Text

Part 3 of 3, Prairie Island Supplement to License Amendment Request Dated December 11, 2000 Conversion to Improved Technical Specifications (ITS)
	ML021060604
Person / Time
Site:	Prairie Island
Issue date:	04/11/2002
From:	Nazar M; Nuclear Management Co
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	NUREG-1431
	Download: ML021060604 (185)
	v • d • e

Diesel Fuel Oil, Lube Oil, and Startling Air IPA3.8-100 3.8.3 c-145 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME B. One or more D*s with B.1 Restore lube oil 488hou.rs lube oil inventory inventory to withi Sgal.limits.

GB. One or more DGs GB.1 Restore fuel oil tank 7 days w-t-hRequired DG total particulates fuel oil tank with properties to within IcL3"8-146I stored fuel oil limit(s).

properti ese-t-a-I partie-&a not within limit(s).

(continued) r- I r* A A.L 9C. One or mforee D~s UL. I 3O days with new fuel oil oil properties to

,.,"-l-a , U *Iq- 4 6 4 L properties not within limits.

C.1 Isolate the 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months

+/--m-i-t-sRequi red associated DG fuel Action and oil tank. ICL3.8-146 associated S - I Completion Time of IR-12 ,

Condition B not L- - -.

met.

WOG STS Rev 1, 04/07/95 3.8.3-2 Markup for PI ITS Part E

Diesel Fuel Oil, Lube Oil and Starting Air 3.8.3 IPA3"8-100 IcL3"8-145 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME E. One or mre *s wi ,th starting air receiver E.1 Restore starting ai 48-hours pressure [225] psig reeeiver-presur to and -- [125]j psig..

DF. Stored DG fuel oil -------------- NOTE--------

supply: Enter applicable Conditions and Required Actions of LCO ICL3.8-146 Unit 1 < 36,000 3.7.8, "CL System" when gal ; Condition D is entered as a IPA3.8-218 result of stored fuel oil Unit 2 < 65,000 properties not within limits.

gal.

OR DF.1 Declare sseeiated Immediately DGs inoperable. _--j----I I I R-12 Required Action and L . . . . . .-

associated Completion Time of Conditions A r. ----- I or C not met.

R-12I L-------

OR

n* or more DGs diesel fuel oil, lube oil, or starting air subsystem not within limits for reasons other than Condition AIn,, C,D',

0rE.

SURVEILLANCE REQUIREMENTS WOG STS Rev 1, 04/07/95 3.8.3-3 Markup for PI ITS Part E

DC Sources - Operating 3.8.4 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME f B. One [er-two] battertyjfie8 B.1 Verify associated battery 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months oin-ne rtaiftj inoperable, charger is OPERABLE. ICL3.8-178 I AND B.2 Verify other train battery 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months is OPERABLE, AND ICL3.8-179 I B.3 Verify other train battery 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months charger is OPERABLE.

AND B.-24 Restore batterfy]{iesi to f2] 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months I OPERABLE status I 171 C. One DC electrical power C.1 Restore DC electrical 121 hours0.0014 days 0.0336 hours 2.000661e-4 weeks 4.60405e-5 months subsystem inoperable for power subsystem to reasons other than OPERABLE status.

Condition A-for B].

D. Required Action and D.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months Associated Completion Time not met. AND D.2 Be in MODE 5. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months R-12 WOG STS Rev. 2, 04/30/01 3.8.4- 2 Markup for PI ITS Part E

DC Sources - Operating 3.8.4 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.4.1 Verify battery terminal voltage is greater than or equal 7 days to the minimum established float voltage.

SR 3.8.4.2 Verify each battery charger supplies >_f413e-250 amps [--81 24 months at greater than or equal to the minimum established float voltage for >_-814 hours.

OR JX3.8-126 Verify each battery charger can recharge the battery to the fully charged state within --24 ]-hours while supplying the largest combined dem'ands of the various continu,,ous steady state loadsdemands of the various continuous steady state loads, after a battery ICL3.8-182 discharge to the bounding design basis event discharge state.

SR 3.8.4.3

- NOTES

1. The modified performance discharge test in SR 3.8.6.6 may be performed in lieu of SR 3.8.4.3.

2. This Surveillance shall not normally be performed in MODE 1, 2, 3, or 4. However, portions of the Surveillance may be performed to reestablish OPERABILITY provided an assessment determines the safety of the plant is maintained or enhanced.

Verify battery capacity is adequate to supply, and [--8124 months maintain in OPERABLE status, the required emergency loads for the design duty cycle when Ix3.8_126 subjected to a battery service test.

1 R-12 WOG STS Rev. 2, 04/30/01 3.8.4- 3 Markup for PI ITS Part E

DC Sources - Shutdown 3.8.5 3.8 ELECTRICAL POWER SYSTEMS 3.8.5 DC Sources - Shutdown LCO 3.8.5 [DG electrical power subsystem shall be OPERABLE to sutppodt the D, electrical power distribution subsystemn(s) requi'red by LCO) 3.8.10, "Distribution Systems Shutdown."]

fOne DC electrical power subsystem shall be OPERABLE.]

- REVIEWER'S NOTE "This second option above applies for plants having a pre"ITS licensing basis (OTS) for electrical power requirements during shutdown conditions that required only one DC electrical power subsystem c nto be OPEBLE.

Action A the bracketed optional wording in Condition B are also eliminated for this case. The first option above is adopted for plants that have a licensing basis (OTS) requtiring the same level of DG electrical powe~r subsystemn suppor as is requiored for power operating conditions.

NT Service Building DC electrical power subsystem components may be used to replace safeguards DC electrical power subsystem components when the required safeguards DC electrical power subsystem is,-------.

inoperable du-e-to testing, ma-intenance, or replacement. R-12 JPA3.8-212 J APPLICABILITY: MODES 5 and 6, During movement of-freeenAl irradiated fuel assemblies.

ACTIONS

- NOTE LCO 3.0.3 is not applicable.

WOG STS Rev. 2, 04/30/01 3.8.5- 1 Markup for PI ITS Part E

DC Sources - Shutdown 3.8.5 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME

[A. One for-twojrequired A." Restore battery terminal 2-h...

battery charger[*sonone voltage to greater than err

,raij inoperable, equ.al to the m. m u established float voltage.

A*D AND The redumedant trai A.2 Verify battery float 1urrent O11e per [12] hours battery and ehargerfs] -[-2] emps.

OPERABLE.

A+49 JPA3.8-198 A.31 Restore battery chargerfsj 7-days ]8 hours to OPERABLE status. ICL3.8-1 71I R-12 WOG STS Rev. 2, 04/30/01 3.8.5- 2 Markup for PI ITS Part E

DC Sources - Shutdown 3.8.5 CONDITION REQUIRED ACTION COMPLETION TIME B. One [eor-mere]-required B4 Deelare affected "m.ed.ate.. r....ired DC electrical power feature(s) incper..... P"9 subsystem[ inoperable PA3.8-199 tfor reasons other than GR Condition A B.2.1 Suspend CORE Immediately OR ALTERATIONS.

Required Action and AND associated Completion Time of Condition A not B-.2.2 Suspend movement of Immediately meti. [feeently] irradiated fuel assemblies.

AND B-.-3 Suspend operations Immediately involving positive reactivity additions that could result in loss of required SDM or boron concentration.

AND B-.2.4 Initiate action to restore Immediately required DC electrical power subsystems to OPERABLE status.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.5.1

- NOTE The following SRs are not required to be performed:

SR 3.8.4.2 and SR 3.8.4.3.

For DC sources required to be OPERABLE, the In accordance with following SRs are applicable: applicable SRs SR 3.8.4.1 SR 3.8.4.2 SR 3.8.4.3 r I.---.

L R-12 WOG STS Rev. 2, 04/30/01 3.8.5- 3 Markup for PI ITS Part E

Battery Parameters 3.8.6 3.8 ELECTRICAL POWER SYSTEMS 3.8.6 Battery Parameters

- PFx,'MirwrP' I.- VI - l*V

ý h'nfTF-Licernsees mffust ifplem-ent a progr=am, as-pecifi-at.on specified in 5.5.17, to monitor battery r.~r ~ *htra ItLi,f :Q~ kLr'.SAUJ h, -r-ý An L -~LI r-ý...-i'-,-,~-~+--

J 11 1/4 I.CII4 1 .4 L.......%JO r-rC C c~i ~

C1"

'..OIU C1 A At..L 4(CQ*IrIr

3 I ----- I /* __!-!

Batteries For Stationary.Applications-.!

LCO 3.8.6 Battery parameters for Train A and Train B batteries shall be within limits.

APPLICABILITY: When associated DC electrical power subsystems are required to be r ... . I. . .I OPERABLE. IR-12 I ACTIONS

- NOTE Separate Condition entry is allowed for each battery.

CONDITION REQUIRED ACTION COMPLETION TIME A. One [or two] batterfyfries A.1 Perform SR 3.8.4.1. 28 hours3.240741e-4 days 0.00778 hours 4.62963e-5 weeks 1.0654e-5 months on o train] with one or more battery cells float AND JPA3.8-171 voltage <-J2.07j-V.

A.2 Perform SR 3.8.6.1. 28 hours3.240741e-4 days 0.00778 hours 4.62963e-5 weeks 1.0654e-5 months AND A.3 Restore affected cell 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months voltage >_f2.071 V.

B. One [or two] batterfy][ies B. 1 Perform SR 3.8.4.1. 28 hours3.240741e-4 days 0.00778 hours 4.62963e-5 weeks 1.0654e-5 months en one-tran] with float current > f2:-amps. AND R-12 WOG STS Rev. 2, 04/30/01 3.8.6- 1 Markup for PI ITS Part E

Diesel Fuel Oil , Lube Oil , and Starting Ai r B 3.8.3 JPA3.8-100 JCL3.8-145 BASES Surveillance Frequency intervals, and proper engine performance has been recently demonstrated (withi-n-3-1-days) it is prudent to allow a brief period prior to declaring the associated DG inoperable or isolating the associated fuel oil tank. Therefore, tiThe 7 day Completion Time allows for further evaluation, resampling and re-analysis of the DG fuel oil.

fDC. 1 With the new fuel oil -properties defined in the Bases fo1'r JCL3.8-146 SR 3.8.3.4 not within the required limits, a period of 30 days is allowed for restoring the stored fueloi properties. This period provides sufficient time to test the stored fuel oil to determine that the new fuel oil, when mixed with previously stored fuel oil, remains acceptable, orto restore the stored fuel oil properties. Thisý restoration may involve feed and bleed procedures, filtering, or combinations of these procedures. Even if a DC start and load was required during this time interval and the fuel oil properties were outside limits, there is a high likelihood that the BG would still be capable of performing its intended functiont.

