 07/18/2007

Control Rod Alignment B 3.1.4 BASES SURVEILLANCE SR 3.1.4.3 REQUIREMENTS (continued) Verifying each full-length control rod is trippable would require that each full-length control rod be tripped. In MODES 1 and 2, tripping each full-length control rod would result in radial or axial power tilts, or oscillations. Therefore, individual full-length control rods are exercised every 92 days to provide increased confidence that all full-length control rods continue to be trippable, even if they are not regularly tripped. A movement of 6 inches is adequate to demonstrate motion without exceeding the alignment limit when only one control rod is being moved.

The 92 day Frequency takes into consideration other information available to the operator in the control room and other surveillances being performed more frequently, which add to the determination of OPER,l\BILITY of the control rods. At any time, if a control rod(s) is immovable, a determination of the trippability of the control rod(s) must be made, and appropriate action taken . Condition 3.1.4 D would apply whenever it is discovered that a single full-length control rod cannot be moved by its operator, yet the control rod is still capable of being tripped (or is fully inserted .)~

SR 3.1.4.4 ~

Demonstrating the rod position deviation alarm is OPERABLE verifies the alarm is functional. The 18 month Frequency takes into account other ~

information continuously a'/ailable to the operator in the control room, so that during control rod movement, deviations can be detected. Insert 3 SR 3.1.4.5 Performance of a CHANNEL CALIBRATION of each control rod position indication channel ensures the channel is OPERABLE and capable of indicating control rod position over the entire length of the control rod's travel with the exception of the secondary rod position indicating channel dead band near the bottom of travel. This dead band exists because the control rod drive mechanism housing seismic support prevents operation of the reed switches. Since this Surveillance must be performed when ~

the reactor is shut down, an 18 month Frequency to be coincident with refueling outage was selected. Operating experience has shown that Insert 3 these components usually pass this Surveillance when performed at a Frequency of once every 18 months. Furthermore, the Frequency takes into account other surveillances being performed at shorter Frequencies, which determine the OPERABILITY of the control rod position indicating systems.

Palisades Nuclear Plant B 3.1.4-12 Revised 07/18/2007

Shutdown and Part-Length Rod Group Insertion Limits B 3.1.5 BASES SURVEILLANCE SR 3.1 .5.1 REQUIREMENTS Verification that the shutdown and part-length rod groups are within their insertion limits prior to an approach to criticality ensures that when the reactor is critical, or being taken critical, the shutdown rods will be available to shut down the reactor, and the required SOM will be maintained following a reactor trip. Verification that the part-length rod groups are within their insertion limits ensures that they do not adversely affect power distribution requirements. This SR aR4 rreql:lency ensure that the shutdown and part-length rod groups are withdrawn before the regulating rods are withdrawn during a plant startup.

Since control rod grol:lps are positioned manl:lally by the control room ~

operator, verification of shl:ltdown and part length rod grol:lp position at a rreql:lency of 12 hOl:lrs is adeql:late to ensl:lre that the shl:ltdovm and Insert 3 part length rod grol:lps are within their insertion limits. Also, the 12 hOl:lr rreql:lency takes into accol:lnt other information a'Jailable to the operator in the control room for the pl:lrpose of monitoring the statl:ls of the shl:ltdown and part length rod grol:lps.

REFERENCES 1. FSAR, Section 5.1

2. FSAR, Section 14.2

3. FSAR, Section 14.6 Palisades Nuclear Plant B 3.1.5-7 Revised 07/02/2004

Regulating Rod Group Position Limits B 3.1 .6 BASES SURVEILLANCE SR 3.1.6.1 REQUIREMENTS With the POlL alarm circuit OPERABLE, verification of each regulating rod group position every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is sufficient to detect rod positions that may approach the acceptable limits, and to provide the operator with time to undertake the Required Action(s) should the sequence or insertion limits be found to be exceeded.

The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Frequency also takes into account the indication prO);ide~

by the POlL alarm circuit and other information about rod group positions available to the operator in the control room . Insert 3 SR 3.1 .6.2 SR 3.1.6.3 Demonstrating the CROOS alarm circuit OPERABLE verifies that the CROOS alarm circuit is functional. The 31 day Frequency takes into~

account other Surveillances being performed at shorter Frequencies that identify improper control rod alignment. Insert 3 REFERENCES 1. FSAR, Section 5.1

2. 10 CFR 50.46

3. FSAR, Section 14.16

4. FSAR, Section 14.4 Palisades Nuclear Plant B 3.1.6-9 Revised 07/30/2003

STE 83.1 .7 8ASES ACTIONS 0.1 (continued)

If Required Actions of Condition A, Condition 8, or Condition C cannot be completed within the required Completion Time, PHYSICS TESTS must be suspended within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . Allowing 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months for suspending PHYSICS TESTS allows the operator sufficient time to change any abnormal rod configuration back to within the limits of LCO 3.1.4, LCO 3.1.5, and LCO 3.1.6, or to restore Primary Coolant System (PCS) temperature to within the limits of LCO 3.4.2.

SURVEILLANCE SR 3.1 .7.1 REQUIREMENTS Verifying that THERMAL POWER is :0::; 2% RTP as specified in the PHYSICS TEST procedure and required by the safety analysis, ensures that adequate LHR and DN8 parameter margins are maintained while LCOs are suspended . The 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months F'requensy is suffisient, based on the~

slow rate of pO'tver shange and insreased operational sontrols in plase during PHYSICS TESTS. Insert 3 SR 3.1.7.2 Verifying Tave ~ 500°F during the PHYSICS TEST ensures that Tave remains in an analyzed range while the LCOs are suspended. The ~

1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months F'requensy is suffisient, based on the slo'# rate of shange and insreased operational sontrols in plase during PHYSICS TESTS. Insert 3 SR 3.1.7.3 Verification that ~ 1% shutdown reactivity is available for trip insertion is performed by a reactivity balance calculation, considering the following reactivity effects:

a. PCS boron concentration;

b. Control rod group position;

c. PCS average temperature;

d. Fuel burnup based on gross thermal energy generation; Palisades Nuclear Plant 83.1.7-5 Revised 05/15/2007

STE B 3.1 .7 BASES SURVEILLANCE SR 3.1.7.3 (continued)

REQUIREMENTS

e. Xenon concentration; and

f. Isothermal Temperature Coefficient (ITC).

Using the ITC accounts for Doppler reactivity in this calculation because reactor power is maintained below 2% RTP, and for most of the PHYSIC TESTS below the point of adding heat the fuel temperature will be changing at the same rate as the PCS.

The Freql.lenGY of 24 hOl.lrs is based on the generally slow Ghange in ~

boron GonGentration and on the lO'N probability of an aGGident oGGl.lrring withol.lt the SDM established by LeO a.1.5. Insert 3 REFERENCES 1. 10 CFR 50, Appendix B, Section XI

2. 10 CFR 50.59

3. Regulatory Guide 1.68, Revision 2, August 1978

4. ANSI/ANS-19.6.1-2005, November 29, 2005 Palisades Nuclear Plant B 3.1.7-6 Revised 05/15/2007

LHR B 3.2.1 BASES SURVEILLANCE SR 3.2.1.1 REQUIREMENTS The Incore Alarm portion of the Incore Monitoring System provides continuous monitoring of LHR through the plant computer. The PIDAL computer program is used to generate alarm setpoints for the plant computer that are based on measured margin to allowed LHR. As the incore detectors are read by the plant computer, they are continuously compared to the alarm setpoints. If the Incore Alarm System LHR monitoring function is inoperable, excore detectors or manual recordings of the incore detector read ings may be used to monitor LHR.

Periodically monitoring LHR ensures that the assumptions made in the Safety Analysis are maintained . This SR is modified by a Note that states that the SR is only required to be met when the Incore Alarm System is being used to monitor LHR. The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Frequency is ~

consistent '/lith an SR which is to ge j:)erformed each shift.

Insert 3 SR 3.2.1 .2 Continuous monitoring of the LHR is provided by the Incore Alarm System which provides adequate monitoring of the core power distribution and is capable of verifying that the LHR does not exceed its specified limits.

Performance of this SR verifies the Incore Alarm System can accurately monitor LHR by ensuring the alarm setpoints are based on a measured power distribution. Therefore, they are only applicable when the Incore Alarm System is being used to determine the LHR.

The alarm setpoints must be initially adjusted following each fuel loading prior to operation above 50% RTP, and periodically adjusted every 31 Effective Full Power Days (EFPD) thereafter. /\ 31 EFPD Frequency is consistent with the historical testing frequency of the reactor monitoring system. The SR is modified by a Note which requires the SR to be met only when the Incore Alarm System is being used to determine LHR. ~

~

Palisades Nuclear Plant B 3.2.1-8 Revised 08/06/2004

LHR B 3.2.1 BASES SURVEILLANCE SR 3.2.1.3 REQUIREMENTS (continued) SR 3.2.1.3 requires, prior to initial use of the excore LHR monitoring function, verification that the absolute difference of the measured ASI and the target ASI has been ~ 0.05 for each OPERABLE channel for the last 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months using the previous 24 hourly recorded values.

Performance of this SR verifies that plant conditions are acceptable for the Excore Monitoring System to accurately monitor the LHR (Ref. 5).

The prior to initial use verification identifies that there have been no significant power distribution anomalies while using other monitoring methods, e.g ., the incore detectors, which may affected the ability of the excore detectors to monitor LHR.

The SR is modified by a Note that states that the SR is only required to be met when the Excore Monitoring System is being used to monitor LHR. Failure of this SR prevents the Excore Monitoring System from being considered OPERABLE for monitoring of LHR.

SR 3.2.1.4 SR 3.2.1.4 requires verification that THERMAL POWER is less than or equal to the Allowable Power Level (APL) which is limited to not more than 10% greater than the THERMAL POWER at which the APL was last determined. Performance of this SR also verifies that plant conditions are acceptable for the Excore Monitoring System to accurately monitor the LHR (Ref. 5). The 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months Frequency is based o~

engineering judgement and the need to assure that conditions remain acceptable for use of the Excore Monitoring System to monitor LHR. Insert 3 The SR is modified by a Note that states that the SR is only required to be met when the Excore Monitoring System is being used to monitor LHR. Failure of this SR prevents the Excore Monitoring System from being considered OPERABLE for monitoring of LHR.

Palisades Nuclear Plant B 3.2.1-9 Revised 08/06/2004

LHR B 3.2.1 BASES SURVEILLANCE SR 3.2.1.5 REQUIREMENTS (continued) SR 3.2.1.5 requires verification that the absolute difference of the measured ASI and the target ASI is ;5; 0.05 every hour. This must be verified on at least 3 of the 4, 2 of the 3, or 2 of the 2 OPERABLE channels, whichever is the applicable case. However, any otherwise OPERABLE channel which indicates an absolute difference of> 0.05 must be considered out of limits. Performance of this SR verifies that

== =::~:~=~: ::!:~:rt;;:;;;C:;~~1 plant conditions are acceptable for the Excore Monitoring System to be used to assure LHR is within limits (Ref. 5) . ~:!:~:~~~:

Insert 3 The SR is modified by a Note that states that the SR is only required to be met when the Excore Monitoring System is being used to monitor LHR. Failure of this SR (when using an OPERABLE Excore Monitoring System) is a failure to verify that LHR is within limits and is therefore considered a failure to meet the LCO due to LHR not within limits as determined by the Excore Monitoring System.

SR 3.2.1.6 SR 3.2.1.6 requires verification that the QUADRANT POWER TILT is

5; 0.03. Performance of this SR also verifies that plant conditions are acceptable for the Excore Monitoring System to be used to assure LHR is within limits (Ref. 5). The 24 hOblr Freqblency is based on engineering~

jbldgement and the need to identify adverse trends in these parameters prior to their affecting the ability of the Excore Monitoring System to Insert 3 monitor U=IR The SR is modified by a Note that states that the SR is only required to be met when the Excore Monitoring System is being used to monitor LHR. Failure of this SR (when using an OPERABLE Excore Monitoring System) is a failure to verify that LHR is within limits and is therefore considered a failure to meet the LCO due to LHR not within limits as determined by the Excore Monitoring System.

Palisades Nuclear Plant B 3.2.1-10 Revised 08/06/2004

Radial Peaking B 3.2.2 BASES SURVEILLANCE SR 3.2.2.1 REQUIREMENTS The periodic Surveillance to determine FRT ensures that FR T remains within the range assumed in the analysis throughout the fuel cycle.

Determining FR T using the incore detectors after each fuel loading prior to the reactor exceeding 50% RTP ensures that the core is properly loaded.

Performance of the S~rveillance every 31 Effectilo'e F~II Power Da~~ s (EFPD) ens~res that ~nacceptable changes in FR+ are promptly detected. Insert 3 REFERENCES None Palisades Nuclear Plant B 3.2.2-3 Revised 09/28/2001

Tq B 3.2.3 BASES ACTIONS (continued)

If T q is > 0.15, or if Required Actions and associated Completion Times are not met, THERMAL POWER must be reduced to s 25% RTP. This requirement ensures that the core is operating within its thermal limits and places the core in a conservative condition. Four hours is a reasonable time to reach 25% RTP in an orderly manner and without challenging plant systems.

SURVEILLANCE SR 3.2.3.1 REQUIREMENTS QUADRANT POWER TILT (Tq) is determined from excore detector readings which are calibrated using incore detector measurements (Ref. 1). Calibration factors are determined using incore measurements and an incore analysis computer program (Ref. 2). Each power range channel provides alarms if Tq exceeds its limits. Therefore, with all power range channels OPERABLE, this SR only requires verification that the channel deviation alarms do not indicate an excessive Tq. If the Excore Monitoring System Tq deviation alarm monitoring function is inoperable, excore detector readings or symmetric incore detector readings may be used to monitor Tq at 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months intervals. The 12 hOU~

rrequensy pre'Jents signifisant xenon redistribution between Surveillanses. Insert 3 REFERENCES 1. FSAR, Section 7.6.2.2

2. FSAR, Section 7 .6.2.4 Palisades Nuclear Plant B 3.2.3-3 Revised 09/28/2001

ASI B 3.2.4 BASES SURVEILLANCE SR 3.2.4.1 REQUIREMENTS Verifying that the ASI is within the limits specified in the COLR ensures that the core is not approaching DNB conditions. ASI is determined from excore detector readings which are calibrated using incore detector measurements (Ref. 1). Calibration factors are determined using incore measurements and an incore analysis computer program (Ref. 2). ASI is normally calculated and compared to the alarm setpoints continuously and automatically. Therefore, this SR only requires verification that alarms do not indicate an excessive ASI. If the Excore Monitoring System ASI Alarm function is inoperable, excore detector or incore indications may be used to monitor AS I. A rrequensy that result in an aJ3J3reash to the ,l\Sllimits, l3esause the meshanisms that affest the ASI, sush as xenon reElistril3ution or sontml mEl Elrive ' - - - - - - '

meshanism malfunstions, sause the ASI to shange slowly anEi shoulEi l3e ElissovereEll3erore the limits are exseeEleEi.

REFERENCES 1. FSAR, Section 7.6.2.2

2. FSAR, Section 7.6.2.4 Palisades Nuclear Plant B 3.2.4-3 Amendment No. 189 Revised 08/09/2000

RPS Instrumentation B 3.3.1 BASES SURVEILLANCE SR 3.3.1.1 (continued)

REQUIREMENTS (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 Containment High Pressure and Loss of Load channels are B

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 tW{) random failures in redundant channels in any 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Insert 3 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, which are the Thermal Margin Monitors. These monitors provide input to both the VHPT Function and the TMILP Trip Function. The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Frequency is reasonable based on engineering judgment and plant operating experience. ~

SR3.3.1.3 ~

A GaHy- calibration (heat balance) is performed when THERMAL POWER is ~ 15%. The GaHy- calibration consists of adjusting the "nuclear power calibrate" 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 GaHy- calibration (heat balance) procedure. Performance of the GaHy-calibration ensures that the two inputs to the Q power measurement are indicating accurately with respect to the much more accurate secondary calorimetric calculation.

Palisades Nuclear Plant B 3.3.1-29 Amendment No. 226 Revised 04/14/2016

RPS Instrumentation B 3.3.1 BASES SURVEILLANCE REQUIREMENTS (continued)

SR 3.3.1 .3 (continued)

The FFe~~ency of 24 hO~FS is based on !'llant o!'leFating eX!'leFience and takes into acco~nt indications and alaFms located in the contFOI FOom to detect deviations in channel o~t!'l~ts .

B Insert 3 The Frequency is modified by a Note indicating this Surveillance must be performed within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after THERMAL POWER is;:: 15% RTP.

The secondary calorimetric is inaccurate at lower power levels. The 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months 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 ASI reflects the true core power distribution as determined by the incore detectors. ASI is utilized as an input to the TMILP 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 Tinlet limitation of LCO 3.4.1. An adjustment of the excore channel is necessary only if reactor power is greater than 25% RTP and individual excore channel ASI differs from AXIAL OFFSET, as measured by the incores, outside the bounds of the following table:

Allowed Group 4 Group 4 Reactor Rods ;:: 128" withdrawn Rods <128" withdrawn Power s 100% -0.020 S (AO-ASI) S 0.020 -0.040 S (AO-ASI) S 0.040

< 95 -0.033 S (AO-ASI) S 0.020 -0.053 S (AO-ASI) S 0.040

< 90 -0.046 S (AO-ASI) S 0.020 -0.066 S (AO-ASI) S 0.040

< 85 -0.060 S (AO-ASI) S 0.020 -0.080 S (AO-ASI) S 0.040

< 80 -0.120 S (AO-ASI) S 0.080 -0.140 S (AO-ASI) S 0.100

< 75 -0.120 S (AO-ASI) S 0.080 -0.140 S (AO-ASI) S 0.100

< 70 -0.120 S (AO-ASI) S 0.080 -0.140 S (AO-ASI) S 0.100

< 65 -0.120 S (AO-ASI) S 0.080 -0.140 S (AO-ASI) S 0.100

< 60 -0.160 S (AO-ASI) S 0.120 -0.180 S (AO-ASI) S 0.140

< 55 -0.160 S (AO-ASI) S 0.120 -0.180 S (AO-ASI) S 0.140

< 50 -0.160 S (AO-ASI) S 0.120 -0.180 S (AO-ASI) S 0.140

< 45 -0.160 S (AO-ASI) S 0.120 -0.180 S (AO-ASI) S 0.140

< 40 -0.160 S (AO-ASI) S 0.120 -0.180 S (AO-ASI) S 0.140

< 35 -0.160 S (AO-ASI) S 0.120 -0.180 S (AO-ASI) S 0.140

< 30 -0.160 S (AO-ASI) S 0.120 -0.180 S (AO-ASI) S 0.140

< 25 Below 25% RTP any AO/ASI difference is acceptable Table values determined with a conservative Pvar gamma constant of -9505.

Palisades Nuclear Plant B 3.3.1-30 Amendment No. 226 Revised 04/14/2016

RPS Instrumentation 83.3.1 8ASES SURVEILLANCE SR 3.3.1.4 (continued)

REQUIREMENTS (continued) 8elow 25% RTP any difference between ASI and AXIAL OFFSET is acceptable. A Note indicates the Surveillance is not required to have been performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months 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 B

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months allows time for plant stabilization, data taking, and instrument calibration.

The 31 day Frequensy is adequate, based on operating experiense of the exsore linear amplifiers and the slow burnup of the detestors. The exsore readings are a strong funstion of the power prod used in the Insert 3 peripheral fuel bundles and do not represent an integrated reading aGFOSS the sore. Slow shanges in neutron flux during the fuel sysle san also be detested at this Frequensy.

SR 3.3.1.5 A CHANNEL FUNCTIONAL TEST is performed on each RPS instrument channel, except Loss of Load and High Startup Rate, evePf 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.

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 B

current setpoint analysis.

The Frequensy of 92 days is based on the reliability analysis presented in topisal report CEN 327, "RPS/ESFAS Extended Test Interval Evaluation" (Ref. 5). Insert 3 Palisades Nuclear Plant 83.3.1-31 Amendment No. 226 Revised 04/14/2016

RPS Instrumentation B 3.3.1 BASES SURVEILLANCE SR 3.3.1.6 REQUIREMENTS (continued) A calibration check of the power range excore channels S e OFftl 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 detector, both 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 .

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. Insert 3 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 .

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 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 Band 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 each reactor startup.

Palisades Nuclear Plant B 3.3.1-32 Amendment No. 226 Revised 04/14/2016

RPS Instrumentation B 3.3.1 BASES SURVEILLANCE SR 3.3.1.8 REQUIREMENTS (continued) SR 3.3.1.8 &tAe performaRSe sf a CHANNEL CALIBRATION evePf 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 shall 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, TMILP 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 ifift:. Insert 3 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 Gafty 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).

Palisades Nuclear Plant B 3.3.1-33 Amendment No. 226 Revised 04/14/2016

RPS Logic and Trip Initiation B 3.3.2 BASES SURVEILLANCE SR 3.3.2.1 (continued)

REQUIREMENTS (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 B

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 topioal report CEN 327, "RPS/ESFAS Extended Test Interval Evaluation" (Ref. 5). Insert 3 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

6. 10 CFR 50.67 Palisades Nuclear Plant B 3.3.2-10 Amendment No. 226 Revised 04/14/2016

ESF Instrumentation B 3.3.3 BASES SURVEILLANCE SR 3.3.3.1 (continued)

REQUIREMENTS (continued) Performance of the CHANNEL CHECK once every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months 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. 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 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.

Offscale low current loop channels are verified to be reading at the bottom of the range and not failed downscale.

~~i~~~~'~:I~~~~~~~

~-==:===12~."~

period is extremely low, the e~AN~IEL e~EeK minimizes the chance 1...-_ _ _.....1 of loss of protective function due to failure of redundant channels. +Re Insert 3 e~AN~IEL e~EeK supplements less formal, but more frequent, checks of e~AN~IEL OPERABILITY during normal operational use of displays associated with the LeO required channels.

Palisades Nuclear Plant B 3.3.3-22 Revised 03/20/2008

ESF Instrumentation B 3.3.3 BASES SURVEILLANCE SR 3.3.3.2 REQUIREMENTS (continued) 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 \¥ith 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 channels using an analog test input to each bistable.

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 E)(tended Test Interval Evaluation" (Reference 5). Insert 3 SR 3.3.3.3 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 surveillances. CHANNEL CALIBRATIONS must be performed consistent with the setpoint analysis.

Palisades Nuclear Plant B 3.3.3-23 Revised 03/20/2008

ESF Instrumentation B 3.3.3 BASES SURVEILLANCE SR 3.3.3.3 (continued)

REQUIREMENTS (continued) The as found and as left values must also be recorded and reviewed for B

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 calibration interval for the determination of the magnitude of equipment drift in the setpoint analysis. Insert 3 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 3.3.3-24 Revised 03/20/2008

ESF Logic and Manual Initiation B 3.3.4 BASES ACTIONS C.1 and C.2 (continued)

Condition C is entered when one or more Functions have two Manual Initiation or Actuation Logic channels inoperable for Functions 5 or 6, or when the Required Action and associated Completion Time of Condition A are not met for Functions 5 or 6. 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 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and to MODE 5 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

SURVEILLANCE SR 3.3.4.1 REQUIREMENTS A funstional test of eash SIS actuation i.Q lEIl es'e channel

~must be performed eash 92 days. This test is te-Ge 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 "DBA 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 Frequensy of 92 days is based on plant operating experiense. ~

SR3.3.4.2 ~

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 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 onse per refueling interval with applisable extensions.

Palisades Nuclear Plant B 3.3.4-11 Revised 09/09/2003

ESF Logic and Manual Initiation B 3.3.4 BASES SURVEILLANCE SR 3.3.4.2 (continued)

REQUIREMENTS (continued) 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 J.JA.2 is in agreement with the ~

conclusions of the reliability analysis presented in topical report CEN J27, "RPS/ESFAS Extended Test Interval Evaluation" (Ref. 2). Insert 3 SR 3.3.4.3 A CHANNEL FUNCTIONAL TEST is performed on the manual ESF initiation channels, Actuation Logic channels, and bypass removal channels for specified ESF Functions. 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 Surveillance verifies that the required channels will perform their intended functions when needed.

The 1B 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 Insert 3 performed 'Nith the reactor at power. Operating experience has shown these components usually pass the Surveillance ".'hen performed at a Frequency of once every 1B 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 Revised 09/09/2003

DG - UV Start B 3.3.5 BASES ACTIONS (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 Undervoltage 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 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

'8 Specifications tests at least onse per refueling interval with applisable extensions.

The Frequensy of 18 months is based on the plant sonditions nesessaF1 to perform the test Insert 3 Palisades Nuclear Plant B 3.3.5-5 Revised 11/08/2017

OG - UV Start B 3.3.5 BASES SURVEILLANCE SR 3.3.5.2 REQUIREMENTS (continued) A CHANNEL CALIBRATION performeEl each 1B 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 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.

TheF:'"-eqoonsy efta mor:rths is a.typ~~ r~ng GYcle. Operating ~

eXiJeFl-AC&has showl+thls~equenc~1s 3CCef'ltable, '1-1;--s.e-rt-3---.

REFERENCES 1. 10 CFR 50, Appendix A GOCs 17 and 21

2. FSAR, Section 8.6

3. Analysis EA-ELEC-VOLT-033

4. Analysis EA-ELEC-VOLT-034

5. Analysis EA-ELEC-EOSA-04

6. FSAR, Chapter 14

7. Analysis EA-ELEC-EOSA-03

8. Analysis A-NL-92-111

9. Analysis 0098-0189-CALC-001 Palisades Nuclear Plant B 3.3.5-6 Revised 11/08/2017

Refueling CHR Instrumentation B 3.3.6 BASES SURVEILLANCE SR 3.3.6.1 (continued)

REQUIREMENTS (continued) 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 Frequensy, about onse every shift, is based on operating ~

experiense that demonstrates the rarity of shannel failure . Sinse the probability of two random failures in redundant shannels in any 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Insert 3 period is low, the CFlANNEL CFlECK minimizes the shanse of loss of protestive funstion due to failure of redundant shannels. TAe CFlANNEL CFlECK supplements less formal, but more frequent, shesks of shannel OPERABILITY during normal operational use of the displays assosiated with the LCO required shannels.

SR 3.3.6.2 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 onse per refueling interval with applisable extensions.

The Frequensy of 31 days is based on plant operating experiense "'lith ~

regard to shannel OPERABILITY, whish demonstrates that failure of more than one shannel of a given Funstion in any 31 day interval is a Insert 3 rare event.

Palisades Nuclear Plant B 3.3.6-5 Amendment No. 226 Revised 04/14/2016

Refueling CHR Instrumentation B 3.3.6 BASES SURVEILLANCE SR 3.3.6.3 REQUIREMENTS 8

(continued) 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

!lAth regard to channel OPERABILITY, and is consistent with the testing of other manually actuated functions. Insert 3 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.

No required setpoint is specified because these instruments are not assumed to function by any of the safety analyses.

Tn~Fr~qu~+s ~ased upo: ~he assu:~P~R~f an 18 month ~

calibratioRWerval~R-the setp9intdetermtnatioA: '-1

..... n--;--e....rt-3~

REFERENCES 1. FSAR, Section 7.3

2. FSAR, Section 14.19 Palisades Nuclear Plant B 3.3.6-6 Amendment No. 226 Revised 04/14/2016

PAM Instrumentation B 3.3.7 BASES SURVEILLANCE A Note at the beginning of the Surveillance Requirements specifies that REQUIREMENTS 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 verify 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 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.

B 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

'Nith regard to channel OPERABILITY and drift, ..-"hich demonstrates that failure of more than one channel of a given Function in any 31 day Insert 3 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.

Palisades Nuclear Plant B 3.3.7-11 Revised 11/08/2012

PAM Instrumentation B 3.3.7 BASES SURVEILLANCE SR 3.3.7.2 REQUIREMENTS (continued) A C~ANNEL CALIBRATION is perfoFR=led 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.

8 For the core exit thermocouples, a CHANNEL CALIBRATION is performed by substituting a known voltage for the thermocouple.

The Frequenoy is based upon operating experienoe and oonsistenoy with the typioal industry refueling oyole and is justified by an 18 month oalibration interval for the determination of the magnitude of equipment Insert 3 Gfift,.

REFERENCES 1. FSAR, Appendix 7C, "Regulatory Guide 1.97 Instrumentation"

2. Regulatory Guide 1.97

3. NUREG-0737, Supplement 1 Palisades Nuclear Plant B 3.3.7-12 Revised 11/08/2012

Alternate Shutdown System B 3.3.8 BASES 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 SR 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-1S0 and C-1S0A and locally, as appropriate. Operation of the equipment from the AHSDPs C-1S0 and C-1S0A is not necessary. The Surveillance can be satisfied by performance of a continuity check. This will ensure that if the control B

room becomes inaccessible, the plant can be maintained in MODE 3 from the auxiliary shutdown panel and the local control stations.

The 18 month Frequensy is based on the need to perform this Surveillanse under the sonditions that apply during a plant outage and the potential for an unplanned transient if the Surveillanse were Insert 3 performed with the Feastor at power. Operating m(periense demonstrates that Alternate Shutdown System sontrol shannels seldom fail to pass the Surveillanse when performed at a Frequensy of onse every 18 months.

Palisades Nuclear Plant B 3.3.8-S Revised 02/24/200S

Alternate Shutdown System B 3.3.8 BASES SURVEILLANCE SR 3.3.8.3 REQUIREMENTS (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 Insert 3 performed with the reactor at power.

Operating experience demonstrates that Alternate Shutdown System instrumentation channels seldom fail to pass the Surveillance ',vhen performed at a Frequency of once every 18 months. Therefore, the Frequency ',vas concluded to be acceptable 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, "Other Safety Related Protection, Control, and Display Systems"

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

Palisades Nuclear Plant B 3.3.8-6 Revised 02/24/2005

Neutron Flux Monitoring Channels B 3.3.9 BASES 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 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . 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 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 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.

!::'i~~=.:V+~ ~~~:~~~f.:.~i~

~="'~!:i:::::~:;~~uj Insert 3 period is e>Gremely low, CHANNEL CHECK minimizes the chance of L...-_ _ _....

loss of protective function due to failure of redundant channels.

CHANNEL CHECK supplements less formal , but more frequent, checks of channel OPERABILITY durin§ normal operational use of displays associated with the LCO required channels.

Palisades Nuclear Plant B 3.3.9-4 Amendment No 189 Revised 08/09/2000

Neutron Flux Monitoring Channels B 3.3.9 BASES SURVEILLANCE SR 3.3.9.2 REQUIREMENTS (continued) 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 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.

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 Freque~cy~s the sameas that employed for the same channels in ...,

the1)the~appHcabl~MODES. ~-s....Le-rt-3--'

' I-I REFERENCES 1. 10 CFR 50, Appendix A, GOC 13

2. FSAR,Cha~er7

3. FSAR, Chapter 14 Palisades Nuclear Plant B 3.3.9-5 Amendment No 189 Revised 08/09/2000

ESRV Instrumentation B3.3.10 BASES ACTIONS A.1 (continued)

(continued)

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

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 B

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 Insert 3 of channel OPERABILITY during normal operational use of the displays associated with the LCO required channels.

Palisades Nuclear Plant B 3.3.10-3 Amendment No. 226 Revised 04/14/2016

ESRV Instrumentation B 3.3.10 BASES 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 Freql::lensy of 31 days is based on plant operating experiense WithB regard to shannel OPERABILITY, whish demonstrates that faill::lre of more than one shannel of a given Fl::lnstion in any 31 day interval is a Insert 3 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 Freql::lensy is based I::Ipon the assl::lmption of an 18 month B salibration interval for the determination of the magnitl::lde of eql::lipment drift in the setpoint analysis. Insert 3 REFERENCES 1. FSAR, Section 7.4.5.2

2. FSAR, Section 14.22 Palisades Nuclear Plant B 3.3.10-4 Amendment No. 226 Revised 04/14/2016

PCS Pressure, Temperature, and Flow DNB Limits B 3.4.1 BASES ACTIONS Pressurizer pressure and cold leg temperature are controllable and measurable parameters. PCS flow rate is not a controllable parameter and is not expected to vary during steady state operation. With any of these parameters not within the LCO limits, action must be taken to restore the parameter.

The 2-hour Completion Time for restoration of the parameters provides sufficient time to adjust plant parameters, to determine the cause of the off normal condition, and to restore the readings within limits.

The Completion Time is based on plant operating experience.

If Required Action A.1 is not met within the associated 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 2 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . In MODE 2, the reduced power condition eliminates the potential for violation of the accident analysis bounds.

Six hours is a reasonable time that permits the plant power to be reduced at an orderly rate without challenging plant systems.

SURVEILLANCE SR 3.4.1.1 and SR 3.4.1.2 B

REQUIREMENTS The Surveillance for monitoring pressurizer pressure and PCS cold leg temperature is performed using installed instrumentation. The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months interval has been shown by operating practice to be sufficient to regularly assess for potential degradation and verify operation is within Insert 3 safety analysis assumptions.

SR 3.4.1.3 Measurement of PCS total flow rate verifies that the actual PCS flow B

rate is within the bounds of the analyses. This verification may be performed by a calorimetric heat balance or other method.

The Frequency of 18 months reflects the importance of verifying flow after a refueling outage where the core has been altered, which may have caused an alteration of flow resistance. PCS flow rate must also Insert 3 be verified after plugging of each 10 or more steam generator tubes since plugging 10 or more tubes could result in an increase in PCS flow resistance. Plugging less than 10 steam generator tubes will not have a significant impact on PCS flow resistance and, as such, does not require a verification of PCS flow rate.

Palisades Nuclear Plant B 3.4.1-4 Revised 08/24/2004

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

REQUIREMENTS The SR is modified by a Note that states the SR is only required to be performed 31 EFPD after THERMAL POWER is ~ 90% RTP. The Note is necessary to allow measurement of the flow rate at normal operating conditions at power in MODE 1. The most common , and perhaps accurate, method used to perform the PCS total flow surveillance is by means of a primary to secondary heat balance (calorimetric) with the plant at or near full rated power. The most accurate results for such a test are obtained with the plant at or near full power when differential temperatures measured across the reactor are the greatest.

Consequently, the test should not be performed until reaching near full power (i.e., ~ 90% RTP) conditions. Similarly, test accuracy is also influenced by plant stability. In order for accurate results to be obtained, steady state plant conditions must exist to permit meaningful data to be gathered during the test. Typically, following an extended shutdown the secondary side of the plant will take up to several days to stabilize after power escalation. It is impracticable to perform a primary to secondary heat balance of the precision required for the PCS flow measurement until stabilization has been achieved. Furthermore, an integral part of the PCS flow heat balance involves the use of Ultrasonic Flow Measurement equipment for measuring steam generator feedwater flow. This equipment requires, stable plant operation at or near full power conditions before it can be used. As such, the Surveillance cannot be performed in MODE 2 or below, and will not yield accurate results if performed below 90% RTP.

REFERENCES 1. FSAR, Section 14.1 Palisades Nuclear Plant B 3.4.1-5 Revised 08/24/2004

PCS Minimum Temperature for Criticality B 3.4.2 BASES APPLICABILITY The reactor has been designed and analyzed to be critical in MODES 1 and 2 only and in accordance with this specification. Criticality is not permitted in any other MODE. Therefore, this LCO is applicable in MODE 1, and MODE 2 when Keft ~ 1.0.

ACTIONS If Tave is below 525°F and cannot be restored in 30 minutes, 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 MODE 2 with Keft < 1.0 within 30 minutes. Rapid reactor shutdown can be readily and practically achieved within a 30 minute period. The allowed time reflects the ability to perform this action and to maintain the plant within the analyzed range .

SURVEILLANCE SR 3.4.2.1 B

REQUIREMENTS PCS loop average temperature is required to be verified at or above 525°F every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . The SR to verify PCS loop average temperature every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months takes into aGGount indiGations and alarms that are Gontinuously available to the operator in the Gontrol room . Insert 3 REFERENCES 1. FSAR, Section 14.1.3 Palisades Nuclear Plant B 3.4.2-2 Amendment No. 189

PCS PIT Limits B 3.4.3 BASES SURVEILLANCE SR 3.4.3.1 REQUIREMENTS Verification that operation is within the limits of Figure 3.4.3-1 and Figure 3.4.3-2 is required every 30 min",tes when PCS pressure and temperature conditions are undergoing planned changes. +Ais Freq",enoy is oonsidered reasonable in view of the oontrol room indioation a~'ailable to monitor pes stat",s. Also, sinoe temperat",re rate of ohange limits are speoified in ho",rly inorements, 30 min",tes permits assessment and oorreotion for minor deviations within a reasonable time. Calculation of the average hourly cool down rate must consider changes in reactor vessel inlet temperature caused by initiating shutdown cooling, by starting primary coolant pumps with a temperature difference between the steam generator and PCS, or by stopping primary coolant pumps with shutdown cooling in service. The additional restrictions in Figure 3.4.3-2, required for the reactor vessel head nozzle repairs, use the average core exit temperature to provide the best indication available of the temperature of the head inside material temperature. This indication may be either the average of the core exit thermocouples or the vessel outlet temperature.

Surveillance for heatup and cooldown operations may be discontinued ~

when the definition given in the relevant plant procedure for ending the ~

activity is satisfied.

This SR is modified by a Note that requires this SR be performed only during PCS heatup and cooldown operations. No SR is given for criticality operations because LCO 3.4.2 contains a more restrictive requirement.

REFERENCES 1. Safety Evaluation for Palisades Nuclear Plant License Amendment No. 245, dated January 19, 2012

2. 10 CFR 50, Appendix G

3. Deleted

4. ASTM E 185-82, July 1982

5. 10 CFR 50, Appendix H

6. ASME, Boiler and Pressure Vessel Code,Section XI, Appendix E

7. Safety Evaluation for Palisades Nuclear Plant License Amendment No. 218, dated November 8, 2004 Palisades Nuclear Plant B 3.4.3-8 Revised 02/17/2012

PCS Loops - MODES 1 and 2 B3.4.4 BASES SURVEILLANCE SR 3.4.4.1 REQUIREMENTS This SR requires verification every 12 hOllrs of the required number of loops in operation. Verification may include indication of PCS flow, temperature, or pump status, which help to ensure that forced flow is providing heat removal while maintaining the margin to DNB. +Ae Freqllensy of 12 hOllrs has been sho'.'lA by operating prastise to be ,...---....----.

sllffisient to reglliarly assess degradation and verify operation vJithin REFERENCES 1. FSAR, Section 14.1 Palisades Nuclear Plant B 3.4.4-4 Revised 09/21/2006

PCS Loops - MODE 3 B 3.4.5 BASES (continued)

SURVEILLANCE SR 3.4.5.1 REQUIREMENTS SR 3.4.5.2 This SR requires verification every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months that the secondary side B

water level in each SG is 2: -84% using the wide range level instrumentation. An adequate SG water level is required in order to have a heat sink for removal of the core decay heat from the primary coolant.

The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months interval has been shown by operating practice to be sufficient to regularly assess degradation and verify operation within the safety analyses assumptions. Insert 3 SR 3.4.5.3 Verification that the required PCP is OPERABLE ensures that the single failure criterion is met and that an additional PCS loop can be placed in operation, if needed, to maintain decay heat removal and primary coolant circulation. Verification is performed by verifying proper breaker alignment and power availability to the required PCP that is not in B

operation such that the PCP is capable of being started and providing forced PCS flow if needed . Proper breaker alignment and power availability means the breaker for the required PCP is racked-in and electrical power is available to energize the PCP motor. The Frequency of 7 days is considered reasonable in view of other administrative controls available and has been shown to be acceptable by operating experience. Insert 3 REFERENCES None Palisades Nuclear Plant B 3.4.5-5 Revised 09/21/2006

PCS Loops - MODE 4 B 3.4.6 BASES ACTIONS C.1! C.2.1! and C.2.2 (continued)

If no PCS loops or SDC trains are OPERABLE, or no PCS loop is in operation and the SDC flow through the reactor core is < 2810 gpm, except during conditions permitted by Note 1 in the LCO section, all operations involving reduction of PCS boron concentration must be suspended. Action to restore one PCS loop or SDC train to OPERABLE status and operation shall be initiated immediately and continue until one loop or train is restored to operation and flow through the reactor core is restored to ~ 2810 gpm. Boron dilution requires forced circulation for proper mixing, and the margin to criticality must not be reduced in this type of operation. The immediate Completion Times reflect the importance of decay heat removal.