With a Required Action and associated Completion Time ofICL3.8-146 Condition B not met, the associated fuel oil tank must be isolated within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months . Isolation of a specific fuel oil tank may not make the associated DG inoperable since the DG can take suction from another fuel oil tank. Isolation 121 of the associated fuel oil tank may cause entry intoL---

Conditions A or 0 which could result in the DG being inoperable.

(conti nued)

WOG STS Rev 1, 04/07/95 B 3.8.3-6 TSev, 0/07/5 WOG PI ITS Part E forB3.83-6Markup

DC Sources - Operating B 3.8.4 BASES ACTIONS (continued) batlery float current is not less than or equal to [2] amps this indicates there may be additional battery problems and the battery mustb declared inoperable.

Required Action A.3 requires, within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , that the ICL3.8-179 diesel generator and safeguards equipment on the other train are verified to be OPERABLE. This verification ensures that the redundant train is OPERABLE ensuring that the plant will be able to mitigate an event as analyzed in the USAR (Ref. 3). 1 Required Action A.34 limits the restoration time for the inoperable R-12 battery charger to 7-days8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . This action is applicable ia---

alternate means of restoring battery termfinal voltage to greater than 71 12~

or equal to the mninimum established float voltage has been used (e.g., balance of plant non-Class 1E battery chargo ,.The 7-day8 hour Completion Time reflects a reasonable time to effect restoration of the qualified battery charger to OPERABLE status.

B.1, B.2, B.3, and B.4 REVIEWER'S NOTE The 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months Completion Times of Requir.ed Actions 0-,.1 and C.1 are in brackets. Any licensee wishing to request a longer Completion TimeI need to demuonstrate that the longer Cmpletion Time is appropriate for the plant In accordance with the guidance in Regulatory Cuide (RC) 1.177, "An Approach for Plant Specific, Risk Informod Decisionmffaki ng:

Technical Spccifications.

Condition B represents onetrain-with-one fer-twNoibatterfy]fiesj inoperable.

With one [or two] batterfy-fies inoperable, the DC bus is being supplied by the OPERABLE battery charger-. Any event that results in a loss of the AC bus supporting the battery charger[sj will also result in loss of DC to that train. Recovery of the AC bus, especially if it is due to a loss of offsite power, will be hampered by the fact that many of the JPA3.8-203 components necessary for the recovery (e.g., diesel generator ICL3.8 ____

control and field flash, AC load shed and diesel generator output 0L3.8-178 circuit breakers, etc.) likely rely upon the batter-jy[iesi. In iadto-CL3.8_179 the energization transients of any DG loads that are beyondth cap ability

. of the battery charger[s] and normally require the CL3.8-171 assistance of the batter[yflies] will not be able to be broughtonie Required Actions B.1, B.2, and B,3 verify that the associated battery charger, the other train battery and associated charger are OPERABLE WOG STS Rev. 2, 04/30/01 B 3.8.4 - 6 Markup for PI ITS Part E

DC Sources - Operating B 3.8.4 BASES ACTIONS (continued) within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months . This time provides for either returning the inoperable battery to OPERABLE status or verifying that the associated charger and other train battery and charger are OPERABLE therefore, ensuring no loss of function exists. CL3.8-171 Required Action B.4 requires the inoperable battery to be restored to OPERABLE within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . The M2]8 hour limit allows sufficient time to effect restoration of an inoperable battery given that the majority of the conditions that lead to battery inoperability (e.g., loss of battery charger, battery cell voltage less than f2.071 V, etc.) are identified in Specifications 3.8.4, 3.8.5, and 3.8.6 together with additional specific completion times.

C.1 Condition C represents one train with a loss of ability to completely respond to an event, and a potential loss of ability to remain energized during normal operation. It is therefore, imperative that the operator's attention focus on stabilizing the unit, minimizing the potential for complete loss of DC power to the affected train. The 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months limit is consistent with the allowed time for an inoperable DC distribution system train.

If one of the required DC electrical power subsystems is inoperable for reasons other than Condition A or B (e.g., inoperable battery charger and associated inoperable battery), the remaining DC electrical power subsystem has the capacity to support a safe shutdown and to mitigate an accident condition. Since a subsequent worst- case single failure could, however, result in the loss of minimum necessary DC ICL3.8-172 electrical subsystems to mitigate a worst case accident, continued power operation should not exceed 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months . The 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months Completion Time is based on Regulatory Guide 1..93 (Ref. 7) and reflects a reasonable time to assess unit status as a function of the inoperable DC electrical power subsystem and, if the DC electrical power subsystem is not restored to OPERABLE status, to prepare to effect an orderly and safe unit shutdown.

D.1 and D.2 If the inoperable DC safeguards electrical power subsystem cannot be R-12 restored to OPERABLE status within the required Completion Time, the ------

unit must be brought to a MODE in which the LCO does not apply. To achieve this status, the unit 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 unit WOG STS Rev. 2, 04/30/01 B 3.8.4 - 7 Markup for PI ITS Part E

DC Sources - Operating B 3.8.4 BASES ACTIONS (continued)

The other option requires that each battery charger be capable of 0CL3.8-182 recharging the battery after a senicedischarge test coincident with I supplying the largest coincident demands of the various continuous steady state loads (irrespective of the status of the plant during which these dem.ands occur)demands of the various continuous steady state loads, after the battery discharge to the bounding design basis event discharge state. This level of loading may not norm,*ally be available following the battery service test and will need to be supplemented with additional 'lads. The duration for this test may be longer than the charger sizing criteria since the battery recharge is affected by float voltage, temperature, and the exponential decay in charging current. The battery is fully recharged when the measured charging current is _<f21 amp-s.

The Surveillance Frequency is acceptable, given the unit conditions required to perform the test and the other administrative controls existing to ensure adequate charger performance during these

[--+824 month} intervals. In addition, this Frequency is intended to be JX3.8-126 consistent with expected fuel cycle lengths.

SR 3.8.4.3 A battery service test is a special test of the 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 Reference 42. ICL3.8-172 The Surveillance Frequency of [1 8 months] is consistent with the recommendations of Regulatory Guide 1.32 (Ref. 9) and Regulator Cuide 1.129 (Ref 10), which state that tThe battery service test should be performed during refueling operations, or at some other outage, with intervals between tests not to exceed [-1-8-24 months. JX3.8-126 This SR is modified by two Notes. Note 1 allows the performance of a modified performance discharge test in lieu of a service test.

The reason for Note 2 is that performing the Surveillance would perturb the electrical distribution system and challenge safety systems. This restriction from normally performing the Surveillance in MODE 1 or 2 is further amplified to allow portions of the Surveillance to be performed for the purpose of reestablishing OPERABILITY (e.g. post work testing following corrective maintenance, corrective modification, deficient or R-12 WOG STS Rev. 2, 04//30/01 B 3.8.4 - 9 Markup for PI ITS Part E

DC Sources - Operating B 3.8.4 BASES SURVEILLANCE REQUIREMENTS (continued) incomplete surveillance testing, and other unanticipated OPERABILITY concerns) provided an assessment determines plant safety is maintained or enhanced. This assessment shall, as a minimum, consider the potential outcomes and transients associated with a failed partial Surveillance, a successful partial Surveillance, and a perturbation of the offsite or onsite system when they are tied together or operated independently for the partial Surveillance; as well as the operator procedures available to cope with these outcomes. These shall be measured against the avoided risk of a plant shutdown and startup to determine that plant safety is maintained or enhanced when portions of the Surveillance are performed in MODE I or 2. Risk insights or deterministic methods may be used for the assessment.

REFERENCES 1. 1 GFR.50, ,-Appendix A-,,GD*, AEC "General Design Criteria for IR-1 R 2I J Nuclear Power Plant Construction Permits." Criterion 39, issued for comment July 10, 1976, as referenced in USAR, Section 1.2.

2. Regulatory Guide 1.6, March 10, 1971.

IEE

3. 38 [--978]L-

. CL3.8-172

42. FUSAR, Ghapte Section 18.

5-3. FUSAR, GhaptefSection f6].

6. FSAf, Ghapter [I 51j.

7. RgtfulatCrI y Guide 1 .93, December 1974.

8. IiEEE-450- r995]

9. Regulatory Guide 1.32, February. 1977.

1U. iRegulaeOry GuIe U D.u, WOG STS Rev. 2, 04/30/01 B 3.8.4 - 10 Markup for PI ITS Part E

DC Sources - Shutdown B 3.8.5 APPLICABLE SAFETY ANALYSES (continued) shutdown tasks and associated electrical support to maintain risk at an acceptable low level. This may require the availability of additional equipment beyond that required by the shutdown Technical Specifications.

The DC sources satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii).

LCO The DC electrical power subsystems, [each required] [the required]

[stibsystern consisting of twea batteryies, one battery charger perbattery, and the corresponding control equipment and interconnecting cabling within [eite]-the train, fa-re]-isi required to be OPERABLE to support ti'eeldiied]fonel trainf-j of the distribution systems required OPERABLE by LCO 3.8.10, "Distribution Systems - Shutdown."j This ensures the availability of sufficient DC electrical power sources to operate the unit in a safe manner and to mitigate the consequences of postulated events during shutdown (e.g., fuel handling accidents [involving. ha. dling ree.ry

radiatted fuel])

A Note has been added to the LCO allowing the service building DC electrical power subsystem components to be used in lieu of the i R-12 required safeguards DC electrical power subsystem components when L-----

the required safeguards DC electrical power subsystem is inoperable due to testing, maintenance, or replacement. The service building DC power electrical components include the battery, associated battery charger, and the interconnecting cabling. When any of the service JPA3.8-212 building DC power electrical components are used in lieu of the safeguards DC electrical power subsystem components, they are required to be maintained in accordance with Specification 5.5.15 for monitoring various battery parameters that is based on the recommendations of r IEEE Standard 450-1995, "IEEE Recommended Practice For I R-12 Maintenance, Testing, and Replacement of Vented Lead-Acid Batteries ------ J For Stationary Applications" (Ref. 3).