SURVEILLANCE SR 3.4.6.1 REQUIREMENTS This SR requires verification every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months that one required loop or train is in operation. This ensures forced flow is providing heat removal and mixing of the soluble boric acid. Verification may include flow rate (SDC only), or indication of flow, temperature, or pump status for the PCP. The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Frequency has been shown by operating practice to ~

be sufficient to regularly assess PCS loop/SOC train status. In addition, control room indication and alarms will normally indicate loop/train status. Insert 3 SR 3.4.6.2 This SR requires verification every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months of secondary side water level in the required SG(s) ~ -84% using the wide range level instrumentation.

An adequate SG water level is required in order to have a heat sink for removal of the core decay heat from the primary coolant. The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ~

interval has been sho¥m by operating practice to be sufficient to regularly assess degradation and verify SG status. Insert 3 Palisades Nuclear Plant B 3.4.6-5 Revised 07/31/2007

PCS Loops - MODE 4 B 3.4.6 BASES SURVEILLANCE SR 3.4.6.3 REQUIREMENTS (continued) Verification that the required pump is OPERABLE ensures that an additional PCS loop or SOC train can be placed in operation, if needed to maintain decay heat removal and primary coolant circulation. Verification is performed by verifying proper breaker alignment and power available to the required pump that is not in operation such that the pump is capable of being started and providing forced PCS flow if needed. Proper breaker alignment and power availability means the breaker for the required pump is racked-in and electrical power is available to energize the pump motor.

The Frequency of 7 days is considered reasonable in view of other ~

administrative controls al"lailable and has been shown to be acceptable by operating experience. Insert 3 REFERENCES None Palisades Nuclear Plant B 3.4.6-6 Revised 07/31/2007

PCS Loops - MOOE 5, Loops Filled B 3.4.7 BASES (continued)

ACTIONS A.1 and A.2 If one SOC train is inoperable and any SG has a secondary side water level < -84% (refer to LCO Bases section), redundancy for heat removal is lost. Action must be initiated immediately to restore a second SOC train to OPERABLE status or to restore the water level in the required SGs. Either Required Action A.1 or Required Action A.2 will restore redundant decay heat removal paths. The immediate Completion Times reflect the importance of maintaining the availability of two paths for decay heat removal.

B.1 and B.2 If no SOC trains are OPERABLE or SOC flow through the reactor core is

< 2810 gpm, except as permitted in Note 1, all operations involving the reduction of PCS boron concentration must be suspended. Action to restore one SOC train to OPERABLE status and operation shall be initiated immediately and continue until one train is restored to operation and flow through the reactor core is restored to ~ 2810 gpm . Boron dilution requires forced circulation for proper mixing and the margin to criticality must not be reduced in this type of operation. The immediate Completion Times reflect the importance of maintaining operation for decay heat removal.

SURVEILLANCE SR 3.4.7.1 REQUIREMENTS This SR requires verification every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months that one SOC train is in operation. Verification of the required flow rate ensures forced flow is providing heat removal and mixing of the soluble boric acid. The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ~

Frequenoy has been shown by operating praotioe to be suffioient to regularly assess SDC train status. In addition, oontrol room indioation Insert 3 and alarms will normally indioate train status.

Palisades Nuclear Plant B 3.4.7-6 Revised 07/31/2007

PCS Loops - MODE 5, Loops Filled B 3.4.7 BASES SURVEILLANCE SR 3.4.7.2 REQUIREMENTS (continued) This SR requires verification every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months of secondary side water level in the required SGs ~ -84% using the wide range level instrumentation.

An adequate SG water level is required in order to have a heat sink for B

removal of the core decay heat from the primary coolant. The Surveillance is required to be performed when the LCO requirement is being met by use of the SGs. If both SDC trains are OPERABLE, this SR is not needed. The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Frequency has been shown by operating practice to be sufficient to regularly assess degradation and verify SG staWs,. Insert 3 SR 3.4.7.3 Verification that the second SDC train is OPERABLE ensures that redundant paths for decay heat removal are available. The requirement also ensures that the additional train can be placed in operation, if needed, to maintain decay heat removal and primary coolant circulation.

Verification is performed by verifying proper breaker alignment and power available to the required pump that is not in operation such that the SDC pump is capable of being started and providing forced PCS flow if needed. Proper breaker alignment and power availability means the 8

breaker for the required SDC pump is racked-in and electrical power is available to energize the SDC pump motor. The Surveillance is required to be performed when the LCO requirement is being met by one of two SDC trains, e.g., both SGs have < -84% water level. The Frequency of 7 days is considered reasonable in vie..., of other administrative controls available and has been shown to be acceptable by operating o)Eperience. Insert 3 REFERENCES 1. NRC Information Notice 95-35, "Degraded Ability of Steam Generators to Remove Decay Heat by Natural Circulation" Palisades Nuclear Plant B 3.4.7-7 Revised 07/31/2007

PCS Loops - MOOE 5, Loops Not Filled B 3.4.8 BASES ACTIONS B.1 and B.2 (continued)

If no SOC trains are OPERABLE or SOC flow through the reactor core is not within limits, except as provided in Note 1, all operations involving the reduction of PCS boron concentration must be suspended. Action to restore one SOC train to OPERABLE status and operation shall be initiated immediately and continue until one train is restored to operation and flow through the reactor core is restored to within limits. Boron dilution requires forced circulation for proper mixing and the margin to criticality must not be reduced in this type of operation. The immediate Completion Time reflects the importance of maintaining operation for decay heat removal.

SURVEILLANCE SR 3.4.8.1 and SR 3.4.8.2 REQUIREMENTS These SRs require verification every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months that one SOC train is in operation. Verification of the required flow rate ensures forced circulation is providing heat removal and mixing of the soluble boric acid . +he Insert 3 L...-_ _ _- '

SR 3.4.8.1 and SR 3.4.8.2 are each modified by a Note to indicate the SR is only required to be met when complying with the applicable portion of the LCO. Therefore, it is only necessary to perform either SR 3.4.8.1, or SR 3.4.8.2 based on the method of compliance with the LCO.

Palisades Nuclear Plant B 3.4.8-4 Revised 07/31/2007

PCS Loops - MODE 5, Loops Not Filled B 3.4.8 BASES SURVEILLANCE SR 3.4.8.3 REQUIREMENTS (continued) This SR requires verification every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months that two of the three charging pumps are incapable of reducing the boron concentration in the PCS below the minimum value necessary to maintain the required SHUTDOWN MARGIN. Making the charging pumps incapable reducing the boron concentration in the PCS may be accomplished by electrically disabling the pump motors, blocking potential dilution sources to the pump suction, or by isolating the pumps discharge flow path to the PCS.

B Verification may include visual inspection of the pumps configuration (e.g., pump breaker position or valve alignment), or the use of other administrative controls. The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Frequensy is based on engineering judgement sonsidering ol'lerating I'lrastise, administrative sontrol a'/ailable, and the unlikeness of inadvertently aligning a sharging I'luml'l Insert 3 for pes injestion during this I'leriod.

SR 3.4.8.3 is modified by a Note to indicate the SR is only required to be met when complying with LCO 3.4.8.b. When SDC flow through the reactor core is ~ 2810 gpm, there is no restriction on charging pump operation .

SR 3.4.8.4 Verification that the required number of trains are OPERABLE ensures that redundant paths for heat removal are available and that additional trains can be placed in operation, if needed, to maintain decay heat removal and primary coolant circulation. Verification is performed by verifying proper breaker alignment and indicated power available to the required pump that is not in operation such that the SDC pump is capable of being started and providing forced PCS flow if needed. Proper breaker B

alignment and power availability means the breaker for the required SDC pump is racked-in and electrical power is available to energize the SDC pump motor. The Frequensy of 7 days is sonsidered reasonable in vie'N of other administrative sontrols available and has been shown to be assel'ltable by ol'lerating eXl'leriense. Insert 3 REFERENCES None Palisades Nuclear Plant B 3.4.8-5 Revised 07/31/2007

Pressurizer B 3.4.9 BASES ACTIONS D.1 and D.2 (continued)

If one or more of the electrical buses' required pressurizer heaters cannot be restored to an OPERABLE status within the associated allowed Completion Times, 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 MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and to MODE 4 within 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . The Completion Time of 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months is reasonable, based on operating experience, to reach MODE 3 from full power in an orderly manner and without challenging safety systems. Similarly, the Completion Time of 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months is reasonable, based on operating experience, to reach MODE 4 from full power in an orderly manner and without challenging plant systems.

SURVEILLANCE SR 3.4.9.1 REQUIREMENTS This SR ensures that during steady state operation, pressurizer water level is maintained below the nominal upper limit to provide a minimum space for a steam bubble. The Surveillance is performed by observing the indicated level. SR 3.4.9.1 is modified by a Note which states that verification of the pressurizer water level is not required to be met until 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months after a bubble has been established in the pressurizer and the pressurizer water level has been lowered to its normal operating band.

The intent of this Note is to prevent an SR 3.0.4 conflict by delaying the performance of this SR until after the water level in the pressurizer is

=::::1==;:;=:: ~=::::==:~~:=~'--II-ns...L-ert-3--'

within its normal operating band following a plant heatup. ~:~~ ~

analyses assumptions. Alarms are also available for early detection of abnormal level indications.

SR 3.4.9.2 The SR is satisfied when the power supplies are demonstrated to be capable of producing the minimum power and the capacity of the associated pressurizer heaters are verified to be ~ 375 kW. (This may be done by testing the power supply output and by performing an electrical Insert 3 check on heater element continuity and resistance.) The Frequency of 18 ~

months is considered adequate to detect heater degradation and has been shown by operating experience to be acceptable.

Palisades Nuclear Plant B 3.4.9-6 Amendment No. 256 Revised 07/29/2015

Pressurizer B 3.4.9 BASES SURVEILLANCE SR 3.4.9.3 REQUIREMENTS (continued) This SR only applies to the pressurizer heaters normally powered from electrical bus 1E since the pressurizer heaters powered from bus 1Dare permanently connected to the engineered safeguards electrical system.

This SR confirms that the pressurizer heaters normally fed from electrical bus 1E are capable of being powered from electrical bus 1C by use of a B

jumper cable. It is not the intent of this SR to physically install the jumper cable, but to verify the necessary components are available for installation and to ensure the procedures and methods used to install the jumper cable are current. The Frequency of 18 months is based on engineering judgement and is considered acceptable when considering the design reliability of the equipment (the jumper cable is left Insert 3 permanently in place and dedicated to providing the emergency feed function only), and administrative control which govern configuration management and changes to plant procedures.

REFERENCES 1. FSAR, Chapter 14

2. FSAR, Section 4.3.7

3. WCAP-16125-NP-A, "Justification for Risk-Informed Modifications to Selected Technical Specifications for Conditions Leading to Exigent Plant Shutdown," Revision 2, August 2010.

Palisades Nuclear Plant B 3.4.9-7 Amendment No. 256 Revised 07/29/2015

Pressurizer PORVs B3.4.11 BASES SURVEILLANCE SR 3.4.11 .1 REQUIREMENTS Block valve cycling verifies that it can be opened and closed if necessary.

The basis for the Frequency of "prior to entering MODE 4 from MODE 5 if not performed in the previous 92 days" reflects the importance of not routinely cycling the block valves during the period when the PCS is pressurized since this practice may result in the associated PORV being opened by the increase inlet pressure to the PORV. The "92 days" portion of the Frequency is consistent with the testing frequency stipulated by ASME Section XI as modified by the Cold Shutdown Testing Basis used in support of the second 120 month interval of the Inservice Valve Testing Program which only requires the block valves to be cycled during Cold Shutdown conditions. If the block valve is closed to isolate a PORV that is capable of being manually cycled, the OPERABILITY of the block valve is of importance because opening the block valve is necessary to permit the PORV to be used for manual control of primary coolant pressure. If a block valve is open and its associated PORV was stuck open, the OPERABILITY of the block valve is of importance because closing the block valve is necessary to isolate the stuck opened PORV.

SR 3.4.11.2 SR 3.4.11 .2 requires complete cycling of each PORV. PORV cycling demonstrates its function and is performed when the PCS temperature is

> 200°F. Stroke testing of the PORVs above 200°F is desirable since it B

closer simulates the temperature and pressure environmental effects on the valves and thus represents a better test condition for assessing PORV performance under normal plant conditions. The Frequency Of 18 months is based on a typical refueling cycle and industry accepted practice. Insert 3 REFERENCES None Palisades Nuclear Plant B 3.4.11-7 Revised 02/24/2005

LTOP System B 3.4.12 BASES ACTIONS 0 .1 (continued)

If two required PORVs are inoperable, or if the Required Actions and the associated Completion Times are not met, or if the LTOP System is inoperable for any reason other than Condition A, B, or C, the PCS must be depressurized and a vent established within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . The vent must be sized to provide a relieving capability of ;: >: 167 gpm at a pressure of 315 psia which ensures the flow capacity is greater than that required for the worst case mass injection transient reasonable during the applicable MODES. This action protects the PCPB from a low temperature overpressure event and a possible brittle failure of the reactor vessel.

The Completion Time of 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months to depressurize and vent the PCS is based on the time required to place the plant in this condition and the relatively low probability of an overpressure event during this time period due to operator attention and administrative requirements.

SURVEILLANCE SR 3.4.12.1 REQUIREMENTS To minimize the potential for a low temperature overpressure event by limiting the mass injection capability, both HPSI pumps are verified to be incapable of injecting into the PCS. The HPSI pumps are rendered incapable of injecting into the PCS by means that assure that a single event cannot cause overpressurization of the PCS due to operation of the pump. Typical methods for accomplishing this are by pulling the HPSI pump breaker control power fuses, racking out the HPSI pump motor circuit breaker, or closing the manual discharge valve.

SR 3.4.12.1 is modified by a Note which only requires the SR to be met when complying with LCO 3.4.12.a. When all PCS cold leg temperature are;::>: 300°F, a start of both HPSI pumps in conjunction with a charging/letdown imbalance will not cause the PCS pressure to exceed the 10 CFR 50 Appendix G limits. Thus, this SR is only required when any PCS cold leg temperature is reduced to less than 300°F.

The 12 hOl:lr interval sonsiaers oJ3eratin~ J3rastise to re~l:Ilarly assess ~

J30tential ae~raaation ana to verify oJ3eration within the safety anaIYSis . ~

Palisades Nuclear Plant B 3.4.12-10 Revised 02/17/2012

LTOP System B 3.4.12 BASES SURVEILLANCE SR 3.4.12.2 REQUIREMENTS (continued) SR 3.4.12.2 requires a verification that the required PCS vent, capable of relieving ~ 167 gpm at a PCS pressure of 315 psia, is OPERABLE by verifying its open condition eitAeF.

a. Once every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months for a valve that is not locked open; or

b. Once every 31 days for a valve that is locked open.

The passive vent arrangement must only be open to be OPERABLE.

This Surveillance need only be performed if vent valves are being used to satisfy the requirements of this LCO. This Surveillance does not need to be performed for vent paths relying on the removal of a steam generator primary manway cover, pressurizer manway cover, safety valve or PORV since their position is adequately addressed using administrative controls and the inadvertent reinstallation of these components is unlikely. The Frequencies consider operating experience with mispositioning of unlocked and locked vent valves, respectively.

SR 3.4.12.3 The PORV block valve must be verified open every 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months to provide the flow path for each required PORV to perform its function when actuated . The valve can be remotely verified open in the main control room .

The block valve is a remotely controlled, motor operated valve. The power to the valve motor operator is not required to be removed, and the manual actuator is not required locked in the inactive position. Thus, the block valve can be closed in the event the PORV develops excessive B

leakage or does not close (sticks open) after relieving an overpressure event.

The 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Frequency considers operating experience with accidental movement of valves having remote control and position indication capabilities available where easily monitored. These considerations Insert 3 include the administrative controls over main control room access and equipment control.

Palisades Nuclear Plant B 3.4.12-11 Revised 02/17/2012

LTOP System B 3.4.12 BASES SURVEILLANCE SR 3.4.12.4 REQUIREMENTS (continued) Performance of a CHANNEL FUNCTIONAL TEST is required every 31 days. A successful FU 0 ES 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 Technical Specifications and non Technical Specifications tests at least once per refueling interval 'Ilith 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 performed 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months 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 Action specifying a plant shutdown, or other plant evolutions requiring an expedited cooldown of the plant.

The test must be performed within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after entering the LTOP MODES.

SR 3.4.12.5 Performance 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.

Insert 3 Palisades Nuclear Plant B 3.4.12-12 Revised 02/17/2012

PCS Operational LEAKAGE B 3.4.13 BASES SURVEILLANCE SR 3.4.13.1 (continued)

REQUIREMENTS (continued) Note 2 states that this SR is not applicable to primary to secondary LEAKAGE because LEAKAGE of 150 gallons per day cannot be measured accurately by an RCS water inventory balance.

The 72 hOloJr rreqloJency is a reasonable interval to trend LEAKAGE and recognizes the importance of early leakage detection in the prevention of accidents. A Note under the Frequency column states that this SR is required to be performed during steady state operation.~

SR 3.4.13.2 ~

This SR verifies that primary to secondary LEAKAGE is less or equal to 150 gallons per day through anyone SG. Satisfying the primary to secondary LEAKAGE limit ensures that the operational LEAKAGE performance criterion in the Steam Generator Program is met. If this SR is not met, compliance with LCO 3.4.17, "Steam Generator Tube Integrity," should be evaluated. The 150 gallons per day limit is measured at room temperature as described in Reference 7. The operational LEAKAGE rate limit applies to LEAKAGE through anyone SG. If it is not practical to assign the LEAKAGE to an individual SG, all the primary to secondary LEAKAGE should be conservatively assumed to be from one SG.

The Surveillance is modified by a Note which states that the Surveillance is not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after establishment of steady state operation. For RCS primary to secondary LEAKAGE determination, steady state is defined as stable RCS pressure, temperature, power level, pressurizer and makeup tank levels, makeup and letdown, and RCP seal injection and return flows .

The SloJrveiliance rreqloJency of 72 hOloJrs is a reasonable interval to trend primary to secondary LEAKAGE and recognizes the importance of early leakage detection in the prevention of accidents. The primary to secondary LEAKAGE is determined using continuous process radiation monitors or radiochemical grab sampling in accordance with the EPRI guidelines (Ref. 7). ~

~

Palisades Nuclear Plant B 3.4.13-6 Amendment No. 226 Revised 04/14/2016

PCS PIV Leakage B 3.4.14 BASES ACTIONS .QJ.

(continued)

The inoperability of the SOC suction valve interlocks renders the SOC suction isolation valves incapable of preventing an inadvertent opening of the valves at PCS pressures in excess of the SOC systems design pressure. If the SOC suction valve interlocks are inoperable, operation may continue as long as the suction penetration is closed by at least one closed deactivated valve within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . This action accomplishes the purpose of the interlock. The 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months Completion Time provides time to accomplish the action and restricts operation with an inoperable interlock.

SURVEILLANCE SR 3.4.14.1 REQUIREMENTS Performance of leakage testing on each PCS PIV or isolation valve used to satisfy Required Action A.1 is required to verify that leakage is below the specified limit and to identify each leaking valve. The leakage limit of up to 5 gpm maximum applies to each valve. Leakage testing requires a stable pressure condition.

For the two PIVs in series, the leakage requirement applies to each valve individually and not to the combined leakage across both valves.

If the PIVs are not individually leakage tested, one valve may have failed completely and not be detected if the other valve in series meets the leakage requirement. In this situation, the protection provided by redundant valves would be lost.

Testing is to be performed every 9 months whenever the plant has been in MODE 5 for 7 days or more, but may be extended up to a maximum of 18 months, a typical refueling cycle, if the plant does not go into MODE 5 for at least 7 days. The 18 month Frequency is ~

consistent '.vith 10 CFR 50.55a(f), as contained in the INSeRVICe TeSTI~IG PROGRAM , and is within the frequency allowed by the Insert 3 American Society of Mechanical engineers (ASMe) Code (Ref. a), and is based on the need to perform the Surveillance under conditions that apply during a plant outage and the potential for an unplanned transient if the SUF\'eiliance '.vere performed with the reactor at power.

The leakage limit is to be met at the PCS pressure associated with MODES 1 and 2. This permits leakage testing at high differential pressures with stable conditions not possible in the MODES with lower pressures.

Palisades Nuclear Plant B 3.4.14-5 Amendment No. 262 Revised 07/26/2017

PCS PIV Leakage B3.4.14 BASES SURVEILLANCE SR 3.4.14.1 (continued)

REQUIREMENTS SR 3.4.14.1 is modified by three Notes. Note 1 states that the SR is only required to be performed in MODES 1 and 2. Entry into MODES 3 and 4 is allowed to establish the necessary differential pressure and stable conditions to allow performance of this surveillance.