APPLICABILITY The DC electrical power sources required to be OPERABLE in MODES 5 and 6, and during movement of [freeently irradiated fuel assemblies, provide assurance that:

a. Required features to provide adequate coolant inventory makeup are available for the irradiated fuel assemblies in the core; WOG STS Rev. 2, 04/30/01 B 3.8.5 - 3 Markup for PI ITS Part E

DC Sources - Shutdown B 3.8.5 APPLICABILITY (continued)

b. Required features needed to mitigate a fuel handling accident

[involving handling recently irradiated fuel (iLe., fuel that has occuipied part of a critical reactor core within the previouis []days)] are available;

c. Required features necessary to mitigate the effects of events that can lead to core damage during shutdown are available; and

d. Instrumentation and control capability is available for monitoring and maintaining the unit in a cold shutdown condition or refueling condition.

The DC electrical power requirements for MODES 1, 2, 3, and 4 are covered in LCO 3.8.4.

ACTIONS LCO 3.0.3 is not applicable while in MODE 5 or 6. However, since irradiated fuel assembly movement can occur in MODE 1, 2, 3, or 4, the ACTIONS have been modified by a Note stating that LCO 3.0.3 is not applicable. 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 operations. Entering LCO 3.0.3, while in MODE 1, 2, 3, or 4 would require the unit to be shutdown unnecessarily.

A.1. A.2ran[ r-__-- -1 IR-12

- REVIEWER'S NOTE AC)TION A is included ontly when plant specific implementation of LCO) 3.0.5 includes the potential to require both trains of the DCG Systemn to be PERADL.LE. If plant speclfic implementation resu*ilts inLO 9.a-.5 reqirig oly one trains of the DG Systemn to be O)PERABLE,-the-n ACTFION1 A is omnitted and AC)TION 0 is renumnbered as AC)TION A.

"- I--"----.----"-------.--

--- --- --- .----- *"T F'[I ' . . . k . .J -- -- ^ -----r-I

A -"%l-l Condition A represents one train with one [ortwojrequired battery PA3.8-198 chargers inoperable (e.g., the voltage limit of SR 3.8.4.1 is not maintained). The ACTIONS provide a tiered response that focu-ses on returning the battery to the fully chaIged state and restoring a fliiy qualified charger to OPEflADLE statuis in a reasonable timne period.

Required Action A.! requires that the battery terminal voltage be restored to greater than or equal to the mninimnum established float voltage within 2 houirs. This tin provides for returning the inoperable charger to OPERABLE statuis or providing an alternate meal ms of restoring battery---------------------

IR-12 I WOG STS Rev. 2, 04/30/01 B 3.8.5 - 4 Markup for PI ITS Part E -

DC Sources - Shutdown B 3.8.5 ACTIONS (continued) terminal voltage to greater than or equal to the minimum established float voltage. Restoring the battery terminal voltage to greater than or equal to the mi'nimurm established float voltage provides good assurance that, within [12] hours, the battery will be restored to its fully charged condition (Required Action A.2) frorm any discharge that might have occurred duet the charger inoperability. IPA3.8-198

- REVIEWER'S NOTE A plant that cannot meet the 12-hour Completion Timne duie to an ihrn battery charging characteristio can propose an alternate time equjal t-o 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months plus the time experienced to accomplish the exponential charginfg current portion of the battery charge profile following the Sefvce test A discharged battcry having termninal voltage of at least the minim 111um established float voltage indlicates that the battery is on the exponential charging current portion (the second padt) of its recharge cycle. The timfe to return a battery to its fully charged state uinder this condition is simply a function of the amount of the previous discharge and the recharge characteristic of the battery. Thus there is good assurance offully recharging the battery within [12] hourfs.

if established battery terminal float voltage cannot be restored to greate-r than or equial to the mninimumr established float voltage within 2 hourfs, and the charger is not operating in the current limiting moedes, a fauilty charger is indicated. A fauilty charger that is incapable of maintaining established battery terminal float voltage does not provide assurance that it can reve to and operate properly in the current limit modes that is necessary during the recovery period following a battery discharge event that the DG system is designed for.

If the charger is operating in the current limit mode after 2 houirs that is a indication that the battery is partially discharged and its capacity mnargins will be reduced. The time to return the battery to its fully charged condition in this case Is a function of the battery charger capacity, the amnount of loads on the associated BC system, the amount of the previousý discharge, and the recharge characteristic of the battery. The charge time can be extensive, and there is not adequate assurance that it can be rechargod within [12] houjrs (Required Action A.2).

Required Action A.2 requires that the battery float current be verified as less than or equal to [2] amps. This indicates that, ifthe battery had R-1 2 WOG STS Rev. 2, 04/30/0 1 B 3.8.5 - 5 Markup for PI ITS Part E

DC Sources - Shutdown B 3.8.5 ACTIONS (continued) diseharged as the resutlt of the inoperable baler; charger, it has oew been fully recharged. if at the expiration of the initial [12] hor period the bailer; float current is not less than or equal to [2] amps this indicates there may be additional batter; problems and the bailer; muistb declared inoperable. ICL3.8-1 71 Required Action A.81 limits the restoration time for the inoperable battery charger to 7-days8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . This action is applicable if an alternate means of restoring battery' termfinal voltage to greater than or equal to the minium etablished float voltage has been used (e.g. balance of plant non Class 1[ bailer; charger).. The 8 hour7-&aty Completion--------

Time reflects a reasonable time to effect restoration of the qualified IR-12 battery charger to OPERABLE status. .--

B-+- B.2.1.. 13.E2. B.-23. and B.2.4 Ifi two trains ar-eurd by LCO) 3.3.10G, the remaining train with DCG power available mfay be capable of supporting sufficient systemsl to allowV eentinuation of CORE ALTERATIONS and fuel moevemnent] [ Iinvlvn handling reenmtly irradiated fuel]. By allowing the option to declare required features inoperable with the associated 96 power source(s) inoperable, appropriate restrictions will be imnplemnented in accordance with the affected required featuires LCO) ACTIO)NS. InFmanyintce this option may involve uindesired admninistrative efforts. Condition B FPA3.8-1 99 represents one train with one required DC electrical power subsystem inoperable for reasons other than Condition A or if the Required Actions and associated Completion Time of Condition A are not met. In this Condition there may not be adequate DC power available to support the subsystems required by LCO 3.8.10. Therefore, the allowance for suffielently conservative actions are required is-meade (i.e., to suspend CORE ALTERATIONS, movement of treeentlyi-irradiated fuel assemblies, and operations involving positive reactivity additions) that assure ecould result in failure-to mteet-the minimum SDMV or boron concentration limit is meti-reuie to assure continued safe operation. Introduction of coolant inventory must be from sources that have a boron concentration greater than that what would be-required in the RCS for minimum SIDM or r----------------

refueling boron concentration. This may result in an overall reduction in R12 RCS boron concentration, but provides acceptable margin toL--

maintaining subcritical operation. Introduction of temperature changes including temperature increases when operating with a positive MTC must also be evaluated to ensure they do not result in a loss of required SDM.

WOG STS Rev. 2, 04/30/01 B 3.8.5 - 6 Markup for PI ITS Part E

DC Sources - Shutdown B 3.8.5 ACTIONS (continued)

Suspension of these activities shall not preclude completion of actions to establish a safe conservative condition. These actions minimize probability of the occurrence of postulated events. It is further required to immediately initiate action to restore the required DC electrical power subsystemfs- and to continue this action until restoration is accomplished in order to provide the necessary DC electrical power to the unit safety systems.

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

SURVEILLANCE SR 3.8.5.1 REQUIREMENTS SR 3.8.5.1 requires performance of all Surveillances required by SR 3.8.4.1 through SR 3.8.4.3. Therefore, see the corresponding I Bases for LCO 3.8.4 for a discussion of each SR. I1 R...

R-12 1 This SR is modified by a Note. The reason for the Note is to preclude ---------

requiring the OPERABLE DC sources from being discharged below their capability to provide the required power supply or otherwise rendered inoperable during the performance of SRs. It is the intent that these SRs must still be capable of being met, but actual performance is not required.

REFERENCES 1. F USAR, Chapter [6] Section 6.

2. F USAR-Ghapter [151 Section 14. JPA3.8-172

3. IEEE-450-1995.

WOG STS Rev. 2, 04/30/01 B 3.8.5 - 7 Markup for PI ITS Part E

Battery Parameters B 3.8.6 BASES APPLICABLE SAFETY ANALYSES (continued)

a. An assumed loss of all offsite AC power; or all o,,site AC power and

b. A worst-case single failure.

Battery parameters satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii).

LCO Battery parameters must remain within acceptable limits to ensure availability of the required DC power to shut down the reactor and maintain it in a safe condition after an anticipated operational occurrence or a postulated DBA. Battery parameter limits are conservatively established, allowing continued DC electrical system function even with limits not met. Additional preventative maintenance, testing, and monitoring performed in accordance with the plant proceduresfileensee

, ,ntrlledprogram] is conducted as specified in Specification 5.5.1-75.

APPLICABILITY The battery parameters are required solely for the support of the associated DC electrical power subsystems. Therefore, battery parameter limits are only required when the DC power source is required to be OPERABLE. Refer to the Applicability discussion in Bases for LCO 3.8.4 and LCO 3.8.5.

ACTIONS A Note has been added to provide clarification that, for this LCO, separate Condition entry is allowed for each battery. This is acceptable, since Required Actions for each Condition provide appropriate JPA3.8-158 compensatory actions.

A.1. A.2, and A.3 IPA3.8-171 With one or more cells in one er-mnoe-batteryies in one tri < f2.071 V, the battery cell is degraded. Within 28 hours3.240741e-4 days 0.00778 hours 4.62963e-5 weeks 1.0654e-5 months verification of the required battery charger OPERABILITY is made by monitoring the battery terminal voltage (SR 3.8.4.1) and of the overall battery state of charge by monitoring the battery float charge current (SR 3.8.6.1). This assures that there is still sufficient battery capacity to perform the intended function.

Therefore, the affected battery is not required to be considered inoperable solely as a result of one or more cells in one er-more-batteryies < f2.071 V, and continued operation is permitted for a limited period up to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

Since the Required Actions only specify "perform," a failure of SRI R-12 3.8.4.1 or SR 3.8.6.1 acceptance criteria does not result in this Required Action not met. However, if one of the SRs is failed the appropriate Condition(s), depending on the cause of the failures, is entered. If WOG STS Rev. 2, 04/30/01 B 3.8.6 - 2 Markup for PI ITS Part E

Battery Parameters B 3.8.6 BASES ACTIONS (continued)

SR 3.8.6.1 is failed then there ismay not be assurance that there is still sufficient battery capacity to perform the intended function and the j battery must be declared inoperable immediately. r I R-12 B.1 and B.2 One r--rnere-batteryies in one tr with float > t2i amps indicates that a partial discharge of the battery capacity has occurred. This may be due to a temporary loss of a battery charger or possibly due to one or PA3.8-171 more battery cells in a low voltage condition reflecting some loss of capacity. Within -8 hours verification of the required battery charger I OPERABILITY is made by monitoring the battery terminal voltage. If the terminal voltage is found to be less than the minimum established I R-12 float voltage there are two possibilities, the battery charger is L----

inoperable or is operating in the current limit mode. Condition A ICL3.8-166 addresses charger inoperability. If the charger is operating in the current limit mode after 28 hours3.240741e-4 days 0.00778 hours 4.62963e-5 weeks 1.0654e-5 months that is an indication that the battery I has been substantially discharged and likely cannot performi-----

required design functions. The time to return the battery to its fully I R-12 is a function of the battery charger charged condition in this case capacity, the amount of loads on the associated DC system, the amount of the previous discharge, and the recharge characteristic of the battery. The charge time can be extensive, and there is not adequate assurance that it can be recharged within-fl 2-hours (Required Action B.2). The battery must therefore be declared inoperable.

If the float voltage is found to be satisfactory but there are one or more battery cells with float voltage less than-f2.071-V, the associated 'QB" statement in Condition F is applicable and the battery must be declared inoperable immediately. If float voltage is satisfactory and there are no cells less than-f2.07]-V there is good assurance that, within-fl 2t-hours, the battery will be restored to its fully charged condition (Required Action B.2) from any discharge that might have occurred due to a temporary loss of the battery charger.

IR-12 I

I I

- REVIEWER'S NOTE i-..- - ...

A plant that cannot meet the 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Comnpletion Tmre duie to an inherent battery charging characteristic can propose an alternate time equal to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months plus the time experienced to acecomnplish the exponential WOG STS Rev. 2, 04/30/01 B 3.8.6 - 3 Markup for PI ITS Part E

Battery Parameters B 3.8.6 BASES ACTIONS (continued) eharging curroent portion of the battery charge prefile following the service A discharged battery with float voltage (the charger setpoint) across its terminals indicates that the battery is on the exponential charging current portion (the second part) of its recharge cycle. The time to return a JPA3.8-187 battery to its fully charged state under this condition is simply a function of the amount of the previous discharge and the recharge characteristic of the battery. Thus there is good assurance of fully recharging the battery within f+2124 hours, avoiding a premature shutdown with its own attendant risk and the battery is not inoperable.

If the condition is due to one or more cells in a low voltage condition but still greater than t2.07i-V and float voltage is found to be satisfactory, this is not indication of a substantially discharged battery and-Fl-2-24 hours is a reasonable time prior to declaring the battery inoperable.

1R-12 Since Required Action B.1 only specifies "perform," a failure of L SR 3.8.4.1 acceptance criteria does not result in the Required Action not met. However, if SR 3.8.4.1 is failed, the appropriate Condition(s),

depending on the cause of the failure, is entered.

0.1. C.2, and C.3 With one eor-nere-batteryees in one-tr with one or more cells electrolyte level above the top of the plates, but below the minimum established design limits, the battery still retains sufficient capacity to perform the intended function. Therefore, the affected battery is not required to be considered inoperable solely as a result of electrolyte level not met.

Within 31 days the minimum established design limits for electrolyte level must be re-established.

With electrolyte level below the top of the plates there is a potential for dryout and plate degradation. Required Actions C.1 and C.2 address this potential (as well as provisions in Specification 5.5.1-5, Battery ICL3.8-172 Monitoring and Maintenance Program). They are modified by a note that indicates they are only applicable if electrolyte level is below the top of the plates. Within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months level is required to be restored to above the top of the plates. The Required Action C.2 requirement to verify that there is no leakage by visual inspection and the Specification 5.5.175.b item to initiate action to equalize and test in accordance with manufacturer's recommendation are taken from Annex D of IEEE Standard 450-1995.

They are performed following the restoration of the electrolyte level to WOG STS Rev. 2, 04/30/01 B63.8.6- 4 Markup for PI ITS Part E

Battery Parameters B 3.8.6 BASES ACTIONS (continued) above the top of the plates. Based on the results of the manufacturer's recommended testing the batterfyjjiesj may have to be declared inoperable and the affected cellfsj replaced.

D.1 With one or more batteryies in one tr with pilot cell temperature less than the minimum established design limits, 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is allowed to restore the temperature to within limits. A low electrolyte temperature limits the current and power available. Since the battery is sized with margin, while battery capacity is degraded, sufficient capacity exists to perform the intended function and the affected battery is not required to be considered inoperable solely as a result of the pilot cell temperature not met.

E..1 PA3.8-161 I With one or more batteries in the redundant trains with battery parameters not within limits there is not sufficient assurance that battery capacity has not been affected to the degree that the batteries IR-12 can still perform their required function, given that redundant batteries are involved. With redundant batteries involved this potential could result in a total loss of function on multiple systems that rely upon the batteries. The longer Completion Times specified for battery parameters on non-redundant batteries not within limits are therefore not appropriate, and the parameters must be restored to within limits on at least one train within 8- hours.

F.1 IR-12 With one or more batteries with any battery parameter outside the allowances of the Required Actions for Condition A, B, C, D, or E, sufficient capacity to supply the ,maximum epeteddesign load requirement is not assured and the corresponding battery must be declared inoperable. Additionally, discovering one or more batteries in one train with one or more battery cells float voltage less than-[2.07j-V and float current greater than t2j-amps indicates that the battery capacity may not be sufficient to perform the intended functions. The battery must therefore be declared inoperable immediately.

IR-12 L - I WOG STS Rev. 2, 04/30/01 B 3.8.6 - 5 Markup for PI ITS Part E

Battery Parameters B 3.8.6 BASES SURVEILLANCE REQUIREMENTS (continued) electrolyte temperature is maintained above this ICL3.8-172 temperature to assure the battery can provided-the required r ---- -n current and voltage to meet the design requirements. Temperatures I R-12 lower than assumed in battery sizing calculations act to inhibit or reduce L-----J battery capacity. The Frequency is consistent with I EEE-450 (Ref. 1).

S R 3.8.6. 6 A battery performance discharge test is a test of constant current capacity of a battery, normally done in the as found condition, after having been in service, to detect any change in the capacity determined by the acceptance test. The test is intended to determine overall battery degradation due to age and usage.

Either the battery performance discharge test or the modified performance discharge test is acceptable for satisfying SR 3.8.6.6 , however, only the modified performanee discharge test may be used to satisfy the batter service test requiremnents of SRl 3.8.4.3-.

A modified discharge test is a test of the battery capacity and its ability to provide a high rate, short duration load (usually the highest rate of the duty cycle). This will often confirm the battery's ability to meet the critical period of the load duty cycle, in addition to determining its percentage of rated capacity. Initial conditions for the modified performance discharge test should be identical to those specified for a service test.

It may consist of just two rates; for instance the one minute rate for the battery or the largest current load of the duty cycle, followed by the test rate employed for the performance test, both of which envelope the duty cycle of the service test. Since the ampere-hours removed by a one minute discharge represents a very small portion of the battery capacity, the test rate can be changed to that for the performance test without compromising the results of the performance discharge test. The battery terminal voltage for the modified performance discharge test must remain above the minimum battery terminal voltage specified in the battery service test for the duration of time equal to that of the service test.

The acceptance criteria for this Surveillance are consistent with IEEE-450 (Ref. 1-3) and IEEE-485 (Ref. 4). These references recommend that the battery be replaced if its capacity is below 80% of the [CL3.8-172 manufacturer's 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. Furthermore, the battery is sized to meet or exceed WOG STS Rev. 2, 04/30/01 B 3.8.6 - 7 Markup for PI ITS Part E

Inverters -Operating B 3.8.7 1PA3.8-100 BASES (continued) maintaining required Reactor Protection Instrument AC Panelsvita--b-sea OPERABLE during accident conditions in the event of:

a. An assumed loss of all offsite AC electrical power IL3.8-163i or all ansite AC elcctrieal power; and

b. A worst case single failure.

Inverters are a part of the distributi'n systo" an", IPA3.8-217 sueh--,satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii)-tilhN ral iey Statemnent.

LCO The inverters ensure the availability of AC electrical power for the systems instrumentation required to shut down the reactor and maintain it in a safe condition after an anticipated operational occurrence (AO0) or a postulated DBA.

Maintaining the required inverters OPERABLE ensures that the redundancy incorporated into the design of the RPS and ESFAS instrumentation and controls is maintained. The four inverters [(twL per train)] ensure an uninterruptible supply of AC electrical power to the Reactor Protection !R-12 i-ta buse

-- even if the 4--16 kV L ........

Instrument AC Panels sa-fetySafeguards buses are de-energized.

(conti nued)

WOG STS Rev 1, 04/07/95 B 3.8.7-2 Markup for PI ITS Part E

Inverters -Operating B 3.8.7 JPA3.8-100I BASES (continued)

OPERABLEpearb-l-e inverters require the associated v-it--Reactor Protection Instrument AC Panel-4tts to be ICL3.8-I C

80O powered by the inverter with output voltage and frequency within tolerances, and power supply~ntu-t to the inverter from a J125 VDC+ station battery.

Altermat4ve3ýyNormally, the power supply may-beis fromiCL3.8-183 an internal AC source via rectifier as long aswith the station battery-+/-s- available as the uninterruptible power supply.

r . ... - - -i Thi s LCO 4s medi fi ed by a Note that a]llows [one/twa+ R-12 inverters to be dT3eVT=etdfoa cmnbterfr L--------1

-t24 hours, if the vital buts(es) is powered fromf a

[Klass 1E constant voltage transformner or inverter IPA3.E ~-1851 using internal AC source] during the period and al4 other inverters are operable. This allows an equalizing charge to be placed on one battery. if the invert rs were not disconnected, the resulting voltage condition milght b a n k. .. %, I ll ',Jl l li I ,. , I I*i III i U I I Y ,. l l S / . V l Thse-

^rovisions-minimize damnage the inverterEs] the loss of equipmfent that would occur in the event of a less of offsite power. The 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months timfe period for the allowance mfinifmizes the timne during which a loss of offsite power.

eould result in the less of equipmnent energized fromf the affected AC vita] bus while taking inte consideration the time required to performf an equalizing eharge on the battery The intent of this Note is to limnit the number of inverters that mfay be disconnected. Only those inýqive-rter's PA3.8-185

-associated with the single battery utndergoing an equalizing charge may be disonnkec*teLd. ll Iothr Inverters mfust be aligned to their associated batteries, regardless of the numfber of inverters or unit design.