Note 2 further restricts the PIV leakage rate acceptance criteria by limiting the reduction in margin between the measured leakage rate and the maximum permissible leakage rate by 50% or greater.

Reductions in margin by 50% or greater may be indicative of PIV degradation and warrant inspection or additional testing. Thus, leakage rates less than 5.0 gpm are considered acceptable if the latest measured rate has not exceeded the rate determined by the previous test by an amount that reduces the margin between measured leakage rate and the maximum permissible rate of 5.0 gpm by 50% or greater.

Note 3 limits the minimum test differential pressure to 150 psid during performance of PIV leakage testing .

SR 3.4.14.2 Verifying that the SOC suction valve interlocks are OPERABLE ensures that PCS pressure will not pressurize the SOC system beyond 125% of its design pressure of 300 psig . The interlock setpoint that prevents the valves from being opened is set so the actual PCS pressure must be < 280 psia to open the valves. This setpoint ensures the SOC design pressure will not be exceeded and the SOC relief valves will not lift. The narrow range pressure transmitters that provide the SOC suction valve interlocks are sensed from the pressurizer. Due to the elevation differences between these narrow range pressure transmitter calibration points and the SOC suction piping, the pressure in the SOC suction piping will be higher than the indicated pressurizer B

pressure. Due to this pressure difference, the SOC suction valve interlocks are conservatively set at or below 280 psia to ensure that the 300 psig (315 psia) design pressure of the suction piping is not exceeded . The 18 month frequency is based on the need to perform these Surveillanoes under oonditions that apply during a plant outage.

The 18 month frequenoy is also aooeptable based on oonsideration of Insert 3 the design reliability (and oonfirming operating e*perienoe) of the equipment.

Palisades Nuclear Plant B 3.4.14-6 Revised 07/26/2017

PCS Leakage Detection Instrumentation B 3.4.15 BASES ACTIONS (continued)

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 reasonable confidence the channel is operating properly. The FrequenGY of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is based on instrurnen~

reliability and is reasonable for deteGting off norrnal Gonditions.

Insert 3 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 FrequenGY of ~

18 rnonths is a typiGal refueling GYGle (perforrnanGe of the test is only praGtiGal during a plant outage) and Gonsiders instrurnent reliability. Insert 3 Operating e*perienGe has shown this FrequenGY is aGGeptable for deteGting degradation.

Palisades Nuclear Plant B 3.4.15-5 Revised 02/24/2005

PCS Leakage Detection Instrumentation B 3.4.15 BASES SURVEILLANCE SR 3.4.15.5. SR 3.4.15.6. and SR 3.4.15.7 REQUIREMENTS (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 Fre~l:Ienoy of 18 months experienoe has shown this FFe~l:Ienoy is aooeptable.

REFERENCES 1. FSAR, Section 5.1.5

2. FSAR, Sections 4.7 and 6.3 Palisades Nuclear Plant B 3.4.15-6 Revised 02/24/2005

PCS Specific Activity B 3.4.16 BASES ACTIONS (continued)

If a Required Action and associated Completion Time of Condition A is not met or if the DOSE EQUIVALENT 1-131 is 40 IJCi/gm or above, or with the gross specific activity in excess of the allowed limit, the plant must be placed in a MODE in which the requirement does not apply.

The change within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months to MODE 3 with PCS average temperature

< 500°F lowers the saturation pressure of the primary coolant below the setpoints of the main steam safety valves and prevents venting the SG to the environment in an SGTR event. The allowed Completion Time of 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months is required to reach MODE 3 below 500°F from full power conditions and without challenging plant systems.

SURVEILLANCE SR 3.4.16.1 REQUIREMENTS The Surveillance requires performing a gamma isotopic analysis as a measure of the gross specific activity of the primary coolant at least once per 7 days. While basically a quantitative measure of radionuclides with half lives longer than 15 minutes, excluding iodines, this measurement is the sum of the degassed gamma activities and the gaseous gamma activities in the sample taken. This Surveillance provides an indication of any increase in gross specific activity.

Trending the results of this Surveillance allows proper remedial action to be taken before reaching the LCO limit under normal operating conditions. The Surveillance is applicable in MODES 1 and 2, and in MODE 3 with PCS average temperature at least 500°F. The 7 day ~

Frequency considers the unlikelihood of a gross fuel failure during the time,. Insert 3 Palisades Nuclear Plant B 3.4.16-4 Amendment No. 226 Revised 04/14/2016

PCS Specific Activity B 3.4.16 BASES SURVEILLANCE SR 3.4.16.2 REQUIREMENTS (continued) This Surveillance is performed to ensure iodine remains within limits during normal operation and following fast power changes when fuel failure is more apt to occur. The 14 say F'Feql:lensy is aseql:late te tFens ellery 7 says. The Frequency, between 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months after any power change of;::: 15% RTP within a 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months period, is established because the iodine levels peak during this time following fuel failure; samples at other times would provide inaccurate results. If any (may be more than one) power change;::: 15% RTP occurs within a 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months period, then more than one sample may be required to ensure that an iodine peak sample is obtained between the 2 and 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months Frequency requirement. This SR is modified by a Note which states that the SR is only required to be performed in MODE 1. Entrance into a lower MODE does not preclude completion of this surveillance.

SR 3.4.16.3 A radiochemical analysis for E determination is required every 184 says (6 menths) with the plant operating in MODE 1 equilibrium conditions.

The E determination directly relates to the LCO and is required to verify plant operation within the specified gross activity LCO limit. The analysis for E is a measurement of the average energies per disintegration for isotopes with half lives longer than ~ 5 minutes, excluding iodines. TRe~

F'reql:lensy ef 184 says resegnizes E sees net shange rapisly .

Insert 3 This SR has been modified by a Note that indicates sampling is requir to be performed within 31 days after 2 effective full power days and 20 days of MODE 1 operation have elapsed since the reactor was last subcritical for at least 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months . This ensures the radioactive materials are at equilibrium so the analysis for E is representative and not skewed by a crud burst or other similar abnormal event.

REFERENCES 1. FSAR, Section 14.15 Palisades Nuclear Plant B 3.4.16-5 Amendment No. 226 Revised 04/14/2016

SITs B 3.5.1 BASES ACTIONS C.1 (continued)

If the SIT cannot be restored to OPERABLE status within the associated 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 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . The allowed Completion Time is reasonable, based on operating experience, to reach the required plant conditions from full power condition in an orderly manner and without challenging plant systems.

If more than one SIT is inoperable, the plant is in a condition outside the accident analyses. Therefore, LCO 3.0.3 must be entered immediately.

SURVEILLANCE SR 3.5.1.1 REQUIREMENTS Verification every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months that each SIT isolation valve is fully open, as indicated in the control room, ensures that SITs are available for injection and ensures timely discovery if a valve should be partially 8

closed . If an isolation valve is not fully open, the rate of injection to the PCS would be reduced. Although a motor operated valve should not change position with power removed, a closed valve could result in not meeting accident analysis assumptions. A 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months rrequenoy is oonsidered reasonable in view of other administrative oontrols that ensure the unlikelihood of a mispositioned isolation valve. Insert 3 SR 3.5.1 .2 and SR 3.5.1.3 SIT borated water volume and nitrogen cover pressure should be verified to be within specified limits every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months in order to ensure adequate injection during a LOCA. Due to the statio design of the SITS~

a 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months rrequenoy usually allo..."s the operator suff.ioient time to identify ohanges before the limits are reaohed. Operating experienoe Insert 3 has shown this rrequenoy to be appropriate for early deteotion and oorreotion of 0# normal trends.

Palisades Nuclear Plant B 3.5.1-7 Amendment No. 189

SITs B 3.5.1 BASES SURVEILLANCE SR 3.5.1.4 REQUIREMENTS (continued)

T::hi~Ft:y~o:n: da:'~s:is:r~~~.~~~.bl:e ~:.0:.v!:r~*fi~;:ti~~~:t:~~:t:~F~:i:::t~h:at:~e::ch~

SJT:S:bor~GOn~~~:the~~d~m~booffiJse~e_

JlV.

'-1-"'

~&"'--'

an~ Insert 3 be chanJed. The 31 day Frequency is adequate to identify chanJes that could occur from mechanisms such as stratification or inleakaJe.

SR 3.5.1.5 Verification every 31 days that power is removed from each SIT isolation valve operator ensures that an active failure could not result in the undetected closure of an SIT motor operated isolation valve. If this were to occur, only two SITs would be available for injection, given a single failure coincident with a LOCA. Since installation and removal Of~

power to the SIT isolation valve operators is conducted under administrative control, the 31 day Frequency 'Nas chosen to provide Insert 3 additional assurance that power is removed.

REFERENCES 1. FSAR, Section 14.17

2. FSAR, Chapter 6.1

3. CE-NPSD-994, "CEOG Joint Applications Report for Safety Injection Tank AOT/STI Extension," May 1995 Palisades Nuclear Plant B 3.5.1-8 Amendment No. +89. 191

ECCS - Operating B 3.5.2 BASES SURVEILLANCE SR 3.5.2.1 REQUIREMENTS Verification of proper valve position ensures that the flow path from the ECCS pumps to the PCS is maintained. CV-3027 and CV-3056 are stop valves in the minimum recirculation flow path for the ECCS pumps.

If either of these valves were closed when the PCS pressure was above the shutoff head of the ECCS pumps, the pumps could be damaged by running with insufficient flow and thus render both ECCS trains inoperable.

Placing HS-3027A and HS-3027B for CV-3027, and HS-3056A and HS-3056B for CV-3056, in the open position ensures that the valves cannot be inadvertently misaligned or change position as the result of an active failure. These valves are of the type described in Reference 4, which can disable the function of both ECCS trains and invalidate the accident analysis. CV-3027 and CV-3056 are capable of being closed from the control room since the SIRWT must be isolated

===:==~=:~:i=::: OA "Alikeli Insert 3 from the containment during the recirculation phase of a LOCA. A SR 3.5.2.2 Verifying the correct alignment for manual, power operated, and automatic valves in the ECCS flow paths provides assurance that the proper flow paths will exist for ECCS operation. This SR does not apply to valves that are locked, sealed, or otherwise secured in position, since these valves were verified to be in the correct position prior to locking ,

sealing , or securing . A valve that receives an actuation signal is allowed to be in a nonaccident position provided the valve automatically repositions within the proper stroke time. This Surveillance does not require any testing or valve manipulation. Rather, it involves verification that those valves capable of being mispositioned are in the correct position.

The 31 day Frequency is appropriate because the valves are operated ~

under procedural control and an improper valve position would only affect a single train . This Frequency has been shown to be acceptable Insert 3 through operating m(perience.

Palisades Nuclear Plant B 3.5.2-9 Revised 07/26/2017

ECCS - Operating B 3.5.2 BASES SURVEILLANCE SR 3.5.2.3 REQUIREMENTS (Continued) SR 3.5.2.3 verifies CV-3006 is in the open position and that its air supply is isolated. CV-3006 is the shutdown cooling flow control valve located in the common LPSI flow path. The valve must be verified in the full open position to support the low pressure injection flow assumptions used in the accident analyses. The inadvertent misposition of this valve could result in a loss of low pressure injection flow and thus invalidate these flow assumptions. CV-3006 is designed to be held open by spring force and closed by air pressure. To ensure the valve cannot be inadvertently misaligned or change position as the result of a hot short in the control circuit, the air supply to CV-3006 is isolated. Isolation of the air supply to CV-3006 is acceptable since the valve does not require automatic repositioning during an accident.

The a1 day Frequency has been shown to be acceptable through ~

operating practice and the unlikely occurrence of the air supply to CV a006 being un isolated coincident with a inadvertent valve Insert 3 misalignment event or a hot short in the control circuit.

SR 3.5.2.4 Periodic surveillance testing of ECCS pumps to detect gross degradation caused by impeller damage or other hydraulic component problems is required by the ASME Code. This type of testing may be accomplished by measuring the pump developed head at only one pOint of the pump characteristic curve. This verifies both that the measured performance is within an acceptable tolerance of the original pump baseline performance and that the performance at the test flow is greater than or equal to the performance assumed in the plant safety analysis. SRs are specified in the INSERVICE TESTING PROGRAM of the ASME Code. The ASME Code provides the activities and Frequencies necessary to satisfy the requirements.

SR 3.5.2.5, SR 3.5.2.6, and SR 3.5.2.7 These SRs demonstrate that each automatic ECCS valve actuates to the required position on an actual or simulated actuation signal, i.e., on an SIS or RAS, that each ECCS pump starts on receipt of an actual or simulated actuation signal, i.e. , on an SIS, and that the LPSI pumps stop on receipt of an actual or simulated actuation signal, i.e., on an RAS . RAS opens the HPSI subcooling valve CV-3071, if the associated HPSI pump is operating. After the containment sump valve CV-3030 opens from RAS, HPSI subcooling valve CV-3070 will open, if the associated HPSI pump is operating . RAS will re-position CV-3001 and CV-3002 to a predetermined throttled position. RAS will close Palisades Nuclear Plant B 3.5.2-10 Amendment No. 262 Revised 07/26/2017

ECCS - Operating B 3.5.2 BASES SURVEILLANCE SR 3.5.2.5, SR 3.5.2.6, and SR 3.5.2.7 REQUIREMENTS (continued) containment spray valve CV-3001, if containment sump valve CV-3030 does not open. This Surveillance is not required for valves that are locked, sealed, or otherwise secured in the required position under administrative controls.

The 18 month Freql::lency is based on the need to perform these Sl::Irveillances l::Inder the conditions that apply dl::lring a plant ol::ltage and ~

the potential for l::Inplanned transients if the Sl::Irveillances were performed with the reactor at power. The 18 month Freql::lency is also Insert 3 acceptable based on consideration of the design reliability of the eql::lipment and operating experience. The actuation logic is tested as part of the Engineered Safety Feature (ESF) testing, and equipment performance is monitored as part of the INSERVICE TESTING PROGRAM.

SR 3.5.2.8 The HPSI Hot Leg Injection motor operated valves and the LPSIIoop injection valves have position switches which are set at other than the full open position . This surveillance verifies that these position switches are set properly.

The HPSI Hot leg injection valves are manually opened during the post-LOCA long term cooling phase to admit HPSI injection flow to the PCS hot leg. The open position limit switch on each HPSI hot leg isolation valves is set to establish a predetermined flow split between the HPSI injection entering the PCS hot leg and cold legs.

The LPSI loop injection MOVs open automatically on a SIS signal. The open position limit switch on each LPSIIoop injection valve is set to establish the maximum possible flow through that valve. The design of these valves is such that excessive turbulence is developed in the valve body when the valve disk is at the full open position. Stopping the valve travel at slightly less than full open reduces the turbulence and results in increased flow. Verifying that the position stops are properly set ensures that a single low pressure safety injection subsystem is capable of delivering the flow rate required in the safety analysis.

The 18 month Freql::lency is based on the same factors as those stated ~

above for SR 3.5.2.5, SR 3,5.2.6, and SR 3.5.2.7.

Insert 3 Palisades Nuclear Plant B 3.5.2-11 Amendment No. 262 Revised 07/26/2017

ECCS - Operating B 3.5.2 BASES SR 3.5.2.9 Periodic inspection of the ECCS containment sump passive strainer assemblies ensures that the post-LOCA recirculation flowpath to the ECCS train containment sump suction inlets is unrestricted. Periodic inspection of the containment sump entrance pathways, which include containment sump passive strainer assemblies, containment sump downcomer debris screens, containment floor drain debris screens, containment sump vent debris screens, and reactor cavity corium plug bottom cup support assemblies, ensures that the containment sump stays in proper operating condition. The migration of LOCA-generated debris larger than the strainer perforation diameter through the two one-inch reactor cavity drain line corium plugs is not considered to be credible. The 18 month FreElbiency is based on the need to periorm this ~

SblF\'eiliance binder obltage conditions. This FreElbiency is sblfficient to detect abnormal degradation and is confirmed by operating experience. Insert 3 REFERENCES 1. FSAR, Section 5.1

2. FSAR, Section 14.17

3. NRC Memorandum to V. Stello, Jr., from R. L. Baer, "Recommended Interim Revisions to LCOs for ECCS Components," December 1, 1975

4. IE Information Notice No. 87-01, January 6, 1987

5. CE-NPSD-994, "CEOG Joint Applications Report for Safety Injection Tank AOT/STI Extension," May 1995 Palisades Nuclear Plant B 3.5.2-12 Revised 07/26/2017

SIRWT B 3.5.4 BASES ACTIONS !U.

(continued)

With SIRWT borated water volume not within limits, it must be returned to within limits within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . In this condition, neither the ECCS nor Containment Spray System can perform their design functions; therefore, prompt action must be taken to restore the tank to OPERABLE status or to place the plant in a MODE in which these systems are not required. The allowed Completion Time of 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months to restore the SIRWT to OPERABLE status is based on this condition simultaneously affecting multiple redundant trains.

C.1 and C.2 If the SIRWT cannot be restored to OPERABLE status within the associated 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 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and to MODE 5 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

SURVEILLANCE SR 3.5.4.1 REQUIREMENTS SIRWT borated water temf')eratl-Jre shall be verified every 24 hOl-Jrs to be ~

within the limits assl-Jmed in the accident analysis. This Freql-Jency has been shown to be sl-Jfficient to identify temf')eratl-Jre changes that Insert 3 af')f')roach either accef')table limit.

SR 3.5.4.2 and SR 3.5.4.3

====~=EE=::I~ ,2----'rt3 Since the SIRIJVT voll-Jme is normally stable and is f')rovided with a Low ' - - - - -.....

Level Alarm, a 7 day Freql-Jency is af')f')rof')riate and has been shown to be accef')table throl-Jgh of')erating eXf')erience.

SR 3.5.4.2 is modified by a Note which states that it is only required to be met in MODES 1, 2, and 3.

Palisades Nuclear Plant B 3.5.4-6 Amendment 227

SIRWT B 3.5.4 BASES SURVEILLANCE SR 3.5.4.2 and SR 3.5.4.3 (continued)

REQUIREMENTS SR 3.5.4.3 is modified by a Note which states that it is only required to be met in MODE 4. The required minimum SIRWT water volume is less in MODE 4 since the PCS temperature and pressure are reduced and a significant volume of water is transferred from the SIRWT to the PCS during MODE 4 to account for primary coolant shrinkage.

SR 3.5.4.4 Boron concentration of the SIRWT shall be verified every 31 days to be within the required range . This Frequency ensures that the reactor will remain subcritical following a LOCA. Further, it ensures that the resulting sump pH will be maintained in an acceptable range such that boron precipitation in the core will not occur earlier than predicted and the effect of chloride and caustic stress corrosion on mechanical systems and components will be minimized .

Since the SIRWT volume is normally stable, a 31 day sampling ~

Frequency is appropriate and has been shown through operating experience to be acceptable. Insert 3 REFERENCES 1. FSAR, Chapter 6 and Chapter 14

2. Design Basis Document (DBD) 2.02, "High-Pressure Safety Injection System," Section 3.3.1

3. EOP 4.0, Loss of Coolant Accident Palisades Nuclear Plant B 3.5.4-7 Amendment 227

STB B 3.5.5 BASES SURVEILLANCE SR 3.5.5.1 REQUIREMENTS Periodic determination of the mass of STB in containment must be performed due to the possibility of leaking valves and components in the containment building that could cause dissolution of the STB during normal operation. A Frequency of 18 months j is required to determine that ~ 8,186 Ibs and ~ 10,553 Ibs of equivalent weight of decahydrate STB are contained in the STB baskets. In the event that the total STB weight is less than the minimum weight, a chemical test is performed to confirm that the weight change is due to the dehydration of the decahydrate form of the STB. It is not necessary to replenish STB if the minimum weight is not met solely due to dehydration of the material.

This requirement ensures that there is an adequate mass of STB to adjust the pH of the post LOCA sump solution to a value ~ 7.0 and

~ 8.0.

~ I Insert 3 the mass of STB placed in the containment building.

SR 3.5.5.2 Periodic testing is performed to ensure the solubility and buffering ability of the STB after exposure to the containment environment. Satisfactory completion of this test assures that the STB in the baskets is "active."

Adequate buffering capability is verified by a measured pH of the sample STB in boric acid solution. The quantity of the STB sample and quantity and boron concentration of the water are chosen to be representative of post-LOCA conditions. The pH is measured at 25°C and is verified to be between 7.0 and 8.0.