(conti nued)

WOG STS Rev 1, 04/07/95 B 3.8.7-3 Markup for PI ITS Part E

Inverters- Operating B 3.8.7 lPA3.8-100 BASES (continued)

APPLICABILITY The inverters are required to be OPERABLE in MODES 1, 2, 3, and 4 to ensure that:

a. Acceptable fuel design limits and reactor coolant pressure boundary limits are not exceeded as a result of AOOs or abnormal transie*; and J -205

b. Adequate core cooling is provided, and containment OPERABILITY and other vital functions are maintained in the event of a postulated DBA.

Inverter requirements for MODES 5 and 6 are covered in the Bases for LCO 3.8.8, "Inverters-Shutdown."

ACTIONS A.1 and A.2 With one Reactor Protection Instrument AC inverter inoperable, Required Action A.1 and A.2 require ICL3.8_183 verification, within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , the Reactor Protection Instrument AC panel with an inoperable inverter is powered from Panel 117 (Unit 2 - 217) or verify that the Reactor Protection Instrument AC panel with an inoperable inverter is powered from its inverter bypass source.

Plant design provides acceptable alternate methods of powering a Reactor Protection Instrument AC panel with an inoperable inverter. Panel 117 (Unit 2 - Panel 217), by plant design, can provide reliable power to a Reactor Protection Instrument AC panel. Alternatively, a Reactor Protection Instrument AC panel may be powered by an inverter internal bypass. In the event an inverter becomes inoperable, the the inverter static transfer bypass switch will automatically bypass, thus providing power to the associated Reactor Protection Instrument AC panel and maintain OPERABILITY. Required Actions A.1 and A.2 require verification that only one Reactor Protection Instrument AC 1R-12 (continued) L-------

WOG STS Rev 1, 04/07/95 B 3.8.7-4 Markup for PI ITS Part E

Inverters -Operating B 3.8.7 IPA3.8-100 BASES (continued) panel is powered from Panel 117 (Unit 2 - Panel 217) or an inverter bypass source. This verification must be completed within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

r R-12 ,l B.1, B.2, and B.3 L

With a req-urectwo Reactor Protection Instrument AC inverters inoperable, the+-t-s associated Reactor Protection Instrument AC panelsvital bus bec.mes inoperable until it is Eafu,-,'ar-"l,-are considered to be inoperable unless they are energized from Panel 117 (Unit 2 - Panel 217) or they are automatically re-energized freamby their inverteri+s- static transfer switch. [Class 1E .nstant voltage sour transfrm.er or -inverter using internal A sourCe].

For this reason a Note has been included in CL3.8-183 1.

Condition AB requiring the entry into the Conditions and Required Actions of LCO 3.8.9, "Distribution Systems-Operating." This ensures that the v+t-a-Reactor Protection Instrument AC panel b*ts-is re-energized within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months . Plant design provides acceptable alternate methods of powering Reactor Protection Instrument AC panels with an inoperable inverter. Panel 117 (Unit 2 - Panel 217), by plant design, can provide reliable power to a Reactor Protection Instrument AC panel. Alternatively, a Reactor Protection Instrument AC panel may be powered by an inverter internal bypass. In the event an inverter becomes inoperable, the inverter static transfer bypass switch will automatically bypass, thus providing power to the associated Reactor Protection Instrument AC panel and maintain OPERABILITY. Therefore, based on plant design, Required Actions B.1 and B.2 require verification that no more than one Reactor Protection Instrument AC inverter will be powered from Panel 117 (Unit 2 - Panel 217) and one or both Reactor Protection Instrument AC panel(s) are powered from an inverter bypass source. This verification must be completed within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

R-12 (continued) ----------

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Inverters - Operating B 3.8.7 1PA3.8-100 BASES I I Required Action AB.J3 allows £248 hours0.00287 days 0.0689 hours 4.100529e-4 weeks 9.4364e-5 months inverter and return it to to fix ICL3.8-183 L-- JI the inoperable service. The -248 hour limit is based upon engineering judgment, taking into consideration the time required to repair an inverter and the additional risk to which the unit is exposed because of the inverter inoperability. This has to be balanced against the risk of an immediate shutdown, along with the potential challenges to safety systems such a shutdown might entail. When the Reactor Protection Instrument AC Panelvital bus is powered from its alternate onstant voltage source, it is relying upon interruptible AC electrical power sources (offsite and onsite). The uninterruptible inverter source to the Reactor Protection is the preferred source for Instrument AC Panel vi*-*, *ab-u.s.s powering instrumentation trip setpoint devices.

r__I R-12 I ACTIONS BC.1 and-PBC.2 I L----------J I

(continued)

If the inoperable devices or components cannot be restored to OPERABLE status within the required Completion Time, the unit must be brought to a MODE in which the LCO does not apply. To achieve this status, the unit 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 unit conditions from full power conditions in an orderly manner and without challenging plant systems.

SURVEILLANCE REQUIREMENTS This Surveillance verifies that the inverters are IPA3.8-102 functioning properly with all required circuit breakers closed and Reactor Protection Instrument AC panelsvita+/--bttseý energized from the inverter. The verification of proper voltage and freq-ueneyoutput ensures WOG STS Rev 1, 04/07/95 B 3.8.7-6 Markup for PI ITS Part E

Inverters -Operating B 3.8.7 JPA3.8-100 BASES that the required power is readily available for the instrumentation of the RPS and ESFAS connected to the Reactor Protection Instrument AC panelsvi-t-l-e-*s. The 7 day Frequency takes into account the redundant capability of the inverters and other indications available in the control room that alert the operator to inverter mal functions.

REFERENCES 1. FUSAR, ha*pterSection t8].

2. FUSAR, ChapterSection f-6114.

3. FSAR, Chapter [15].

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Inverters- Shutdown B 3.8.8 IPA3.8-100 BASES (continued)

c. Systems necessary to mitigate the effects of events that can lead to core damage during shutdown are available; and

d. Instrumentation and control capability is available for monitoring and maintaining the unit in a cold shutdown condition or refueling condition.

Inverter requirements for MODES 1, 2, 3, and 4 are covered in LCO 3.8.7.

ACTIONS LCO 3.0.3 is not applicable while in MODES 5 and 6.

However, since irradiated fuel assembly movement ITA3.8-140 can occur in MODE 1, 2, 3, or 4, the ACTIONS have been modified by a Note stating that LCO 3.0.3 is not applicable. 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 operations.

Entering LCO 3.0.3 while in MODE 1, 2, 3, or 4 would require the unit to be shutdown unnecessarily.

R-12 AT-,-A.2-I. A.£--.2, A.£Z-.3. and A.2--4 If the required inverter is inoperable, the remaining OPERABLE Reactor Protection Instrument AC Panel power ICL3.8-177 suppliestwo trains arc as required by LCO 3.8.10, "Distribution Systems-Shutdown," the remaining OPPRABL[E inverters may be capable of supporting sufficient required features to allow continuation of CORE ALTERATIONS, fuel movement, frrdor operations with a potential for positive reactivity additions. By the allowance of the option to declare A.1. A.2.1. A.2.2, A.2.3. and A.2.4 (continued)

(conti nued)

WOG STS Rev 1, 04/07/95 B 3.8.8-5 Markup for PI ITS Part E

Inverters - Shutdown B 3.8.8 IPA3.8-100 BASES required features inoperable with the associated inverter(s) inoperable, appropriate restrictions will be imnplemfented in accordance with the affected required features L..s' Required Actions. Inmany insta nes, t 4- PA3.8-215 option may involve undesired ad'inistrative ef**rts. -Therefore, the allowance for sufficiently r conservative actions is made (i.e., to suspend CORE ALTERATIONS, movement of irradiated fuel assemblies, and operations involving positive reactivity additions). The Required Action to suspend positive reactivity additons TA3.8-117 does not preclude actions to mfaintain or increasereco vessel inventory, provided the required SUM is maintained-that could result in loss of required SDM (MODE

5) or boron concentration (MODE 6)). Suspending positive reactivity additions that could result in failure to meet the minimum SDM or boron concentration limit is required to assure continued safe operation. Introduction of coolant inventory must be from sources that have a boron concentration greater than that required in the RCS for minimum SDM or refueling boron concentration. This may result in an overall reduction in RCS boron concentration, but provides acceptable margin to maintaining subcritical operation. Introduction of temperature changes including temperature increases when operating with a positive MTC must also be evaluated to ensure they do not result in a loss of required SDM.

Suspension of these activities shall not preclude completion of actions to establish a safe conservative condition.

These actions minimize the probability of 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 necessary inverter power to the unit safety systems.

(continued)

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Inverters- Shutdown B 3.8.8 IPA3.8-100 BASES (continued)

The Completion Time of immediately is consistent with the required times for actions requiring prompt attention. The restoration of the required inverter- should be completed as quickly as possible in order to minimize the time the unit safety systems may be without power or powered from CL3.8-177 a constant voltage source transformer.

SURVEILLANCE SR 3.8.8.1 REQUIREMENTS This Surveillance verifies that the required inverter& eireis functioning properly with all required circuit breakers closed and Reactor Protection Instrument AC Panel-vital b-us-es--energized from the inverter. The verification of proper voltage and ensures that the

,reqtenty-output PA3.8-102 required power is readily available for the instrumentation connected to the Reactor Protection Instrument AC vital busesPanel. The 7 day Frequency takes into account the redundant capab+li-tlyreliability of the ir,+/-ert-ersinstrument panel power sources and other indications available in the control room that alert the operator to in-rve-rt-er malfunctions.

REFERENCES 1. FSAR, [6].