Palisades Nuclear Plant B 3.5.5-4 Amendment No. 227

STB B 3.5.5 BASES SURVEILLANCE SR 3.5.5.2 (continued)

=::=::=I===b=lel ~

REQUIREMENTS I

Insert 3 I REFERENCES 1. FSAR, Section 6.4 Palisades Nuclear Plant B 3.5.5-5 Amendment No. 227

Containment Air Locks B 3.6.2 BASES SURVEILLANCE SR 3.6.2.2 REQUIREMENTS (continued) The air lock interlock is designed to prevent simultaneous opening of both doors in a single air lock. Since both the inner and outer doors of an air lock are designed to withstand the maximum expected post accident containment pressure, closure of either door will support containment OPERABILITY. Thus, the door interlock feature supports containment OPERABILITY while the air lock is being used for personnel transit into and out of containment. Periodic testing of this interlock demonstrates that the interlock will function as designed and that simultaneous opening of the inner and outer doors will not inadvertently occur. Due to the purely mechanical nature of this interlock, and given that the interlock mechanism is not normally challenged when the airlock is used for entry and exit (procedures require strict adherence to single door opening)., tI:Hs test is only reEl~ires to be l3eoormes ellery 24 months.

and:oos;defed~dequat9g~Zwerk)Gk;no;norrn;y challenges s~ring ~se of the airlock.

1Insert 3 I REFERENCES 1. FSAR, Chapter 14

2. FSAR, Section 5.8

3. 10 CFR 50, Appendix J, Option B Palisades Nuclear Plant B 3.6.2-8 Revised 03/15/2012

Containment Isolation Valves B 3.6.3 BASES ACTIONS C.1 and C.2 (continued)

Required Action C.2 is modified by a Note that applies to isolation devices located in high radiation areas and allows these devices to be verified closed by use of administrative means. Allowing verification by administrative means is considered acceptable, since access to these areas is typically restricted. Therefore, the probability of misalignment of these devices, once they have been verified to be in the proper position, is small.

The purge exhaust and air room supply isolation valves have not been qualified to close following a LOCA and are required to be locked closed.

If one or more of these valves is found not locked closed, the potential exists for the valves to be inadvertently opened. One hour is provided to lock closed the affected valves. The 1-hour Completion Time provides a period of time to correct the problem commensurate with the importance of maintaining these valves closed.

E.1 and E.2 If the Required Actions and associated Completion Times 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 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and to MODE 5 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

SURVEILLANCE SR 3.6.3.1 REQUIREMENTS This SR ensures that the 8-inch purge exhaust and 12 inch air room supply valves are locked closed as required . If a valve is open, or closed but not locked, in violation of this SR, the valve is considered inoperable.

Valves may be locked closed electrically, mechanically, or by other physical means. These valves may be unable to close in the environment following a LOCA. Therefore, each of the valves is required to remain closed during MODES 1, 2, 3, and 4. The 31 day FrequenGY is GOnsistent~

with other Gontainment isolation valve requirements disGussed in SR 3.6.3.2. Insert 3 Palisades Nuclear Plant B 3.6.3-9 Revised 07/26/2017

Containment Isolation Valves B 3.6.3 BASES SURVEILLANCE SR 3.6.3.2 REQUIREMENTS (continued) This SR requires verification that each manual containment isolation valve and blind flange located outside containment, and not locked, sealed, or otherwise secured in position, and required to be closed during accident conditions, is closed . The SR helps to ensure that post accident leakage of fission products outside the containment boundary is within design limits. This SR does not require any testing or device manipulation. Rather, it involves verification that those containment isolation devices outside containment and capable of being mispositioned are in the correct position. Since verification of device position for containment isolation devices outside containment is relatively easy, the 31 day Frequency is based on engineering judgment and was chosen to provide added assurance of the correct positions. Containment isolation valves that are open under administrative controls are not required to meet the SR during the time the valves are open. This SR does not apply to devices that are locked, sealed, or otherwise secured in the closed position, since these were verified to be in the correct position upon locking, sealing, or seCUring. ~

The Note applies to valves a Insert 3 s located in high radiation areas and allows these devic s v I led closed by use of administrative means. Allowing verification by administrative means is considered acceptable, since access to these areas is typically restricted during MODES 1, 2, 3, and 4 for ALARA reasons. Therefore, the probability of misalignment of these containment isolation devices, once they have been verified to be in the proper position, is small.

SR 3.6.3.3 This SR requires verification that each containment isolation manual valve and blind flange located inside containment and not locked, sealed or otherwise secured in position, and required to be closed during accident conditions, is closed. The SR helps to ensure that post accident leakage of fission products outside the containment boundary is within design limits. For containment isolation devices inside containment, the Frequency of "prior to entering MODE 4 from MODE 5 if not performed within the previous 92 days" is appropriate, since these containment isolation devices are operated under administrative controls and the probability of their misalignment is low. Containment isolation valves that are open under administrative controls are not required to meet the SR during the time that they are open. This SR does not apply to devices that are locked, sealed, or otherwise secured in the closed position, since these were verified to be in the correct position upon locking, sealing, or securing .

Palisades Nuclear Plant B 3.6.3-10 Revised 07/26/2017

Containment Isolation Valves B 3.6.3 BASES SURVEILLANCE SR 3.6.3.3 (continued)

REQUIREMENTS The Note allows valves and blind flanges located in high radiation areas to be verified closed by use of administrative means. Allowing verification by administrative means is considered acceptable, since access to these areas is typically restricted during MODES 1, 2, and 3 for ALARA reasons. Therefore, the probability of misalignment of these containment isolation devices, once they have been verified to be in their proper position, is small.

SR 3.6.3.4 Verifying that the isolation time of each automatic power operated containment isolation valve is within limits is required to demonstrate OPERABILITY. The isolation time test ensures the valve will isolate in a time period less than or equal to that assumed in the safety analysis. The isolation time and Frequency of this SR are in accordance with the INSERVICE TESTING PROGRAM.

SR 3.6.3.5 For containment 8 inch purge exhaust and 12 inch air room supply valves with resilient seals, additional leakage rate testing beyond the test requirements of 10 CFR 50, Appendix J, Option B (Ref. 3), is required to ensure the valves are physically closed (SR 3.6.3.1 verifies the valves are locked closed). Operating experience has demonstrated that this type of seal has the potential to degrade in a shorter time period than do other seal types. Based on this observation and the importance of maintaining ~

this penetration leak tight (due to the direct path between containment and the en\'ironment), a Frequency of 184 days 'Nas established as part Insert 3 of the NRC resolution of Generic Issue B 20, "Containment Leakage Due to Seal Deterioration" (Ref. 4) as specified in the Safety Evaluation for Amendment No. 90 to the Facility Operating License.

Palisades Nuclear Plant B 3.6.3-11 Amendment 262 Revised 07/26/2017

Containment Isolation Valves B 3.6.3 BASES SURVEILLANCE SR 3.6.3.6 REQUIREMENTS (continued) Automatic containment isolation valves close on a containment isolation signal to minimize leakage of fission products from containment following a DBA. This SR ensures each automatic containment isolation valve will actuate to its isolation position on an actual or simulated actuation Signal, i.e., CHP or CHR. This Surveillance is not required for valves that are locked, sealed, or otherwise secured in the required position under administrative controls. The 18 month FrequenGY was developed Gonsidering it is prudent that this SR l::le performed only during a plant outage, sinGe isolation of penetrations would eliminate Gooling water flow and disrupt normal operation of many GritiGal Gomponents. Operating e*perienGe has shown that these Gomponents usually pass this SR when performed on the 18 month FrequenGy. Therefore, the FrequenGY was GonGluded to l::le aGGeptal::lle from a relial::lility standPoint..-_.-,-'-_--,

REFERENCES 1. FSAR, Section 5.8 L...-------Ir--

Insert 3

2. FSAR, Section 6.7.2 and Table 6-14

3. 10 CFR 50, Appendix J, Option B

4. Generic Issue B-20

5. FSAR, Chapter 14

6. FSAR, Section 1.4.16 Palisades Nuclear Plant B 3.6.3-12 Revised 07/26/2017

Containment Pressure B 3.6.4 BASES ACTIONS B.1 and B.2 (continued)

If containment pressure cannot be restored to within limits 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 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and to MODE 5 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

SURVEILLANCE SR 3.6.4.1 REQUIREMENTS Verifying that containment pressure is within limits ensures that operation remains within the limits assumed in the accident analyses. The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Frequency of this SR was developed after taking into consideration operating experience related to trending of containment pressure variations during the applicable MODES. Furthermore, the 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Frequency is considered adequate in view of other indications available in the control room, including alarms, to alert the operator to an abnormal containment pressure condition. The limit of 1.0 psig for MODES 1 and 2, 1.5 psig for MODES 3 and 4 are the actual limits used in the accident analysis and do not account for instrument inaccuracies.

REFERENCES 1. FSAR, Section 14.18 Palisades Nuclear Plant B 3.6.4-3 Revised 03/15/2012

Containment Air Temperature B 3.6.5 BASES ACTIONS B.1 and B.2 (continued)

If the containment average air temperature cannot be restored to within its limit 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 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and to MODE 5 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

SURVEILLANCE SR 3.6.5.1 REQUIREMENTS Verifying that containment average air temperature is within the LCO limit ensures that containment operation remains within the limit assumed for the containment analyses. The 145°F limit is the actual limit assumed for the accident analyses and does not account for instrument inaccuracies.

Instrument uncertainties are accounted for in the surveillance procedure.

The 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Frequency of this SR is considered acceptable based on th~

observed slow rates of temperature increase within containment as a result of environmental heat sources (due to the large volume of Insert 3 containment).

REFERENCES 1. FSAR, Section 5.8

2. FSAR, Section 14.18 Palisades Nuclear Plant B 3.6.5-3 Revised 03/15/2012

Containment Cooling Systems B 3.6.6 BASES ACTIONS C.1 (continued)

If the Containment Spray side (tube side) of SOC Heat Exchanger E-60B is out of service, 100% of the required post accident cooling capability can be provided, if other equipment outages are limited. One hundred percent of the post accident cooling can be provided with the Containment Spray side of SOC Heat Exchanger E-60B out of service if the following equipment is OPERABLE: three safety related Containment Air Coolers, two Containment Spray Pumps, two spray headers, CCW pumps P-52A and P-52B, two SWS pumps, and both CCW Heat Exchangers, and if

1. One CCW Containment Isolation Valve, CV-0910, CV-0911, or CV-0940, is OPERABLE, and

2. Two CCW isolation valves for the non-safety related loads outside the containment, CV-0944A and CV-0944 (or CV-0977B), are OPERABLE.

With less than 100% of the required post accident containment cooling capability available, the plant is in a condition outside the assumptions of the safety analyses. Therefore, LCO 3.0.3 must be entered immediately.

SURVEILLANCE SR 3.6.6.1 REQUIREMENTS Verifying the correct alignment for manual , power operated , and automatic valves, excluding check valves, in the Containment Spray System provides assurance that the proper flow path exists for Containment Spray System operation. This SR also does not apply to valves that are locked, sealed, or otherwise secured in position since these were verified to be in the correct positions prior to being secured.

This SR also does not apply to valves that cannot be inadvertently misaligned, such as check valves. This SR does not require any testing or valve manipulation. Rather, it involves verification that those valves outside containment and capable of potentially being mispositioned, are in the correct position. ~

SR3.6.6.2 ~

Operating each safety related Containment Air Cooler fan unit for

~ 15 minutes ensures that all trains are OPERABLE and are functioning properly. The 31 day Frequency '...,as developed considering the known reliability of the fan units, the two train redundancy available, and the low probability of a significant degradation of the containment cooling train occurring beF.a;een surveillances . .--_.-,----L.._---.

Insert 3 Palisades Nuclear Plant B 3.6.6-10 Revised 07/26/2017

Containment Cooling Systems B 3.6.6 BASES SURVEILLANCE SR 3.6.6.3 REQUIREMENTS (continued) Verifying the containment spray header is full of water to the 735 ft elevation minimizes the time required to fill the header. This ensures that spray flow will be admitted to the containment atmosphere within the time frame assumed in the containment analysis. The 31 day Frequency is based on the static nature of the fill header and the low probability of a significant degradation of the water level in the piping occurring between surveillances. ~

SR 3.6.6.4 ~

Verifying a total service water flow rate of ~ 4800 gpm to CACs VHX-1 ,

VHX-2, and VHX-3, when aligned for accident conditions, provides assurance the design flow rate assumed in the safety analyses will be achieved (Ref. 8). Also considered in selecting this Frequency were the known reliability of the cooling water system, the two train redundancy,

RsS;_illaRSej ,2rt I and the low probability of a significant degradation of flow occurring 3

Verifying that each containment spray pump's developed head at the flow test point is greater than or equal to the required developed head ensures that spray pump performance has not degraded during the cycle. Flow and differential pressure are normal tests of centrifugal pump performance required by the ASME Code (Ref. 5).

Since the containment spray pumps cannot be tested with flow through the spray headers, they are tested on recirculation flow. This test confirms one point on the pump design curve and is indicative of overall performance. Such inservice inspections confirm component OPERABILITY, trend performance, and detect incipient failures by indicating abnormal performance. The Frequency of this SR is in accordance with the INSERVICE TESTING PROGRAM.

Palisades Nuclear Plant B 3.6.6-11 Amendment No. 262 Revised 07/26/2017

Containment Cooling Systems B 3.6.6 BASES SURVEILLANCE SR 3.6.6.6 and SR 3.6.6.7 REQUIREMENTS (continued) SR 3.6.6.6 verifies each automatic containment spray valve actuates to its correct position upon receipt of an actual or simulated actuation signal.

This Surveillance is not required for valves that are locked, sealed, or otherwise secured in the required position under administrative controls.

SR 3.6.6.7 verifies each containment spray pump starts automatically on an actual or simulated actuation signal. The 18 month Frequency is ~

based on the need to perform these Surveillances under the conditions that apply during a plant outage and the potential for an unplanned Insert 3 transient if the Surveillances 'Nere performed with the reactor at power.

Operating experience has sho'lm that these components usually pass the Surveillances when performed at the 18 month Frequency. Therefore, the Frequency was concluded to be acceptable from a reliability standpoint.

Where the surveillance of containment sump isolation valves is also required by SR 3.5.2.5, a single surveillance may be used to satisfy both requirements.

SR 3.6.6.8 This SR verifies each safety related containment cooling fan actuates upon receipt of an actual or simulated actuation signal. The 18 month ~

Frequency is based on engineering judgement and has been shown to be acceptable through operating experience. See SR 3.6.6.6 and SR Insert 3 3.6.6.7, above, for further discussion of the basis for the 18 month Frequency.

SR 3.6.6.9 With the containment spray inlet valves closed and the spray header drained of any solution, an inspection of spray nozzles, or a test that blows low-pressure air or smoke through test connections can be completed. Performance of this SR demonstrates that each spray nozzle is unobstructed and provides assurance that spray coverage of the containment during an accident is not degraded. Verification following maintenance which could result in nozzle blockage is appropriate because this is the only activity that could lead to nozzle blockage.

Palisades Nuclear Plant B 3.6.6-12 Revised 07/26/2017

MSIVs B 3.7.2 BASES SURVEILLANCE SR 3.7.2.1 REQUIREMENTS This SR verifies that the closure time of each MSIV is:::; 5.0 seconds on an actual or simulated actuation signal from each train under no flow conditions. Specific signals (e.g., Containment High Pressure, Steam Generator Low Pressure, handswitch) are tested under Section 3.3, "Instrumentation ." The MSIV closure time is assumed in the MSLB and containment analyses. This SR is normally performed during a refueling outage. The MSIVs are not tested at power since even a part stroke exercise increases the risk of a valve closure with the plant generating power. As the MSIVs are not tested at power, they are exempt from the ASME Code,Section XI (Ref. 5) requirements during operation in MODES 1 and 2.

The Frequency for this SR is every 18 months. This 18 month FreqUency ~

demonstrates the valve closure time at least once l3er refueling cycle.

Ol3erating eXl3erience has shown that these coml3onents usually l3ass the Insert 3 SR when l3erformed at the 18 month Frequency. Therefore, the Frequency is accel3table from a reliability standl3oint.

REFERENCES 1. FSAR, Section 10.2

2. FSAR, Section 14.18

3. FSAR, Section 14.14

4. 10 CFR 50.67

5. ASME, Boiler and Pressure Vessel Code,Section XI, Inservice Inspection, Article IWV-3400 Palisades Nuclear Plant B 3.7.2-6 Amendment No. 226 Revised 04/14/2016

MFRVs and MFRV Bypass Valves B 3.7.3 BASES ACTIONS A.1 and A.2 (continued)

(continued)

Therefore, while Required Action 3.7.3 A.2 must be initially performed within 7 days without any SR 3.0.2 extension, subsequent performances may utilize the 25% SR 3.0.2 extension .

B.1 and B.2 If the MFRVs or MFRV bypass valves cannot be restored to OPERABLE status, closed, or isolated in the associated Completion Time, the plant must be placed in a MODE in which the LCO does not apply. To achieve this status, the plant must be placed in at least MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and in MODE 4 within 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

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

SURVEILLANCE SR 3.7.3.1 REQUIREMENTS This SR verifies the closure time for each MFRV and MFRV bypass valve is :S: 22.0 seconds on an actual or simulated actuation signal. Specific signals (e.g., steam generator low pressure and containment high pressure) are tested under Section 3.3, "Instrumentation ." The MFRV and MFRV bypass valves closure times are bounding values assumed in the MSLB containment response and core response (DNB) analyses (Refs. 3 and 4) . This SR is normally performed during a refueling outage. The MFRVs and MFRV bypass valves should not be tested at power since even a part stroke exercise increases the risk of a valve closure with the plant generating power. As these valves are not stroke tested at power, they are exempt from the ASME Code,Section XI (Ref. 2) requirements during operation in MODES 1 and 2.

The FreE(l:IeRoy is 18 mORths. The 18 mORth FreE(l:IeRoy for valve Olosl:lre~

time is based OR the refl:leliRg oyole. OperatiRg experieRoe has sho'lm that these oompoReRts I:Isl:lally pass the SR 'lfheR performed at the Insert 3 18 mORth FreE(UeRoy.

Palisades Nuclear Plant B 3.7.3-4 Revised 12/02/2002

ADVs B 3.7.4 BASES SURVEILLANCE SR 3.7.4.1 REQUIREMENTS To perform a controlled cooldown of the PCS, the ADVs must be able to Er"===:

be cycled through their full range. This SR ensures the ADVs are tested through a full control cycle at least onse per 18 months. Performance of inservice testing or use of an ADV during a plant cooldown may satisfy this requirement *l

=s;:~ ~SS:~SR¥lhe~:e~ih:1S:n:::it~ Insert 3 standpoint.

REFERENCES 1. FSAR, Section 10.2

2. FSAR, Section 9.5.3 Palisades Nuclear Plant B 3.7.4-4 Revised 07/16/2008

AFW System B 3.7.5 BASES SURVEILLANCE SR 3.7.5.1 REQUIREMENTS Verifying the correct alignment for the required manual, power operated, and automatic valves in the AFW water and steam supply flow path provides assurance that the proper flow paths exist for AFW operation.

This SR does not apply to valves that are locked, sealed, or otherwise secured in position, since these valves are verified to be in the correct position prior to locking, sealing, or securing. This SR also does not apply to valves that cannot be inadvertently misaligned, such as check valves. This Surveillance does not require any testing or valve manipulations; rather, it involves verification that those valves capable of potentially being mispositioned are in the correct position .

This test need not be performed for the steam driven AFW pump for MODE 4 operation.

The 31 day Frequenoy is based on engineering judgment, is oonsistent~

with the prooedural oontrols governing valve operation, and ensures oorrect valve positions. Insert 3 SR 3.7.5.2 Verifying that each required AFW pump's developed head at the flow test point is greater than or equal to the required developed head ensures that AFW pump performance has not degraded during the cycle. Flow and differential head are normal tests of pump performance required by the ASME Code (Ref. 2) . This test confirms one point on the pump design curve and is indicative of overall performance. Such inservice tests confirm component OPERABILITY, trend performance, and detect incipient failures by indicating abnormal performance.

This SR is modified by a Note indicating that this SR for the turbine driven AFW pump does not have to be met in MODE 3 when steam pressure is below 800 psig. This is because there is insufficient steam pressure and pump discharge pressure to allow the turbine driven pump to reach the normal test conditions.

Performance of inservice testing as discussed in the ASME Code (Ref.

2), and the INSERVICE TESTING PROGRAM satisfies this requirement.

Palisades Nuclear Plant B 3.7.5-8 Amendment No. 262 Revised 07/26/20 17

AFWSystem B 3.7.5 BASES SURVEILLANCE SR 3.7.5.3 REQUIREMENTS (continued) This SR ensures that AFW can be delivered to the appropriate steam generator, in the event of any accident or transient that generates an AFAS, by demonstrating that each automatic valve in the flow path actuates to its correct position on an actual or simulated actuation signal.

Specific signals (e.g., AFAS) are tested under Section 3.3, "Instrumentation." This Surveillance is not required for valves that are locked, sealed , or otherwise secured in the required position under administrative controls. The 18 month Frequency is acceptable, based~

on the design reliability and operating experience of the equipment.

Insert 3 This SR is modified by a Note which states the SR is only required to b met in MODES 1, 2, and 3 when AFW is not in operation. With AFW in operation, the required trains are already aligned with the flow control valves in manual control.

SR 3.7.5.4 This SR ensures that the AFW pumps will start in the event of any accident or transient that generates an AFAS by demonstrating that each AFW pump starts automatically on an actual or simulated actuation signal. Specific signals (e.g., AFAS, handswitch) are tested under Section 3.3, "Instrumentation."

This test need not be performed for the steam driven AFW pump for MODE 4 operation .

The 18 month Frequency is acceptable, based on the design reliabilitY~

and operating experience of the equipment.

Insert 3 This SR is modified by a Note. The Note states that the SR is only required to be met in MODES 1, 2, and 3. In MODE 4, the required pump is already operating and the autostart function is not required .

REFERENCES 1. FSAR, Section 9.7

2. ASME Code for Operation and Maintenance of Nuclear Power Plants.

3. Palisades Design Basis Document 1.03, Auxiliary Feedwater System, Section 3.4.1 .

Palisades Nuclear Plant B 3.7.5-9 Amendment No. 262 Revised 07/26/2017

Condensate Storage and Supply B 3.7.6 BASES SURVEILLANCE SR 3.7.6.1 REQUIREMENTS This SR verifies that the combination of CST and T-81 contain the required useable volume of cooling water. (This volume ~ 100,000

=:========~'- I-?. .L. -ert "

Frequency is considered adequate in view of other indications in the control room, including alarms, to alert the operator to abnormal CST and L...-_ _----J T 81 level deviations.

REFERENCES 1. FSAR, Section 9.7

2. Analysis EA-GOTHIC-CST-01 Palisades Nuclear Plant B 3.7.6-4 Revised 05/30/2018

CCW System B 3.7.7 BASES SURVEILLANCE SR 3.7.7.1 (continued)

REQUIREMENTS The 31 day Frequency is based on engineering judgment, is consistent~

with the procedural controls gOI/erning valve operation, and ensures correct valve positions. Insert 3 SR 3.7.7.2 This SR verifies proper automatic operation of the CCW valves on an actual or simulated actuation signal. Specific signals (e.g., safety injection, RAS) are tested under Section 3.3, "Instrumentation ." This Surveillance is not required for valves that are locked, sealed, or otherwise secured in the required position under administrative controls.

This SR is modified by a Note which states this SR is only required to be met in MODES 1, 2, and 3. The instrumentation providing the input Signal is not required in MODE 4, therefore, to keep consistency with Section 3.3, "Instrumentation," the SR is not required to be met in this MODE.

Operating experience has shown that these components usually pass the ~

Surveillance when performed at the 18 month Frequency. Therefore, the Frequency is acceptable from a reliability standpoint. Insert 3 SR 3.7.7.3 This SR verifies proper automatic operation of the CCW pumps on an actual or simulated actuation signal in the "with standby power available" mode which tests the starting of the pumps by the SIS-X relays. The starting of the pumps by the sequencer is performed in Section 3.8, "Electrical Power Systems." This SR is modified by a Note which states this SR is only required to be met in MODES 1, 2, and 3. The instrumentation providing the input signal is not required in MODE 4, therefore, to keep consistency with Section 3.3, "Instrumentation," the SR is not required to be met in this MODE. Operating experience has shQ!,'Vn ~

these components usually pass the Surveillance when performed at the 18 month Frequency. Therefore, the Frequency is acceptable from a Insert 3 reliability standpoint.

REFERENCES 1. FSAR, Section 9.3 Palisades Nuclear Plant B 3.7.7-9 Revised 06/07/2005

SWS B 3.7.8 BASES ACTIONS (continued)

1. The non-critical SWS header isolation valve, CV-1359, is OPERABLE, or

2. Plant conditions allow adequate containment cooling to be provided without reliance on CACs and one SWS Containment Isolation Valve, CV-0824 or CV-0847, is OPERABLE.

One hundred percent of the required SWS post accident cooling capability can be provided by three SWS pumps even with SWS flow being provided to both the CACs and the Non-critical SWS header.

With less than 100% of the required SWS post accident cooling capability available, the plant is in a condition outside the assumptions of the safety analyses. Therefore, LCO 3.0.3 must be entered immediately.

SURVEILLANCE SR 3.7.8.1 REQUIREMENTS Verifying the correct alignment for manual, power operated, and automatic valves in the SWS flow path ensures that the proper flow paths exist for SWS operation. This SR does not apply to valves that are locked, sealed, or otherwise secured in position, since they are verified to be in the correct position prior to locking, sealing, or securing. This SR also does not apply to valves that cannot be inadvertently misaligned, such as check valves. This Surveillance does not require any testing or valve manipulation; rather, it involves verification that those valves capable of potentially being mispositioned are in the correct position.

This SR is modified by a Note indicating that the isolation of SWS to components or systems may render those components inoperable but does not affect the OPERABILITY of the SWS.

The a1 aay Frequency is basea on engineering juagment, is consistent~

'It'ith the proceaural controls governing valve operation, ana ensures correct valve positions. Insert 3 Palisades Nuclear Plant B 3.7.8-7 Revised 10/29/2009

SWS B 3.7.8 BASES SURVEILLANCE SR 3.7.8.2 REQUIREMENTS This SR verifies proper automatic operation of the SWS valves on an actual or simulated actuation signal. Specific signals (e.g., safety injection) are tested under Section 3.3, "Instrumentation." This Surveillance is not required for valves that are locked, sealed, or otherwise secured in the required position under administrative controls.

This SR is modified by a Note which states this SR is only required to be met in MODES 1, 2, and 3. The instrumentation providing the input signal is not required in MODE 4, therefore, to keep consistency with Section 3.3, "Instrumentation," the SR is not required to be met in this MODE.

Operating experience has sholNn that these components usually pass the~

Surveillance when performed at the 18 month Frequency. Therefore, the Frequency is acceptable from a reliability standpoint. Insert 3 SR 3.7.8.3 The SR verifies proper automatic operation of the SWS pumps on an actual or simulated actuation signal in the "with standby power available" mode which tests the starting of the pumps by the SIS-X relays. The starting of the pumps by the sequencer is performed in Section 3.8, "Electrical Power Systems." This SR is modified by a Note which states this SR is not required to be met in MODE 4. The instrumentation providing the input signal is not required in MODE 4, therefore, to keep consistency with Section 3.3, "Instrumentation," the SR is not required to be met in this MODE. Operating experience has shown that these ~

components usually pass the Surveillance when performed at the 18 month Frequency. Therefore, the Frequency is acceptable from a Insert 3 reliability standpoint.

REFERENCES 1. FSAR, Section 9.1

2. FSAR, Section 6.1 Palisades Nuclear Plant B 3.7.8-8 Revised 10/29/2009

UHS B 3.7.9 BASES ACTIONS A.1 and A.2 If the UHS is inoperable, the plant must be placed in a MODE in which the LCO does not apply. To achieve this status, the plant must be placed in at least MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in MODE 5 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

SURVEILLANCE SR 3.7.9.1 REQUIREMENTS This SR verifies adequate cooling can be maintained. The level specified also ensures sufficient NPSH is available for operating the SWS pumps. The 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Frequency is based on operating m<perience related to the trending of the parameter variations during the applicable MODES. This SR verifies that the UHS water level is ~ 568.25 ft above mean sea level as measured within the boundaries of the intake structure. ~

Insert 3 SR 3.7.9.

This SR verifies that the SWS is available to provide adequate cooling for normal design heat loads and maximum accident conditions following a DBA. The 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Frequency is based on operating experience related to the trending of the parameter variations during the applicable MODES. This SR verifies that the water temperature from the UHS is::; 85°F. ~

!Insert 3 REFERENCES 1. FSAR, Section 9.

2. FSAR, Section 14.18

3. Design Basis Document (DBD) 1.02, "Service Water System" Palisades Nuclear Plant B 3.7.9-3 Revised 04/14/2011

CRV Filtration B 3.7.10 BASES ACTIONS E.1 . E.2. and E.3 (continued)

(continued) trains inoperable due to an inoperable CRE boundary, action must be taken immediately to suspend activities that could result in a release of radioactivity that might enter the CRE. This places the plant in a '

condition that minimizes the accident risk, This does not preclude the movement of fuel assemblies or a fuel cask to a safe position ,

F,1 and F.2 If an inoperable CRV Filtration or the CRE boundary cannot be restored to OPERABLE status within the required Completion Time in MODE 1, 2, 3, or 4, the plant must be placed in a MODE that minimizes the accident risk, To achieve this status, the plant must be placed in at least MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and in MODE 5 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems, SURVEILLANCE SR 3,7,10,1 REQUIREMENTS Standby systems should be checked periodically to ensure that they function properly. Since the environment and normal operating conditions on this s stem are not severe, testing each train once every month t ic c provides an adequate check on this system.

Monthly hB eater operations dry out any moisture accumulated in the B

charcoal from humidity in the ambient air. Each train must be operated for

~ 10 continuous hours with the associated heater, VHX-26A or VHX-26B, energized . The 31 say Fre1l:Jency is bases on the known reliability ofthe e1l:Jipment, ans the two train resl:Jnsancy available.

Insert 3 SR 3.7.10.2 This SR verifies that the required CRV Filtration testing is performed in accordance with the Ventilation Filter Testing Program (VFTP). The CRV Filtration filter tests are in accordance with the VFTP. The VFTP includes testing HEPA filter performance, charcoal adsorber efficiency, minimum system flow rate, and the physical properties of the activated charcoal (general use and following specific operations) . Specific test Frequencies and additional information are discussed in detail in the VFTP.

Palisades Nuclear Plant B 3.7.10-7 Amendment No. 256 Revised 07/29/2015

CRV Filtration B3.7.10 BASES SURVEILLANCE SR 3.7.10.3 REQUIREMENTS (continued) This SR verifies that each CRV Filtration train starts and operates on an actual or simulated actuation signal. Specific signals (e.g., containment high pressure, containment high radiation) are tested under Section 3.3, "Instrumentation ." This SR is modified by a Note which states this SR is only required to be met in MODES 1, 2, 3 and 4 and during movement of irradiated fuel assemblies in containment. The instrumentation providing the input signal is not required in other plant conditions, therefore, to keep consistency with Section 3.3, "Instrumentation," the SR is not required to be met. The F'FeEll:lency of 18 months is easeEl on inEil:lstFy opeFating ~

m<peFience anEl is consistent '.... ith the typical FeNeling cycle. ~

SR 3.7.10.4 This SR verifies the OPERABILITY of the CRE boundary by testing for unfiltered air in leakage past the CRE boundary and into the CRE. The details of the testing are specified in the Control Room Envelope Habitability Program.

The CRE is considered habitable when the radiological dose to CRE occupants calculated in the analyses of DBA consequences is no more than 5 rem TEDE and the CRE occupants are protected from hazardous chemicals and smoke. This SR verifies that the unfiltered air in leakage into the CRE is no greater than the flow rate assumed in the analyses of DBA consequences. When unfiltered air inleakage is greater than the assumed flow rate, Condition B must be entered . Required Action B.3 allows time to restore the CRE boundary to OPERABLE status provided mitigating actions can ensure that the CRE remains within the licensing basis habitability limits for the occupants following an accident.

Compensatory measures are discussed in Regulatory Guide 1.196, Section C.2.7.3, (Ref. 4) which endorses, with exceptions, NEI 99-03, Section 8.4 and Appendix F (Ref. 5). These compensatory measures may also be used as mitigating actions as required by Required Action B.2.

Temporary analytical methods may also be used as compensatory measures to restore OPERABILITY (Ref. 6). Options for restoring the CRE boundary to OPERABLE status include changing the DBA consequence analysis, repairing the CRE boundary, or a combination of these actions. Depending upon the nature of the problem and the corrective action , a full scope inleakage test may not be necessary to establish that the CRE boundary has been restored to OPERABLE status.

Palisades Nuclear Plant B 3.7.10-8 Amendment No. 256 Revised 07/29/2015

CRVCooling 83.7.11 8ASES SURVEILLANCE SR 3.7.11 .1 REQUIREMENTS This SR verifies that the heat removal capability of the system is sufficient to meet design requirements. This SR consists of a combination of testing and calculations. An 18 month FreEll:Jensy is aJ3J3F9J3riate, sinse ~

signifisant degradation of the CRV Coaling is slav.' and is nat eXJ3ested aver this time J3eriad. Insert 3 REFERENCES 1. FSAR, Section 9.8

2. WCAP-16125-NP-A, "Justification for Risk-Informed Modification to Selected Technical Specifications for Conditions Leading to Exigent Plant Shutdown," Revision 2, August 2010.

Palisades Nuclear Plant 83.7.11-5 Amendment No. 256 Revised 07/29/2015

Fuel Handling Area Ventilation System B 3.7.12 BASES SURVEILLANCE SR 3.7.12.1 REQUIREMENTS This SR verifies the performance of Fuel Handling Area Ventilation System filter testing in accordance with the Ventilation Filter Testing Program. The Fuel Handling Area Ventilation System filter tests are in accordance with the Regulatory Guide 1.52 (Ref. 6) as described in Ventilation Filter Testing Program. The Ventilation Filter Testing Program includes testing HEPA filter performance, charcoal adsorber efficiency, minimum system flow rate, and the physical properties of the activated charcoal (general use and following specific operations). Specific test frequencies and additional information are discussed in detail in the Ventilation Filter Testing Program.

SR 3.7.12.2 This SR verifies the Fuel Handling Area Ventilation System has not degraded and is operating as assumed in the safety analysis. The flow rate is periodically tested to verify proper function of the Fuel Handling Ventilation System. When aligned to the "emergency filter bank", the Fuel Handling Area Ventilation System is designed to reduce the amount of unfiltered leakage from the fuel handling building which, in the event of a fuel handling accident, lowers the dose at the site boundary to within the applicable limits of 10 CFR 50.67. The Fuel Handling Area Ventilation System is designed to lower the dose to these levels at a flow rate of

5840 cfm and

s; 8760 cfm. The Freq~enGy of 18 months is GOnsistent ~

with the test for filter performanGe and other filtration SRs.

Insert 3 Palisades Nuclear Plant B 3.7.12-6 Amendment No. 226 Revised 04/14/2016

ESRV Dampers 83.7.13 8ASES SURVEILLANCE SR 3.7.13.1 REQUIREMENTS This SR verifies that each ESRV Damper train closes on an actual or simulated actuation signal. The 31 Elay Freql:lency is baseEi on ~

operating experience which has sho',tm that these components l:Isl:lally pass the SR when testeEi at this Freql:lency. Therefore, the Freql:lency is Insert 3 acceptable from a reliability stanEipoint.

REFERENCES 1. 10 CFR 50.67

2. FSAR, Section 7.4.5.2

3. FSAR, Section 14.22 Palisades Nuclear Plant 83.7.13-3 Amendment No. 226 Revised 04/14/2016

SFP Water Level B 3.7.14 BASES SURVEILLANCE SR 3.7.14.1 B

REQUIREMENTS This SR verifies sufficient SFP water is available in the event of a fuel handling or fuel cask drop accident. The water level in the SFP must be checked periodically. The 7 day Frequenoy is appropriate beoause the volume in the pool is normally stable. Water level ohanges are oontrolled by plant prooedures and are aooeptable, based on oporating experienoe. Insert 3 During refueling operations, the level in the SFP is at equilibrium with that of the refueling cavity, and the level in the refueling cavity is checked Gaily ef eal~ in accordance with LCO 3.9 .6, "Refueling Cavity Water Level."

REFERENCES 1. FSAR, Section 9.11

2. FSAR, Section 9.4

3. FSAR, Section 14.19

4. FSAR, Section 14.11

5. Regulatory Guide 1.183

6. 10 CFR 50.67 Palisades Nuclear Plant B 3.7.14-3 Amendment No. 226 Revised 04/14/2016

SFP Boron Concentration 83.7.15 8ASES ACTIONS The ACTIONS are modified by a Note indicating that LCO 3.0.3 does not apply.

If moving irradiated fuel assemblies while in MODE 5 or 6, LCO 3.0.3 would not specify any action. If moving irradiated fuel assemblies while in MODE 1, 2, 3, or 4, the fuel movement is independent of reactor operation . Therefore, inability to suspend movement of fuel assemblies is not sufficient reason to require a reactor shutdown .

A.1 . and A.2 When the concentration of boron in the spent fuel pool is less than required, immediate action must be taken to preclude an accident from happening or to mitigate the consequences of an accident in progress.

This is most efficiently achieved by immediately suspending the movement of fuel assemblies. This does not preclude the movement of fuel assemblies to a safe position. In addition, action must be immediately initiated to restore boron concentration to within limit.

SURVEILLANCE SR 3.7.15.1 REQUIREMENTS This SR verifies that the concentration of boron in the spent fuel pool is within the required limit. As long as this SR is met, the analyzed incidents are fully addressed . The 7 day Frequency is appropriate because no ~

major replenishment of pool water is expected to take place over a short period of time. Insert 3 REFERENCES None Palisades Nuclear Plant 83.7.15-2 Amendment No. 43B,~, 236

Secondary Specific Activity B 3.7.17 BASES ACTIONS A.1 and A.2 DOSE EQUIVALENT 1-131 exceeding the allowable value in the secondary coolant is an indication of a problem in the PCS and contributes to increased post accident doses. If secondary specific activity cannot be restored to within limits in the associated Completion Time, the plant must be placed in a MODE in which the LCO does not apply. To achieve this status, the plant must be placed in at least MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and in MODE 5 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

SURVEILLANCE SR 3.7.17.1 REQUIREMENTS This SR ensures that the secondary specific activity is within the limits of the accident analysis. A gamma isotope analysis of the secondary coolant, which determines DOSE EQUIVALENT 1-131 , confirms the validity of the safety analysis assumptions as to the source terms in post accident releases. It also serves to identify and trend any unusual isotopic concentrations that might indicate changes in primary coolant activity or LEAKAGE. The 31 day Frequensy is based on the detestion Of ~

insreasing trends of the level of DOSE EQUIVALENT I 131, and allows for appropriate astion to be taken to maintain levels below the LeO limit. Insert 3 REFERENCES 1. 10 CFR 50.67

2. FSAR, Section 14.14 Palisades Nuclear Plant B 3.7.17-3 Amendment No. 226 Revised 04/14/2016

AC Sources - Operating B 3.8.1 BASES SURVEILLANCE SR 3.8.1.1 (continued)

REQUIREMENTS The 7 day Freql::lenGY is adeql::late beGal::lse disGonneGt switGh Positions~

Gannot Ghange without operator action and beGause their status is displayed in the Gontrol room. Insert 3 SR 3.8 .1.2 This SR helps to ensure the availability of the standby electrical power supply to mitigate DBAs and transients and to maintain the plant in a safe shutdown condition .

The monthly test starting of the DG provides assurance that the DG would start and be ready for loading in the time period assumed in the safety analyses. The monthly test, however does not, and is not intended to, test all portions of the circuitry necessary for automatic starting and loading. The operation of the bus undervoltage relays and their auxiliary relays which initiate DG starting, the control relay, which initiates DG breaker closure, and the DG breaker closure itself are not verified by this test. Verification of automatic operation of these components requires de-energizing the associated 2400 V bus and cannot be done during plant operation. For this test, the 10-second timing is started when the DG receives a start signal, and ends when the DG voltage sensing relays actuate. For the purposes of SR 3.8.1.2, the DGs are manually started from standby conditions. Standby conditions for a DG mean the diesel engine is not running, its coolant and oil temperatures are being maintained consistent with manufacturer recommendations, and ~ 20 minutes have elapsed since the last DG air roll.

Three relays sense the terminal voltage on each DG. These relays, in conjunction with a load shedding relay actuated by bus undervoltage, initiate automatic closing of the DG breaker. During monthly testing, the actuation of the three voltage sensing relays is used as the timing point to determine when the DG is ready for loading .