2. FSAR, Chapter [15]*None.

(conti nued)

WOG STS Rev 1, 04/07/95 B 3.8.8-7 Markup for PI ITS Part E

Distribution Systems -Operating B 3.8.9 PA3.8-100 B 3.8 ELECTRICAL POWER SYSTEMS B 3.8.9 Distribution Systems-Operating BASES BACKGROUND The onsite safeguardsClass !E AC7 and DC, and AC vita] bus electrical power distribution systems are divided by train into +two+ redundant and independent AC, DG, and AC vital bs-u&-electrical power distribution subsystems. The onsite Reactor Protection Instrument AC Distribution System is divided by channels into four separate subsystems (Ref. [CL3.8-167 1).

EachT-ie AC electrical power subsystem for each train consists of a pri-mary safeguards"ngine.red Safety Feature I and .120] V kV bus and twoseeandary {480 SE4--1-6 ,

buses. 7- These in turn supply power to distribution 1 R-12 I L

L . . . .I I panels- and motor control centers (MCCs) and lead een-ters. Each safeguards f4.-1-6 kV E-Sf-bus+/- has twofat-l-ea*-s Lone separate and independent offsite sources of power+/- as well as a dedicated onsite diesel generator (DG) source.

Each safeguards f4--.-6 kV E-S-bus- is normally connected to an preferred-offsite source. After a loss of thise pre-ferred offsite power source to a 4.16 kV ESF bus, a transfer to the alternate offsite source is accomplished by a load sequencer, initiated byutilizing a timo,, delay**d bus undervoltage relays. If all offsite sources are unavailable, the onsite emergency DG supplies power to the safeguards 4--1-6 kV -SF-bus. Control power for the 4-.1-6 kVlcL3.8_167 and 480 V bus breakers is supplied from the safeguards DC '-l distributionClass 1E batteries system. Additional description of the safeguards ACt*h-i-h system may be found in the Bases for LCO 3.3.4, "4 kV Safeguards Bus Voltage Instrumentation," and the Bases for LCO 3.8.1, "AC Sources-Operating," and the Bases for LGO 3.8.4, "DC

-nue ottreei (continued)

WOG STS Rev 1, 04/07/95 B 3.8.9-1 Markup for PI ITS Part E

Distribution Systems- Operating B 3.8.9 JPA3.8-100 I BASES (continued)

The seeondary AC electrical power distribution system for each train includes the safety related busesload eenters, motor control centers and MCCsT and distribution pa..,s shown in Table B 3.8.9-1.

1R-12 ,

1 The 120 V Reactor Protection Instrument AC vita -

b-us-esPanels are arranged in fourtwe load groups-pe-r train and are normally powered from the-inverters. AnT-he alternate power supply for the instrument panels isv-t-a- CL3.8-167 1 buses are Class 1E constant voltage sou the inverter bypass transformers powered from the same MCCt-ra-i- as the associated inverter. Another alternate power supply is from the unit 208/120 VAC interruptable panel., and its Uttse of these supplies is governed by LCO 3.8.7, "Inverters-Operating." Each constant voltage source transformer is powered from a Class 1E AC bus.

There are two independent 125/2-50 VDC electrical power distribution subsystems (one for each train). The 125 VDC safeguards electrical power system consists of two independent and redundant safety related DC safeguards CL3.8-167 electrical power subsystems (Train A and Train B). The sources for each train are a 125 VDC battery, a battery charger, and all the associated control equipment and interconnecting cabling.

The list of the-4l required Reactor Protection Instrument AC and safeguards DC distribution panelsbtses is presented in Table B 3.8.9-1.

APPLICABLE The initial conditions of Design Basis Accident (DBA) and SAFETY ANALYSES transient analyses in the UFSAR, Chapter [6] (Ref. 1), and in the ,SAR, Chapter [*,5 (Ref. 2)- assume ESF systems are OPERABLE. The safeguards AC, DC, and Reactor Protection Instrument AC vital bus electrical power distribution systems are designed to provide sufficient capacity, (continued)

WOG STS Rev 1, 04/07/95 B 3.8.9-2 Markup for PI ITS Part E

Distribution Systems -Operating B 3.8.9 lPA3.8-100 BASES Maintaining the Train A and Train B safeguards AC7 and DC, and Reactor Protection Instrument AC vital bus electrical power distribution subsystems OPERABLE ensures that the redundancy incorporated into the design of ESF is not defeated. Therefore, a single failure within any system or within the electrical power distribution subsystems will not prevent safe shutdown of the reactor. This does not preclude redundant safeguards 4 kV buses from being powered from the same offsite path.

subsystems ICL3.8-167 LCO OPERABLE AC electrical power distribution (continued) require the associated buses, load centers, and MCCsmnot*-or ontrol centers, and distribution pans to be energized to their proper voltages. OPERABLE DC electrical power ,

distribution subsystems require the associated !R-12 L -----

panelsu-ses to be energized to their proper voltage from either the associated battery or charger. OPERABLE Reactor Protection Instrument AC vital bus electrical power distribution subsystems require the associated panelsbd-s-es to be energized to their proper voltage from thc associated

[inverter via inverted DC voltage, inverter using internal AC *soudre,

enstant or Class voltage transformer].

in addition, tie S... breakers between ......

redund~ant safety lCL3 -167 related AC, DC, and AC vital bus power distribution subsystemfs, if they exist, must be open. This prevents any electrical mfalfunction in any power distribution~

suibsystemf fromn propagating to the redundant subsystemf, that could cause thc failure of a redundant subsystem and a loss of esscntial safety function(s). if any tie breakers are closed, the affected redundant electrical power distributiont subsystemns arc considered inoperable. This applies to the onsite, safety' related redundant electrical power distribution subsystems. it does not, however, preclude rcdundant Class 1E 4.16 kV buses from bigpowvered fromi the samce offsite circuit.

(conti nued)

WOG STS Rev 1, 04/07/95 B 3.8.9-4 Markup for PI ITS Part E

Distribution Systems -Operating B 3.8.9 BASES (continued)

IPA3 "8-1°° APPLICABILITY The electrical power distribution subsystems are required to be OPERABLE in MODES 1, 2, 3, and 4 to ensure that:

a. Acceptable fuel design limits and reactor coolant pressure boundary limits are not exceeded as a result R-12 of AOOs or abnormal transients andL ICL3.8-205

b. Adequate core cooling is provided, and containment OPERABILITY and other vital functions are maintained in the event of a postulated DBA.

Electrical power distribution subsystem requirements for MODES 5 and 6 are covered in the Bases for LCO 3.8.10, "Distribution Systems -Shutdown."

ACTIONS A.1 and A.2 With one or more req-u-ired-safeguards AC electrical power I distribution subsystemsbus-es, load centers, motor : R-12 ontrol centers, or distribution panels, exeept JP3.-213I L-..........

AC vital buses, in one trn inoperable, the L I remaining AC electrical power distribution subsystems in-t-he other train isare capable of supporting the minimum safety functions necessary to shut down the reactor and maintain IR-12 I I I it in a safe shutdown condition, assuming no single failure. The overall reliability is reduced, however, because a single failure in the remaining power distribution subsystems could result in the minimum required ESF functions not being supported. Therefore, JPA3"8-213 there are two Required Actions that can be taken. Required Action A.1 would allow declaring the I

affected supported feature(s), being powered from the inoperable portion of the safeguards AC electrical power distribution system, inoperable. If Required Action A.1 is used, LCO 3.0.6 would also be entered to verify that no loss of function would exist. If LCO 3.0.6 identifies that a loss of function did exist, Condition E would be entered.

R-12 L----------

(continued)

WOG STS Rev 1, 04/07/95 B 3.8.9-5 Markup for PI ITS Part E

Distribution Systems -Operating B 3.8.9 BASES (continued) 1PA3*8-IO0I Required Action A.2 requiresd safeguards AC electrical power buses, Ilad centers, .t.r. ontrol centers, and distribution subsystems panels-mus+ to be restored to OPERABLE status within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . 'R ondi"tion-AThe worst scenario is one train without AC power (i.e., no offsite power to the train and the associated DG inoperable). In this Condition, the unit is more vulnerable to a complete loss of AC power. It is, therefore, imperative that the unit operator's attention be focused on minimizing the potential for loss of power to the remaining train by stabilizing the unit, and on restoring power to the affected train. The 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months time limit before requiring a unit shutdown in this Condition is acceptable because of: R-12 L-

a. The potential for decreased safety if the unit operator's attention is diverted from the evaluations and actions necessary to restore power to the affected train, to the actions associated with taking the unit to shutdown within this time limit; and

b. The potential for an event in conjunction with a single failure of a redundant component in the train with AC power.

The second Completion Time for Required Action A.2:

establishes a limit on the maximum time allowed for any combination of required distribution subsystems to be inoperable during any single contiguous occurrence of failing to meet the LCO. If Condition A is entered while, for instance, a DC bus is inoperable and subsequently restored OPERABLE, the LCO may already have been not met for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months . This could lead to a total of 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months , since initial failure of the LCO, to restore the AC distribution system. At this time, a DC circuit could again , IR-12, ,

L--

(conti nued)

WOG STS Rev 1, 04/07/95 B 3.8.9-6 Markup for PI ITS Part E

Distribution Systems- Operating B 3.8.9 IPA3.8-100 BASES

__ " i ! FRF become inoperable, and AC distribution restored OPERABLE.

This could continue indefinitely.

The Completion Time allows for an exception to the normal "time zero" for beginning the allowed outage time "clock."

This will result in establishing the "time zero" at the time the LCO was initially not met, instead of the time Condition A was entered. The 16 hour1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months Completion Time is an acceptable limitation on this potential to fail to meet the LCO indefinitely.

Required Action A.1 and A.2 are modified by a Note that requires the applicable Conditions and RequiredIPA3.8-213]

Actions of LCO 3.8.4, "DC Sources - Operating," to be entered for DC trains made inoperable by inoperable AC power distribution subsystems. This is an exception to LCO 3.0.6 and ensures the proper actions are taken for these components. Inoperability of a distribution system can result in loss of charging power to batteries and eventual loss of DC power. This Note ensures that the appropriate attention is given to restoring charging power to batteries, if necessary, after loss of distribution systems.