The 31 day Freql::lenoy for performanGe of SR 3.8 .1.2 agrees ',',lith the~

originalliGensing basis for the Palisades plant.

Insert 3 Palisades Nuclear Plant B3.8.1-15 Revised 11/08/2012

AC Sources - Operating B 3.8 .1 BASES SURVEILLANCE SR 3.8.1.3 REQUIREMENTS (continued) This Surveillance verifies that the DGs are capable of synchronizing with the offsite electrical system and accepting loads greater than or equal to the equivalent of the maximum expected accident loads for at least 15 minutes. A minimum total run time of 60 minutes is required to stabilize engine temperatures .

During the period when the DG is paralleled to the grid, it must be considered inoperable. This is because there are no provisions to automatically shift the DG controls from parallel mode to unit mode.

Additionally, when paralleled, there are certain conditions where the protection schemes may not prevent DG overloading and subsequent breaker trip and lockout.

The 31 day Frequency for this Surveillance is consistent with the original Palisades licensing basis.

The SR is modified by three Notes. Note 1 states that momentary transients outside the required band do not invalidate this test. This is to assure that a minor change in grid conditions and the resultant change in DG load, or a similar event, does not result in a surveillance being unnecessarily repeated. Note 2 indicates that this Surveillance should be conducted on only one DG at a time in order to avoid common cause failures that might result from offsite circuit or grid perturbations. Note 3 stipulates a prerequisite requirement for performance of this SR. A successful DG start must precede this test to credit satisfactory performance.

SR 3.8.1.4 This SR provides verification that the level of fuel oil in the day tank is at or above the level at which fuel oil is automatically added . The specified level is adequate for a minimum of 13.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months of DG operation at full load.

The 31 day Frequency is adequate to assure that a sufficient supply Of ~

fuel oil is a'Jailable, since low lelJel alarms are prolJided and plant operators would be aware of any uses of the DG during this period . Insert 3 Palisades Nuclear Plant B 3.8.1-16 Revised 11/08/2012

AC Sources - Operating B 3.8.1 BASES SURVEILLANCE SR 3.8.1.5 REQUIREMENTS (continued) Each DG is provided with an engine overspeed trip to prevent damage to the engine. The loss of a large load could cause diesel engine overspeed, which, if excessive, might result in a trip of the engine. This Surveillance demonstrates the DG load response characteristics and capability to reject the largest single load without exceeding predetermined voltage and frequency and while maintaining a specified margin to the overspeed trip. This Surveillance may be accomplished with ttie DG in the "Parallel" mode.

An acceptable method is to parallel the DG with the grid and load the DG to a load equal to or greater than its single largest post-accident load. The DG breaker is tripped while its voltage and frequency (or speed) are being recorded. The time, voltage, and frequency tolerances specified in this SR are derived from the recommendations of RG 1.9, Revision 3 (Ref. 5).

RG 1.9 (Ref. 5) recommends that the increase in diesel speed during the transient does not exceed 75% of the difference between synchronous speed and the overspeed trip setpoint, or 15% above synchronous speed, whichever is lower. The Palisades DGs have a synchronous speed of 900 rpm and an overspeed trip setting range of 1060 to 1105 rpm. Therefore, the maximum acceptable transient frequency for this SR is 68 Hz.

The minimum steady state voltage is specified to provide adequate margin for the switchgear and for both the 2400 and 480 V safeguards motors; the maximum steady state voltage is 2400 +10% V as recommended by RG 1.9 (Ref. 5) .

The minimum acceptable frequency is specified to assure that the safeguards pumps powered from the DG would supply adequate flow to meet the safety analyses. The maximum acceptable steady state frequency is slightly higher than the +2% (61.2 Hz) recommended by RG 1.9 (Ref. 5) because the test must be performed with the DG controls in the Parallel mode. The increased frequency allowance of 0.3 Hz is based on the expected speed differential associated with performance of the test while in the "Parallel" mode.

The 18 month surveillance Frequency is consistent with the recommendation of RG 1.9 (Ref. 5).

Palisades Nuclear Plant B 3.8.1-17 Revised 11/08/2012

AC Sources - Operating B 3.8.1 BASES SURVEILLANCE SR 3.8.1.6 REQUIREMENTS (continued) This Surveillance demonstrates the DG capability to reject a full load without overspeed tripping or exceeding the predetermined voltage limits. The DG full load rejection may occur because of a system fault or inadvertent breaker tripping . This Surveillance ensures proper engine and generator load response under a complete loss of load.

These acceptance criteria provide DG damage protection . The 4000 V limitation is based on generator rating of 2400/4160V and the ratings of those components (connecting cables and switchgear) that would experience the voltage transient. While the DG is not expected to experience this transient during an event and continue to be available, this response ensures that the DG is not degraded for future application, including re-connection to the bus if the trip initiator can be corrected or isolated.

In order to ensure that the DG is tested under load conditions that are as close to design basis conditions as possible, yet still provide adequate testing margin between the specified power factor limit and the DG design power factor limit of 0.8, testing must be performed using a power factor::; 0.9. This is consistent with RG 1.9 (Ref. 5).

The 18 month Frequency is consistent with the recommendation of ~

RG 1.9 (Ref. 5) and is intended to be consistent with expected fuel cycle lengths. Insert 3 SR 3.8.1 .7 As recommended by RG 1.9 (Ref. 5) this Surveillance demonstrates the as designed operation of the standby power sources during loss of the offsite source. This test verifies all actions encountered from the loss of offsite power, including shedding of the nonessential loads and re-energizing of the emergency buses and respective loads from the DG.

The requirement to energize permanently connected loads is met when the DG breaker closes, energizing its associated 2400 V bus.

Permanently connected loads are those that are not disconnected from the bus by load shedding relays. They are energized when the DG breaker closes. It is not necessary to monitor each permanently connected load.

Palisades Nuclear Plant B3.8.1-18 Revised 11/08/2012

AC Sources - Operating B 3.8.1 BASES SURVEILLANCE SR 3.8.1.7 (continued)

REQUIREMENTS The DG auto-start and breaker closure time of 10 seconds is derived from requirements of the accident analysis to respond to a design basis large break LOCA. For this test, the 10-second timing is started when the DG receives a start signal, and ends when the DG breaker closes.

The safety analyses assume 11 seconds from the loss of power until the bus is re-energized .

The requirement to verify that auto-connected shutdown loads are energized refers to those loads that are actuated by the Normal Shutdown Sequencer. Each load should be started to assure that the DG is capable of accelerating these loads at the intervals programmed for the Normal Shutdown Sequence. The sequenced pumps may be operating on recirculation flow.

The requirements to maintain steady state voltage and frequency apply to the "steady state" period after all sequenced loads have been started . This period need only be long enough to achieve and measure steady voltage and frequency.

The Surveillance should be continued for a minimum of 5 minutes in order to demonstrate that all starting transients have decayed and stability has been achieved. The requirement to supply permanently connected loads for ~ 5 minutes, refers to the duration of the DG connection to the associated safeguards bus. It is not intended to require that sequenced loads be operated throughout the 5-minute period. It is not necessary to monitor each permanently connected load.

The requ irement to verify the connection and supply of permanently and automatically connected loads is intended to demonstrate the DG loading logic. This testing may be accomplished in any series of sequential, overlapping, or total steps so that the required connection and loading sequence is verified.

The Frequency of 18 months is consistent with the recommendations of ~

RG 1.9 (Ref. 5) .

This SR is modified by a Note. The reason for the Note is that Insert 3 performing the Surveillance would remove a required offsite circuit from service, perturb the electrical distribution system, and challenge safety systems.

Palisades Nuclear Plant B 3.8.1-19 Revised 11/08/2012

AC Sources - Operating B 3.8 .1 BASES SURVEILLANCE SR 3.8.1.8 REQUIREMENTS (continued) RG 1.9 (Ref. 5) recommends demonstration onGe per 18 months that the DGs can start and run continuously at full load capability for an interval of not less than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , ~ 120 minutes of which is at a load above its analyzed peak accident loading and the remainder of the time at a load equivalent to the continuous duty rating of the DG. SR 3.8 .1.8 only requires ~ 100 minutes at a load above the DG analyzed peak accident loading. The 100 minutes required by the SR satisfies the intent of the recommendations of the RG, but allows some tolerance between the time requirement and the DG rating . Without this tolerance, the load would have to be reduced at precisely 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months to satisfy the SR without exceeding the manufacturer's rating of the DG.

The DG starts for this Surveillance can be performed either from standby or hot conditions.

In order to ensure that the DG is tested under load conditions that are as close to design conditions as possible, yet still provide adequate testing margin between the specified power factor limit and the DG design power factor limit of 0.8, testing must be performed using a power factor of ~ 0.9. The load band is provided to avoid routine overloading of the DG. Routine overloading may result in more frequent inspections in accordance with vendor recommendations in order to maintain DG OPERABILITY.

In addition, a Note to the SR states that momentary transients outside the required band do not invalidate this test. This is to assure that a minor change in grid conditions and the resultant change in DG load, or a similar event, does not result in a surveillance being unnecessarily repeated.

During the period when the DG is paralleled to the grid, it must be considered inoperable. This is because there are no provisions to automatically shift the DG controls from parallel mode to unit mode.

Additionally, when paralleled, there are certain conditions where the protection schemes may not prevent DG overloading and subsequent breaker trip and lockout.

The 18 month Frequency is Gonsistent with the reGommenaations of RG 1.9 (Ref. 5).

Palisades Nuclear Plant B 3.8.1-20 Revised 11/08/2012

AC Sources - Operating B 3.8 .1 BASES SURVEILLANCE SR 3.8.1.9 REQUIREMENTS (continued) As recommended by RG 1.9 (Ref. 5), this Surveillance ensures that the manual synchronization and load transfer from the DG to the offsite source can be made and that the DG can be returned to ready to load status when offsite power is restored . The test is performed while the DG is supplying its associated 2400 V bus, but not necessarily carrying the sequenced accident loads. The DG is considered to be in ready to load status when the DG is at rated speed and voltage, the output breaker is open, the automatic load sequencer is reset, and the DG controls are returned to "Unit."

During the period when the DG is paralleled to the grid, it must be considered inoperable. This is because there are no provisions to automatically shift the DG controls from parallel mode to unit mode.

Additionally, when paralleled, there are certain conditions where the protection schemes may not prevent DG overloading and subsequent breaker trip and lockout.

The Frequency of 18 months is consistent with the recommendations of ~

RG 1.9 (Ref. 5) .

This SR is modified by a Note. The reason for the Note is that Insert 3 performing the Surveillance would remove a required offsite circuit from service, perturb the electrical distribution system, and challenge safety systems.

Palisades Nuclear Plant B 3.8.1-21 Revised 11/08/2012

AC Sources - Operating B 3.8.1 BASES SURVEILLANCE SR 3.8.1 .10 REQUIREMENTS (continued) If power is lost to bus 1C or 1D, loads are sequentially connected to the bus by the automatic load sequencer. The sequencing logic controls the permissive and starting signals to motor breakers to prevent overloading of the DGs by concurrent motor starting currents. The 0.3-second load sequence time tolerance ensures that sufficient time exists for the DG to restore frequency and voltage prior to applying the next load and ensures that safety analysis assumptions regarding ESF equipment time delays are met. Logic Drawing E-17 Sheet 4 (Ref. 7) provides a summary of the automatic loading of safety related buses.

This SR is modified by a Note. The reason for the Note is that performing the Surveillance would remove a required offsite circuit from service, perturb the electrical distribution system, and challenge safety systems.

SR 3.8.1.11 In the event of a DBA coincident with a loss of offsite power, the DGs are required to supply the necessary power to ESF systems so that the fuel, PCS, and containment design limits are not exceeded .

The requirement to energize permanently connected loads is met when the DG breaker closes, energizing its associated 2400 V bus.

Permanently connected loads are those that are not disconnected from the bus by load shedding relays. They are energized when the DG breaker closes. It is not necessary to monitor each permanently connected load. The DG auto-start and breaker closure time of 10 seconds is derived from requirements of the accident analysis to respond to a design basis large break LOCA. For this test, the 10-second timing is started when the DG receives a start signal, and ends when the DG breaker closes. The safety analyses assume 11 seconds from the loss of power until the bus is re-energized .

Palisades Nuclear Plant B 3.8.1-22 Revised 11/08/2012

AC Sources - Operating B 3.8.1 BASES SURVEILLANCE SR 3.8.1.11 (continued)

REQUIREMENTS In addition, a Note to the SR states that momentary transients outside the required band do not invalidate this test. This is to assure that a minor change in grid conditions and the resultant change in DG load, or a similar event, does not result in a surveillance being unnecessarily repeated.

The requirement to verify that auto-connected shutdown loads are energized refers to those loads that are actuated by the DBA Sequencer. Each load should be started to assure that the DG is capable of accelerating these loads at the intervals programmed for the DBA Sequence. Since the containment spray pumps do not actuate on SIS generated by Pressure Low Pressure, the test should be performed such that spray pump starting by the sequencer is also verified along with the other SIS loads. The sequenced pumps may be operating on recirculation flow or in other testing modes. The requirements to maintain steady state voltage and frequency apply to the "steady state" period after all sequenced loads have been started. This period need only be long enough to achieve and measure steady voltage and frequency .

The Surveillance should be continued for a minimum of 5 minutes in order to demonstrate that all starting transients have decayed and stability has been achieved. The requirement to supply permanently connected loads for ~ 5 minutes, refers to the duration of the DG connection to the associated 2400 V bus. It is not intended to require that sequenced loads be operated throughout the 5-minute period. It is not necessary to monitor each permanently connected load.

The Frequency of 18 months takes into consideration plant conditions ~

required to perform the Surveillance and is intended to be consistent with an expected fuel cycle length of 18 months. Insert 3 This SR is modified by a Note. The reason for the Note is that performing the Surveillance would remove a required offsite circuit from service, perturb the electrical distribution system, and challenge safety systems.

Palisades Nuclear Plant B 3.8.1-23 Revised 11/08/2012

Diesel Fuel, Lube Oil, and Starting Air B 3.8.3 BASES SURVEILLANCE SR 3.8.3.1 REQUIREMENTS This SR provides verification that there is an adequate inventory of fuel oil in the storage subsystem to support either DG's operation for 7 days at full post-accident load. The fuel oil inventory equivalent to a 7 day supply is 33,054 gallons (Ref. 5) when calculated in accordance with References 1 and 2 . This inventory is conservatively based on an uprated 2600 kW DG capacity. The required fuel storage volume is determined using the most limiting energy content of the stored fuel.

Using the known correlation of diesel fuel oil absolute specific gravity or API gravity to energy content, the required diesel generator output, and the corresponding fuel consumption rate, the onsite fuel storage volume required for 7 days of operation can be determined. SR 3.8.3.3 requires new fuel to be tested to verify that the absolute specific gravity or API gravity is not less than the value assumed in the diesel fuel oil consumption calculations. The 7 day period is sufficient time to place the plant in a safe shutdown condition and to bring in replenishment fuel from an offsite location.

The 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Frequency is specified to ensure that a sufficient supply of ~

fuel oil is available, since the Fuel Oil Storage Tank is the fuel oil supply for the diesel fire pumps, heating and evaporator boilers, in addition to I rt 3 the DGs. nse SR 3.8.3.2 This Surveillance ensures that sufficient stored lube oil inventory is available to support at least 7 days of full accident load operation for one DG . The lube oil inventory equivalent to a 7 day supply is 313 gallons and is based on an estimated consumption of 1.0% of fuel oil consumption (Ref. 5). This inventory is also conservatively based on an uprated 2600 kW DG capacity.

A 31 day Frequency is adequate to ensure that a sufficient lube oil ~

supply is onsite, since DG starts and run times are closely monitored by the plant staff. Insert 3 SR 3.8.3.3 The tests listed below are a means of determining whether new fuel oil and stored fuel oil are of the appropriate grade and have not been contaminated with substances that would have an immediate, detrimental impact on diesel engine combustion.

Testing for viscosity, specific gravity, and water and sediment is completed for fuel oil delivered to the plant prior to its being added to the Fuel Oil Storage Tank. Fuel oil which fails the test, but has not been Palisades Nuclear Plant B 3.8.3-5 Revised 08/08/2017

Diesel Fuel, Lube Oil, and Starting Air B 3.8.3 BASES SURVEILLANCE SR 3.8.3.3 (continued)

REQUIREMENTS added to the Fuel Oil Storage Tank does not imply failure of this SR and requires no specific action . If results from these tests are within acceptable limits, the fuel oil may be added to the storage tank without concern for contaminating the entire volume of fuel oil in the storage tank.

Fuel oil is tested for other of the parameters specified in ASTM 0975 (Ref. 3) in accordance with the Fuel Oil Testing Program required by Specification 5.5.11. Fuel oil determined to have one or more measured parameters, other than viscosity or water and sediment, outside acceptable limits will be evaluated for its effect on OG operation.

Fuel oil which is determined to be acceptable for short term OG operation, but outside limits will be restored to within limits in accordance with LCO 3.8.3 Condition F.

SR 3.8.3.4 This Surveillance ensures that, without the aid of the refill compressor, sufficient air start capacity for each OG is available. The pressure specified in this SR is intended to reflect the acceptable margin from which successful starts can be accomplished.

The 31 day FreEluency takes into account the capacity, capal3ility, ~

redundancy, and diversity of the /I.e sources and other indications availal3le in the control room, including alarms, to alert the operator to l3elow normal air start pressure. Insert 3 SR 3.8.3.5 Microbiological fouling is a major cause of fuel oil degradation. There are numerous bacteria that can grow in fuel oil and cause fouling, but all must have a water environment in order to survive. Removal of water from the Fuel Oil Storage Tank once every 92 days eliminates the necessary environment for bacterial survival. This is the most effective means of controlling microbiological fouling . In addition , it reduces the potential for water entrainment in the fuel oil during OG operation.

Water may come from any of several sources, including condensation, ground water, rain water, contaminated fuel oil, and from breakdown of the fuel oil by bacteria. Frequent checking for and removal of accumulated water minimizes fouling and provides data regarding the watertight integrity of the fuel oil system. The Surveillance FreEluencies ~

and acceptance criteria are estal3lished in the Fuel Oil Testing Program eased, in part, on those recommended l3y RG 1.137 (Ref. 1). This SR I 3 is for preventative maintenance. nsert Palisades Nuclear Plant B 3.8.3-6 Revised 08/08/2017

Diesel Fuel, Lube Oil, and Starting Air B 3.8.3 BASES SURVEILLANCE SR 3.8.3.5 (continued)

REQUIREMENTS The presence of water does not necessarily represent failure of this SR provided the accumulated water is removed in accordance with the requirements of the Fuel Oil Testing Program.

SR 3.8.3.6 This SR demonstrates that the fuel transfer systems can, as applicable, automatically and manually transfer fuel from the Fuel Oil Storage Tank to each day tank, and automatically from each day tank to each engine mounted tank. Automatic or manual transfer of fuel oil is required to support continuous operation of standby power sources.

This SR provides assurance that the following portions of the fuel transfer system are OPERABLE:

a. Fuel transfer pumps;

b. Day and engine mounted tank filling solenoid valves;

c. Day tank fill via automatic level controls or manual operation; and

d. Engine mounted tank fill via automatic level controls.

The 92 day Frequency corresponds to the testing requirements for ~

pumps in the ASME Code,Section XI (Ref. 4). Additional assurance of fuel transfer system OPERABILITY is provided during the monthly starting and loading tests for each DG when the fuel oil system will Insert 3 function to maintain level in the day and engine mounted tanks.

REFERENCES 1. Regulatory Guide 1.137

2. ANSI N195-1976

3. ASTM Standards, 0975, Table 1

4. ASME, Boiler and Pressure Vessel Code,Section XI

5. Engineering Analysis EA-EC6432-01 Palisades Nuclear Plant B 3.8.3-7 Revised 08/08/2017

DC Sources - Operating B 3.8.4 BASES SURVEILLANCE SR 3.8.4.1 REQUIREMENTS Verifying battery terminal voltage while on float charge helps to ensure the effectiveness of the charging system and the ability of the batteries to perform their intended function. Float charge is the condition in which the charger is supplying the continuous current required to overcome the internal losses of a battery and maintain the battery in a fully charged state. The specified voltage is the nominal rating of the battery. Surveillance voltage measurements may be adjusted for cable losses and for installed plant instrumentation to ensure that battery terminal voltage requirements are satisfied. At that terminal voltage, the battery has sufficient charge to provide the analyzed capacity for either accident loading or station blackout loading. The 7 day Frequency is ~

consistent with manufacturer and IEEE 4§Q (Ref. 4) recommendations.