BC.1 and C.2 With one Reactor Protection Instrument AC Panel vital bus R-12 1 inoperable, the remaining OPERABLE Reactor Protection L Instrument AC Panelst,1bu-*oe are capable of supporting the minimum safety functions necessary to shut down the unit and maintain it in the safe shutdown condition. Overall reliability is reduced, however, since an additional single failure could result in the minimum Erequiredl-ESF IPA3.8-213 functions not being supported. Therefore, there aren two Required Actions that can be taken. Required Action C.1 would allow declaring the affected supported feature(s)

R-12 (continued) L---

WOG STS Rev 1, 04/07/95 B 3.8.9-7 Markup for PI ITS Part E

Distribution Systems -Operating B 3.8.9 IPA3"8-1001 BASES being powered from the inoperable portion of the Reactor Protection Instrument AC panel, inoperable. If iPA3.8_213 Required Action C.1 is used, LCO 3.0.6 would also A be entered to verify that no loss of function would exist.

If LCO 3.0.6 identifies that a loss of function did exist, Condition E would be entered. Required Action C.2 The requiresd the Reactor Protection Instrument AC panel+at--l bus ,must to be restored to OPERABLE status within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months ---- 1 by powering the panelbtti- from the associated {inverter vt-a R-12 inverted --, inverter using internal AC source, or Class 1E

.onstant voltagebypass transformer+/-, or interruptible panel.

Condition BC represents one Reactor Protection ICL3.8-167 IIR-12 Instrument AC vital buspanel without power, I . . . I potentially both the DG source and the associated-AC source are nonfunctioning. In this situation, the unit is significantly more vulnerable to a complete loss of all noninterruptible power. It is, therefore, imperative that the operator's attention focus on stabilizing the unit, minimizing the potential for loss of power to the remaining instrumentvital buses-panels and restoring power to the affected instrumentvital buspanel.

This 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months limit is more conservative than Completion Times allowed for the vast majority of components that are without adequate instrumentv-i-t-a-l AC power. Taking exception to LCO 3.0.2 for components without adequate instrumentv+ta- AC power, that would have the Required Action Completion Times shorter than 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months if declared inoperable, is acceptable because of:

ACTIONS fi.1 (continued)

(conti nued)

WOG STS Rev 1, 04/07/95 B 3.8.9-8 Markup for PI ITS Part E

Distribution Systems -Operating B 3.8.9 1PA3.8-1001 BASES

a. The potential for decreased safety by requiring a change in unit conditions (i.e., requiring a shutdown) and not allowing stable operations to continue;

b. The potential for decreased safety by requiring entry into numerous Aapplicable Conditions and Required Actions for components without adequate instrument'-t-& AC power and not providing sufficient time for the operators to perform the necessary evaluations and actions for restoring power to the affected train; and

c. The potential for an event in conjunction with a single failure of a redundant component.

The 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months Completion Time takes into account the importance to safety of restoring the Reactor Protection Instrument AC v ,talbusPanel to OPERABLE status, the redundant capability afforded by the other OPERABLE instrumenti-t-albuseS panels, and the low probability of a DBA occurring during this period.

The second Completion Time for Required Action CB.2+/-

establishes a limit on the maximum allowed for any combination of required distribution subsystems to be inoperable during any single contiguous occurrence of failing to meet the LCO. If Condition CB is entered while, for instance, an AC bus is inoperable and subsequently returned OPERABLE, the LCO may already have been not met for up to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . This could lead to a total of 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months , since initial failure of the LCO, to restore the vital bus distribution system. At this time, an AC train could again become inoperable, and vital bus distribution restored OPERABLE. This could continue indefinitely.

This Completion Time allows for an exception to the normal "time zero" for beginning the allowed outage time "clock."'

R-12 (continued)

WOG STS Rev 1, 04/07/95 B 3.8.9-9 Markup for PI ITS Part E

Distribution Systems -Operating B 3.8.9 lPA3.8-100 BASES This will result in establishing the "time zero" at the time the LCO was initially not met, instead of the time Condition CB was entered. The 16 hour1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months Completion Time is an acceptable limitation on this potential to fail to meet the LCO indefinitely. r --.. .. 1 1R-12 L

ACTIONS LB.1 and B.2 (continued) ... ,

power JP...

With one or more safeguards DC electrical distribution subsystem panel (s)bus(es) in one train R-12 inoperable, the remaining safeguards DC electrical power L------- I distribution subsystems-a-ire is capable of supporting the minimum safety functions necessary to shut down the reactor and maintain it in a safe shutdown condition, assuming no single failure. The overall reliability is reduced, however, because a single failure in the remaining safeguards DC electrical power distribution subsystem could result in the minimum required ESF functions not being supported. Therefore, there are two Required Actions that can be taken. Required Action B.1 would allow declaring the affected supported feature(s), being powered from the inoperable portion of the safeguards DC panel, inoperable.

If Required Action B.1 is used, LCO 3.0.6 would also be entered to verify that no loss of function would exist. If LCO 3.0.6 identifies that a loss of function did exist, Conditon E would be entered. Required Action B.2 t-hre

-required-s the DC panelsbuts-es--mfts-+ be restored to OPERABLE status within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months by powering the bus from the associated battery, o-i-charger, or portable charger. R-12

C Condition represents one trainThe worst case scenario is one train without adequate safeguards DC power; potentially with both-with the battery significantly degraded and the associated charger nonfunctioning. In this situation, the unit is significantly more vulnerable to a complete loss of all DC power. It is, therefore, imperative that the (conti nued)

WOG STS Rev 1, 04/07/95 B 3.8.9-10 Markup for PI ITS Part E

Distribution Systems -Operating B 3.8.9 IPA3.8-1001 BASES operator's attention focus on stabilizing the unit, minimizing the potential for loss of power to the remaining trains and restoring power to the affected train.

This 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months limit is more conservative than Completion Times allowed for the vast majority of components that would be without power. Taking exception to LCO 3.0.2 for components without adequate DC power, which would have Required Action Completion Times shorter than 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , is acceptable because of:

a. The potential for decreased safety by requiring a R-12 1 L------- J change in unit conditions (i.e., requiring a shutdown) while allowing stable operations to continue;

b. The potential for decreased safety by requiring entry into numerous applicable Conditions and Required Actions for components without DC power and not providing sufficient time for the operators to perform the necessary evaluations and actions for restoring power to the affected train; and

c. The potential for an event in conjunction with a single failure of a redundant component.

The 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months Completion Time for DC buses is consistent with Regulatory Guida 1.93 (Ref. 3).

ACTIONS G.-1---(continued) I 169 The second Completion Time for Required Action BK.2+/- establishes a limit on the maximum time allowed for any combination of required distribution subsystems to be inoperable during any single contiguous occurrence of failing to meet the LCO. If Condition BK is entered while, for instance, an AC bus is inoperable and subsequently returned OPERABLE, the LCO may already have been not met r IR-12 L- -

(conti nued)

WOG STS Rev 1, 04/07/95 B 3.8.9-11 Markup for PI ITS Part E

Distribution Systems -Operating B 3.8.9 IPA3.8-100 BASES for up to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . This could lead to a total of 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months , since initial failure of the LCO, to restore the DC distribution system. At this time, an AC train could again become inoperable, and DC distribution restored OPERABLE.

This could continue indefinitely.

This Completion Time allows for an exception to the normal "time zero" for beginning the allowed outage time "clock."

This will result in establishing the "time zero" at the time the LCO was initially not met, instead of the time Condition BK was entered. The 16 hour1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months Completion Time is an acceptable limitation on this potential to fail to meet the LCO indefinitely.

D.1 and D.2 R-12 If the inoperable distribution subsystem cannot be restored to OPERABLE status within the required Completion Time, the unit must be brought to a MODE in which the LCO does not apply. To achieve this status, the unit 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 unit conditions from full power conditions in an orderly manner and without challenging plant systems.

E.1 Condition E addresses Wit-h--two trains with inoperable distribution subsystems that result in a loss of safety r function, adequate core cooling, containment OPERABILITY R-12 and other vital functions for DBA mitigation would be L-----

compromised-. Condition E also addresses two ICL3.8-214 (continued)

WOG STS Rev 1, 04/07/95 B 3.8.9-12 Markup for PI ITS Part E

Distribution Systems -Operating B 3.8.9 1PA3.8-100 BASES (continued) or more Reactor Protection Instrument AC Panels inoperable.

If the plant is in this Condition, and immediate plant [

shutdown in accordance with LCO 3.0.3 is required.R... SR-12 SURVEILLANCE SR 3.8.9.1 REQUIREMENTS This Surveillance verifies that the +required+ safeguards AC, DC, and Reactor Protection Instrument AC vital bus electrical power distribution systems, presented in Table B.3.8.9-1, are functioning properly, with the correct circuit breaker and switch alignment. The correct breaker and switch alignment ensures the appropriate separation and independence of the electrical divisions is maintained, and the appropriate voltage is available to each required subsystem-btrs. The verification of proper voltage availability on the buses ensures that the required voltage is readily available for motive as well as control functions for critical system loads ..nne.ted to these buses. Various indications are available to the operators which demonstrate correct voltage for the subsystems. The 7 day Frequency takes into account the redundant capability of the safeguards AC, DC, and Reactor Protection Instrument AC vital bus electrical power distribution subsystems, and other indications available in the control room that alert the operator to subsystem malfunctions.

REFERENCES 1. UFSAR, Section 86.apter [61.