SR 3.8.4.2 Insert 3 Visual inspection to detect corrosion of the battery terminals and connectors, or measurement of the resistance of each inter-cell and terminal connection, provides an indication of physical damage or abnormal deterioration that could potentially degrade battery performance.

The specified limits of S 50 1J0hm for inter-cell connections and terminal connections, and S 360 1J0hms for inter-tier and inter-rack connections are in accordance with the manufacturers recommendations. The 50 1J0hm value is based on the minimum battery design voltage.

Battery sizing calculations show the first minute load on the ED-02 battery as the load that determines battery size, hence, battery voltage will be at its lowest value while the battery supplies this current.

Calculations also show that at a minimum temperature and end of life (80% battery performance), battery voltage during this first minute load will be about 1.815 V per cell, assuming nominal connection resistance.

But if all the connections were at the ceiling value of 50 1J0hms, the battery manufacturer indicates that the additional voltage drop would result in a battery voltage of about 1.79 V per cell, which is still above the minimum design voltage (Ref. 5) .

The 360 1J0hm value is based on 120% of the nominal cumulative resistance of the components which make up the connections:

resistance of the connecting cable, and for each end of the cable, the battery post to cable lug connection, the cable lug itself, and the lug to cable connection.

Palisades Nuclear Plant B 3.8.4-5 Revised 07/13/2006

DC Sources - Operating B 3.8.4 BASES SURVEILLANCE SR 3.8.4.2 (continued)

REQUIREMENTS The resistance values determined during initial battery installation are recorded with the battery replacement specifications, FES 95-206-ED-01 and FES 95-206-ED-02.

The Surveillance Frequency for these inspections, which can Eletect ~

conElitions that can cause power losses Elue to resistance heating, is 92 Elays. This Frequency is consiElereEl acceptable baseEl on operating I t3 experience relateEl to Eletecting corrosion trenEls. nser SR 3.8.4.3 Visual inspection of the battery cells, cell plates, and racks provides an indication of physical damage or abnormal deterioration that could potentially degrade battery performance. The presence of physical damage or deterioration does not necessarily represent a failure of this SR, provided an evaluation determines that the physical damage or deterioration does not affect the OPERABILITY of the battery (its ability to perform its design function) .

The 12 month Frequency for this SR is consistent with IEEE 450 ~

(Ref. 4), hich recommenEls EletaileEl visual inspection of cell conElition anEl rack integrity on a yearly basis. Insert 3 SR 3.8.4.4 and SR 3.8.4.5 Visual inspection and resistance measurements of inter-cell and terminal connections provide an indication of phYSical damage or abnormal deterioration that could indicate degraded battery condition.

The anticorrosion material is used to help ensure good electrical connections and to reduce terminal deterioration. The visual inspection for corrosion is not intended to require removal of and inspection under each terminal connection. The removal of visible corrosion is a preventive maintenance SR. The presence of visible corrosion does not necessarily represent a failure of this SR provided visible corrosion is removed during performance of SR 3.8.4.4.

The specified limits for connection resistance are discussed in the Bases for SR 3.8.4.2.

Palisades Nuclear Plant B 3.8.4-6 Revised 07/13/2006

DC Sources - Operating B 3.8.4 BASES SURVEILLANCE SR 3.8.4.4 and SR 3.8.4.5 (continued)

REQUIREMENTS The S~p/eillance FFe~~encies of 12 FAonths is consistent with IEEE 450~

(Ref. 4), which recoFAFAends cell to cell and terFAinal connection resistance FAeaS~FeFAent on a yea Fly basis. Insert 3 SR 3.8.4.6 This SR requires that each required battery charger be capable of supplying 180 amps at 125 V for ~ 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . These requirements are based on the design capacity of the chargers. The chargers are rated at 200 amps; the specified 180 amps provides margin between the charger rating and the test requirement.

The specified FFe~~ency Fe~~iFes each Fe~~iFed batte!), chaFgeF to be ~

tested each 18 FAonths. The S~p/eillance FFe~~ency is acceptable, given the otheF adFAinistrative contFOls existing to ens~re ade~~ate chaFgeF peFfoFFAanCe d~Fing these 18 FAonth intervals. In addition, this Insert 3 FFe~~ency is intended to be consistent with expected f~el cycle lengths.

SR 3.8.4.7 A battery service test is a special test of battery capability, as found, to satisfy the design requirements (battery duty cycle) of the DC electrical power system. The discharge rate and test length should correspond to the design duty cycle requirements as specified in FSAR Chapter 8 (Ref. 2).

The S~p/eillance Fre~~ency of 18 FAonths is consistent with the ~

FecoFAFAendations of RG 1.32 (Ref. 6) and RG 1.129 (Ref. 7), which state that the batte!), sep/ice test sho~ld be peFfoFFAed d~Fing ref~eling Insert 3 opeFations, OF at SOFAe otheF o~tage , with intervals between tests not to exceed 18 FAonths.

Either the battery performance discharge test or the modified performance discharge test is acceptable for satisfying SR 3.8.4.8; however, only the modified performance discharge test may be used to satisfy SR 3.8.4.8 while satisfying the requirements of SR 3.8.4.7 at the same time.

Palisades Nuclear Plant B 3.8.4-7 Revised 07/13/2006

DC Sources - Operating B 3.8.4 BASES SURVEILLANCE SR 3.8.4.8 (continued)

REQUIREMENTS The acceptance criteria for this Surveillance are consistent with the recommendations of IEEE-450 (Ref. 4) and IEEE-485 (Ref. 3). These references recommend that the battery be replaced if its capacity is below 80% of the manufacturer rating. A capacity of 80% shows that the battery rate of deterioration is increasing, even if there is ample capacity to meet the load requirements.

The Surveillance Frequency for this test is normally 60 mont~ s. he battery shows degradation, or if the battery has reached 85% .

expected life and capacity is < 100% of the manufacturer's r I rt 3 Surveillance Frequency is reduced to 12 months. However, nse battery shows no degradation but has reached 85% of its expected life, the Surveillance Frequency is only reduced to 24 months for batteries that retain capacity ~ 100% of the manufacturer's rating. Degradation is indicated, according to IEEE-450 (Ref. 4), when the battery capacity drops by more than 10% relative to its capacity on the previous performance test or when it is ~ 10% below the manufacturer's rating .

These Frequencies are consistent with the recommendations in IEEE-450 (Ref. 4).

The reason for the restriction that the plant be outside of MODES 1, 2, 3, and 4 is that performing the Surveillance requires disconnecting the battery from the DC distribution buses and connecting it to a test load resistor bank. This action makes the battery inoperable and completely unavailable for use.

REFERENCES 1. 10 CFR 50, Appendix A, GDC 17

2. FSAR, Chapter 8

3. IEEE-485-1983, June 1983

4. IEEE-450-1995

5. Letter; Graham Walker, C&D Charter Power Systems, Inc to John Slinkard, Consumers Power Company, 12 July 1996

6. Regulatory Guide 1.32, February 1977

7. Regulatory Guide 1.129, December 1974 Palisades Nuclear Plant B 3.8.4-9 Revised 07/13/2006

Battery Cell Parameters B 3.8.6 BASES ACTIONS (continued)

With the temperature of representative cells below the design temperature, or with one or more battery cells with parameters outside the Category C limits, sufficient capacity to supply the maximum expected load requirement is not assured and the corresponding battery must be declared inoperable.

Additionally, if battery cells cannot be restored to meeting Category A or B limits within 31 days, a serious difficulty with the battery is indicated and the battery must be declared to be inoperable.

SURVEILLANCE SR 3.8.6.1 REQUIREMENTS This SR verifies that Category A battery cell parameters are consistent with IEEE-4S0 (Ref. 1), which recommends regular battery inspections (at least one per month) including voltage, specific gravity, and electrolyte temperature of pilot cells. ~

SR 3.8.6.2 I 3 nsert This Surveillance verification that the average temperature of representative cells is ~ 70°F is consistent with a recommendation of IEEE-4S0 (Ref. 1), whish states that the temperature of elestrolytes in representative sells should be determined on a quarterly basis. The ~

monthly frequensy spesified is a feature of the initial Palisades lisense, and is the same as those other pilot sell tests spesified in SR d.8.6.1. I 3 nsert Lower than normal temperatures act to inhibit or reduce battery capacity. This SR ensures that the operating temperatures remain within an acceptable operating range. This limit is based on manufacturer recommendations.

SR 3.8.6.3 The quarterly inspection of specific gravity and voltage is consistent with the recommendations of IEEE-4S0 (Ref. 1). ~

~

Palisades Nuclear Plant B 3.8.6-3 Amendment No. 189 Revised 08/09/2000

Inverters - Operating B 3.8.7 BASES SURVEILLANCE SR 3.8.7.1 REQUIREMENTS This Surveillance verifies that the inverters are functioning properly and energizing the Preferred AC buses. The verification of proper voltage and frequency output ensures that the required power is readily available for the instrumentation of the RPS and ESF connected to the Preferred AC buses. The7 da}L1=requenc}Liakes 1nto~t ...,

~=~::::::in the contr~4"OOR14hatmert Inv~mal.functtons.

th&operator toI ~

Insert 3 REFERENCES None Palisades Nuclear Plant B 3.8.7-3 Amendment No. 189

Inverters - Shutdown B 3.8.8 BASES ACTIONS A.2.1, A.2.2, A.2 .3, and A.2.4 (continued)

These ACTIONS minimize the probability or the occurrence of postulated events. It is further required to immediately initiate action to restore the required inverters (and to continue this action until restoration is accomplished) in order to provide the required inverter supplied Preferred AC power to the plant instrument and control systems.

The Completion Time of "immediately" is consistent with the required times for actions requiring prompt attention. The restoration of the required inverters should be completed as quickly as possible in order to minimize the time during which the plant safety systems may be without inverter supplied Preferred AC power.

SURVEILLANCE SR 3.8.8.1 REQUIREMENTS 1

A description of the basis for this SR is provided in the Bases for SR 3.8.7.1.

Insert 3 REFERENCES None Palisades Nuclear Plant B 3.8.8-3 Revised 11/06/2001

Distribution Systems - Operating B 3.8.9 BASES SURVEILLANCE SR 3.8.9.1 REQUIREMENTS This surveillance verifies that the required AC, DC, and Preferred AC bus electrical power distribution subsystems are functioning properly, with the correct circuit breaker alignment. The correct breaker alignment ensures the appropriate separation and independence of the electrical divisions is maintained.

For those buses which have undervoltage alarms in the control room, correct voltage may be verified by the absence of an undervoltage alarm.

For those buses which have only one possible power source and have undervoltage alarms in the control room, correct breaker alignment may be verified by the absence of an undervoltage alarm.

A Preferred AC Bus may be considered correctly aligned when powered from either the associated inverter or from the bypass regulator. A mechanical interlock prevents connecting two or more Preferred AC Buses to the Bypass Regulator. LCO 3.8.7 and LCO 3.8.8 address the condition of supplying a Preferred AC Bus from the bypass regulator.

The 7 day Freql:Jency takes into accol:Jnt the redl:Jndant capability of the AC, DC, and Preferred AC bl:Js electrical power distribl:Jtion sl:Jbsystems, and other indications available in the control room that alert the operator 1

to sl:Jbsystem malfl:Jnctions.

Insert 3 REFERENCES None Palisades Nuclear Plant B 3.8.9-6 Revised 11/06/2001

Distribution Systems - Shutdown B 3.8.10 BASES ACTIONS A.2.1, A.2.2, A.2.3, A.2.4, and A.2.5 (continued)

The Completion Time of "immediately" is consistent with the required times for actions requiring prompt attention. The restoration of the required distribution subsystems should be completed as quickly as possible in order to minimize the time during which the plant safety systems may be without sufficient power.

SURVEILLANCE SR 3.8.10.1 REQUIREMENTS 1

A description of the basis for this SR is provided in the Bases for SR 3.8.9.1.

Insert 3 REFERENCES None Palisades Nuclear Plant B 3.8.10-3 Revised 11/06/2001

Boron Concentration B 3.9.1 BASES ACTIONS A.3 (continued)

Once boration is initiated, it must be continued until the boron concentration is restored. The restoration time depends on the amount of boron that must be injected to reach the required concentration.

SURVEILLANCE SR 3.9.1 .1 REQUIREMENTS This SR ensures the coolant boron concentration in the PCS and the refueling cavity is within the limit. The boron concentration of the coolant in each volume is determined periodically by chemical analysis.

A minimum Frequency of once every 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months is therefore a reasonable~

amount of time to \Ierify the boron concentration of representative samples. The Frequeney is based on operating experience, which has Insert 3 shown 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months to be adequate.

REFERENCES 1. FSAR, Section 5.1

2. FSAR, Section 14.3 Palisades Nuclear Plant B3.9.1-4 Amendment No 189 Revised 08/09/2000

Nuclear Instrumentation B 3.9.2 BASES ACTIONS B.2 (continued)

As stated in SR 3.0.2, the 25% extension allowed by SR 3.0.2 may be applied to Required Actions whose Completion Time is stated as "once per . . ." ..

.however, the 25% extension does not apply to the initial performance of a Required Action with a periodic Completion Time that requires performance on a "once per ... " basis. The 25% extension applies to each performance of the Required Action after the initial performance . .. .Therefore, while Required Action 3.9.2 B.2 must be initially performed within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months without any SR 3.0.2 extension, subsequent performances may utilize the 25% SR 3.0.2 extension.

SURVEILLANCE SR 3.9.2.1 REQUIREMENTS SR 3.9.2.1 is the performance of a CHANNEL CHECK, which is a comparison of the parameter indicated on one channel to a similar parameter on other channels. It is based on the assumption that the two indication channels should be consistent with core conditions, but does not require the two source range channels to have the same reading.

Changes in fuel loading and core geometry can result in significant differences between source range channels, but each channel should be consistent with its local conditions. The Frequency of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is ~

consistent with the CHANNEL CHECK Frequency specified similarly for the same instruments in LCO 3.3.9. Insert 3 SR 3.9.2.2 SR 3.9.2.2 is the performance of a CHANNEL CALIBRATION every 18 months. The CHANNEL CALIBRATION for the source range neutron flux monitors consists of obtaining the detector plateau or preamp discriminator curves, evaluating those curves, and comparing the curves to the manufacturer's data. 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 Insert 3 usually pass the Surveillance when performed on the 18 month Frequency.

REFERENCES 1. FSAR, Section 7.6

2. FSAR, Section 14.3 Amendment No 189 Palisades Nuclear Plant B 3.9.2-3 Revised 02/12/2001

Containment Penetrations B 3.9.3 BASES ACTIONS A.1 and A.2 (continued)

(continued)

This is accomplished by immediately suspending CORE ALTERATIONS and movement of irradiated fuel assemblies within containment. Performance of these actions shall not preclude completion of movement of a component to a safe position.

SURVEILLANCE SR 3.9.3.1 REQUIREMENTS This Surveillance demonstrates that each of the containment penetrations required to be in its closed position is in that position. The Surveillance on the valves in unisolated penetrations which provide a direct path from the containment atmosphere to the outside atmosphere will demonstrate that the valves are not blocked from closing. Also, the Surveillance will demonstrate that each valve operator has motive B

power, which will ensure each valve is capable of being closed by an OPERABLE Refueling Containment High Radiation signal.

The Surveillanoe is performed every 7 days during CORE ALTERATIONS or during movement of irradiated fuel assemblies within the containment. The Surveillanoe interval is seleoted to be Insert 3 oommensurate with the normal duration of time to oomplete fuel handling operations. As suoh, this Surveillanoe provides assuranoe that a postulated fuel handling aooident that releases fission produot radioaotivity within the oontainment will not result in an exoessive release of fission produot radioaotivity to the environment.

SR 3.9.3.2 This Surveillance demonstrates that each automatic isolation valve providing direct access from the containment atmosphere to the outside atmosphere valve actuates to its isolation position on an actual or simulated high radiation signal.

Palisades Nuclear Plant B 3.9.3-5 Amendment No. 226 Revised 04/14/2016

Containment Penetrations B 3.9.3 BASES SURVEILLANCE SR 3.9.3.2 (continued)

REQUIREMENTS (continued) The SR is modified by a Note which requires only the valves in un isolated penetrations to be tested . The 18 month F"requency maintains consistency with other similar ESF"AS instrumentation and valve testing requirements. lCO a.a.6, "Refueling Containment High

II

a CHANNEL F"UNCTION,I\l TEST every a1 days and a CHANNEL CALIBRATION every 18 months to ensure the channel OPERABILITY during refueling operations. These surveillances performed during MODE 6 will ensure that the valves are capable of closing after a postulated fuel handling accident to limit a release of fission product radioactivity from the containment.

REFERENCES 1. FSAR, Section 14.19 Palisades Nuclear Plant B 3.9.3-6 Amendment No. 226 Revised 04/14/2016

SOC and Coolant Circulation - High Water Level B 3.9.4 BASES ACTIONS (continued)

If SOC train requirements are not met, actions shall be taken immediately to suspend loading irradiated fuel assemblies in the core. With no forced circulation cooling, decay heat removal from the core occurs by natural circulation to the heat sink provided by the water above the core. A minimum refueling cavity water level equivalent to the 647 ft elevation provides an adequate available heat sink. Suspending any operation that would increase the decay heat load, such as loading a fuel assembly, is a prudent action under this condition.

If SOC train requirements are not met, all containment penetrations providing direct access from the containment atmosphere to the outside atmosphere must be closed to prevent fission products, if released by a loss of decay heat removal event, from escaping to the environment. The 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months Completion Time is based on the low probability of the coolant boiling in that time and allows time for fixing most SOC problems.

SURVEILLANCE SR 3.9.4.1 REQUIREMENTS This Surveillance demonstrates that the SOC train is in operation and circulating primary coolant. The flow rate is sufficient to provide decay heat removal capability and to prevent thermal and boron stratification in the core. The 1000 gpm flow rate has been determined by operating experience rather than analysis. The Frequency of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is SUfficient'B considering the flO'.v, temperature, pump control, and alarm indications available to the operator in the control room for monitoring the SDC Insert 3 System.

REFERENCES 1. FSAR, Sections 6.1 and 14.3 Palisades Nuclear Plant B 3.9.4-4 Revised 07/31/2007

SOC and Coolant Circulation - Low Water Level B 3.9.5 BASES SURVEILLANCE SR 3.9.5.1 REQUIREMENTS This Surveillance demonstrates that one SOC train is operating and circulating primary coolant. The flow rate is sufficient to provide decay heat removal capability and to prevent thermal and boron stratification in the core.

In addition, during operation of the SOC train with the water level in the vicinity of the reactor vessel nozzles, the SOC train flow rate determination must also consider the SOC pump suction requirements.

The 1000 gpm flow rate has been determined by operating experience rather than analysis. The Frequency of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is sufficient, ~

considering the flow, temperature, pump control, and alarm indications available to the operator to monitor the SOC System in the control Insert 3 fGGffi:.

SR 3.9.5.2 Verification that the required pump is OPERABLE ensures that an additional SOC pump can be placed in operation, if needed, to maintain decay heat removal and primary coolant circulation. Verification is performed by verifying proper breaker alignment and power available to the required pump. The Frequency of 7 days is considered reasonable ~

in view of other administrative controls available and has been sho'....n to be acceptable by operating m<perience. Insert 3 REFERENCES 1. FSAR, Sections 6.1 and 14.3 Palisades Nuclear Plant B 3.9.5-4 Revised 07/31/2007

Refueling Cavity Water Level B 3.9.6 BASES SURVEILLANCE SR 3.9.6.1 REQUIREMENTS Verification of a minimum water level corresponding to the 647 ft elevation ensures that the design basis for the postulated fuel handling accident analysis during refueling operations is met. Water at the required elevation limits the consequences of damaged fuel rods that are postulated to result from a fuel handling accident inside containment (Ref. 2) .

The Frequency of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months is based on engineering judgment and is ~

considered adequate in view of the large volume of water and the normal procedural controls of valve positions, which make significant Insert 3 unplanned level changes unlikely.

REFERENCES 1. Regulatory Guide 1.183

2. FSAR, Section 14.19 Palisades Nuclear Plant B 3.9.6-3 Amendment No. 226 Revised 04/14/2016

ML18270A326

Text

Navigation menu

Search