2. UFSAR, Section 14Chapter E1!5. ICL3.8-172 1

3. Regulatory Guide 1.93, December 1974 WOG STS Rev 1, 04/07/95 B 3.8.9-13 Markup for PI ITS Part E

Distribution Systems - Operating B 3.8.9 JPA3 .8-100 Tklf *) 0 n i 1 :R. 1. 1 iULJI a )

rj %P .1. U C, 4-

.A.ý- A lq H Ap ý. PA Awizi Q ý. H Uri :5 TYP-EVOT T-RAIN A~ TRAIN1 1D*

AC buses E46 V ESP--Bus]-ENBOIT f- Bs N

-E4-SO-V] Load Centers Load Centers ENG91,NG031 NG02, GGO4]

-E480--V1 Motor Cantrol! Motor Contro Cent+ers Centers tNGOIA, -- 1-NeG[G0ANG 7 1 NG91B,-N~G93C- G0B N0 NG031,NG03D+ NGG4-I--NG04-0j E0V.D-i-3+ribut~i-n Distributo Panels Panels NG1[N4P02,N"O4 ENP0, DC busesE15VBusEK1 BsN02

-Bus-ENK03]- Bus[EN KO4]

Distribution1 Distributo PanelsPanels

[NK(41, NK43, NK(51] [NK42, NK44-,-*K52]

AC vital E1-20 V1 -Bus [NN011+- Bs[

Bus EN031Bus [ENNO4]

J r k-, 4-~-.- ~-r 4,1~

L, A f na 1nd elect'1rical~ powe~ur distiribuition svsteflfl - s subsystcmý.-

WOG STS Rev 1, 04/07/95 B 3.8.9-14 TSev10407/5 WOG PI ITS Part E forB.8.-14Markup

Distribution Systems-Operating B 3.8.9 Table B 3.8.9-1 (page 1 of 1)

Safeguards AC and DC Electrical Power Distribution Systems TYPE DISTRIBUTION UNIT 1 UNIT 2 EQUIPMENT TRAIN A AND B TRAIN A AND B Safeguards 4 kV Buses 15, 16 25, 26 AC 480 V Buses 111, 112, 121, 122 211, 212, 221, 222 Motor Control 1A1, 1A2 2A1, 2A2 Centers 1AB1*, 1AB2* 1AB1*, 1AB2*

1AC1, 1AC2 2AC1, 2AC2 1K1, 1K2, 1KA2 2K1, 2K2, 2KA2 iL1, 1L2 2L1, 2L2 1LA1, 1LA2 2LA1, 2LA2 1M1, 1M2 2M1, 2M2 1MAI*, 1MA2* IMAI*, 1MA2*

1R1, 1S1 2R1, 2S1 1Ti*, 1T2* 1T1*, 1T2*

ITA1, ITA2 2TA1, 2TA2 lXI, 1X2 2X1, 2X2 Safeguards 125 VDC Panels 11, 12 21, 22 DC 15, 16 25, 26 14*, 19* 14*, 19*

17*, 18* 17*, 18*

151, 161 27, 28 152, 162 251, 261 153, 163 252, 262 191 253, 263 Reactor 120 VAC Panels 111, 112, 113, 114 211, 212, 213, 214 Protection Instrument AC

Denotes MCC's or Panels that are transferrable between Ir - - - - - - -II units. R-12 I L--

WOG STS Rev 1, 04/07/95 B 3.8.9-15 Markup for PI ITS Part E

Part F Package 3.8 Part F Package 3.8 Difference Difference Category Number Justification for Differences 3.8-159 Not used.

CL 160 Additional discussion of the diesel fuel oil storage system is provided since Prairie Island has a unique design that involves many tanks, different systems for the emergency diesel generators (EDGs) for each unit and sharing of the storage capacity between the Unit 1 EDGs and the diesel driven cooling water pumps.

Prairie Island Units 1 and 2 28 4/1/02

Part F Package 3.8 Part F Package 3.8 Difference Difference Category Number Justification for Differences 3.8-PA 161 NUREG-1431, Rev. 2, LCO 3.8.6 Condition E and associated Bases have been revised by deleting the verbiage of one or more batteries. PI only has two batteries, one on each train. Therefore, PI physically cannot have more than one battery on a redundant train. In addition, the associated Completion Time has also been changed to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . The 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months is acceptable based on PI design, and in accordance with CTS which allows the batteries to be inoperable for 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months before initiating a reactor shutdown. P1 has installed batteries that far exceed the loads they would be required to provide in the event of an accident. Due to being oversized, if Condition E were entered, the batteries may still be capable of performing their intended safety function. Based on the above, requiring both batteries to be declared inoperable, in a time less than 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , would be inappropriate.

162 Not used.

Prairie Island Units 1 and 2 29 4/1/02

Part F Package 3.8 Part FPakg3.

Difference Difference Category Number Justification for Differences 3.8-CL 163 NUREG-1431, Rev 1, Bases 3.8.1 and 3.8.9, and NUREG-1431, Rev. 2, Bases 3.8.4 and 3.8.6, Applicable Safety Analyses Section have been revised by deleting, "or all onsite AC power". PI Safety Analysis for this system does not assume a loss of all onsite power. Therefore, this statement is deleted to be consistent with PI Safety Analysis.

164 Not used.

Prairie Island Units 1 and 2 30 4/1/02

Part F Package 3.8 Difference Difference Category Number Justification for Differences 3.8 165 Not used.

CL 166 NUREG-1431, Rev. 2, Bases 3.8.6, Action B has been revised deleting the following phrase, " ... and likely cannot perform its required design functions."

This statement is not applicable to PI. At P1, even though the battery charger is operating in the current limit mode for 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , the battery will still be able to perform its intended function. Therefore, this statement is deleted.

CL 167 NUREG-1431 Bases 3.8.9, has been revised throughout providing additional information or deleting detail in the Bases to make them more applicable to P1 design, operations, and testing. For example, in the LCO Section, deleted the paragraph discussing tie breakers between redundant trains since PI design does not include tie breakers between the trains.

Prairie Island Units 1 and 2 31 4/1/02

Part F Package Part F Package 3.8 3.8 Difference Difference Category Number Justification for Differences 3.8-PA 168 NUREG-1431 SR 3.8.10.1 has been revised by adding the following, "Verify correct breaker "and switch" alignments ... ." Adding switches makes this a more accurate SR for the PI design. PI has both breakers and switches in the safeguards AC, DC, and Reactor Protection Instrument AC electrical power distribution subsystems. This change is consistent with PI design and current operating practices.

PA 169 NUREG-1431 Bases 3.8.9, Action C.1, is being revised by deleting the following sentence, "The 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months Completion Time for DC buses is consistent with Regulatory Guide 1.93 (Ref. 3)." This sentence is being deleted because it is not referenced in the subject Regulatory Guide; in addition, reference JFD CL3.8-172 for PI position on Regulatory Guides.

170 Not used.

Prairie Island Units 1 and 2 32 4/1/02

Part F Package 3.8 Difference Difference Category Number Justification for Differences 3.8 CL 171 NUREG-1431, Rev. 2, LCO 3.8.4, Required Actions A.4 and B.4 Completion Times and their associated Bases have been revised to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months to be consistent with CLB as in the PI CTS 3.7.B.7 and 3.7.B.8. In addition, LCO 3.8.5 Required Action A.3 and LCO 3.8.6 Required Actions A.1, A.2, and B.1 have been changed to be consistent with LCO 3.8.4 Required Actions A.4 and B.4. Maintaining CLB was agreed to be acceptable between the industry and NRC during the onset of the ITS conversion project. This change is consistent with that agreement.

CL 172 NUREG-1431 Bases 3.8 has been revised deleting references to specific Regulatory Guides, IEEE Standards, and 10 CFR 50 criteria that PI is either not committed or designed to. PI was designed, built, and licensed prior to 10 CFR 50 Appendix A GDC and other noted NRC/industry design criteria. Where specific Industry Standards or Regulatory Guides are referenced, within the ITS, it does not mean PI is committing to them. They are only used as reference to support the ITS Bases or NRC criteria, Frequencies, SRs, etc., that are consistent with PI CLB.

CL 173 NUREG-1431, Rev. 2, Bases 3.8.4, Background Section, has been revised by changing the information about the "spare" battery charger to be applicable to the PI "portable" charger design and usage. At PI, there is a portable battery charger that may be moved into place to be used in either unit. This portable charger has been approved in the PI initial SER.

Prairie Island Units I and 2 33 4/1/02

Part F Package 3.8 Part F Package 3.8 Difference Difference Category Number Justification for Differences 3.8 174 Not used.

TA 175 NUREG-1431 LCO 3.8.5, 3.8.8, and associated Bases have been revised consistent with the guidance of TSTF-204, Rev. 3.

PA 176 NUREG-1431 Bases 3.8.8, LCO Section has been revised by deleting the first sentence. This is consistent with TSTF-204, Rev. 3, which clarifies that safety analyses for Shutdown MODES operation does not consider Operating DBA's. The sentence is not consistent with PI CLB since PI does not currently have Shutdown Technical Specifications.

CL 177 NUREG-1431 Bases 3.8.8 has been revised providing additional information and clarification consistent with PI design, terminology, and operating practices, since PI does not currently have Shutdown Technical Specifications. This clarification takes TSTF-204, Rev.

3, into account.

Prairie Island Units 1 and 2 34 4/1/02

Part F Package 3.8 Part F Package 3.8 Difference Difference Category Number Justification for Differences 3.8-181 Not used.

CL 182 NUREG-1431, Rev. 2, SR 3.8.4.2 and associated Bases, have been revised replacing the statement, "largest combined demands of the various continuous steady state loads ... ." with the statement, " ...

demands of the various continuous steady state loads, .... " PI battery charger design requirements were based on the demands of the various continuous steady state loads not the largest combined demands of the of the various continuous steady state loads. Revising this statement as proposed, brings the SR into agreement with the PI design and consistent with the PI USAR.

Prairie Island Units 1 and 2 36 4/1/02

Part F Package 3.8 Difference Difference Category Number Justification for Differences 3.8 CL 183 NUREG-1431 LCO 3.8.7 requires four inverters to be OPERABLE when in MODES 1, 2, 3, or 4.

LCO 3.8.7, Condition A, provides Required Actions A.1 and A.2 in the event that one Reactor Protection Instrument AC inverted is inoperable. In this case, Required Action A.1 requires verifying only one Reactor Protection Instrument AC panel is powered from Panel 117. Required Action A.2 requires verification that only one Reactor Protection Instrument AC Panel is powered from its inverter bypass source. These Required Actions verify and ensure that only one Reactor Protection Instrument AC Panel is powered from Panel 117 (Unit 2 - 217) or powered from its bypass source in accordance with PI design. P1 design does not allow long term operation with one Reactor Protection Instrument AC Panel powered from Panel 117 and at the same time another Reactor Protection Instrument AC Panel supplied power from its bypass source. Required Action A.1 requires verifying only one Reactor Protection Instrument AC panel is powered from Panel 117. Alternatively, Required Action A.2 requires verification that only one Reactor Protection Instrument AC Panel is powered from its inverter bypass. In the Condition when two Reactor Protection Instrument AC inverters are inoperable, PI would enter Condition B which requires verification that at most one Reactor Protection Instrument AC panel is powered from Panel 117 (217) and one or both Reactor Protection Instrument AC panel(s) are powered from their inverter bypass source. In this case, one inverter would have to be restored in 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

Prairie Island Units 1 and 2 37 4/1/02

Part F Package 3.8 Difference Difference Category Number Justification for Differences 3.8-CL 183 (continued)

Required Action A.1 (ITS Required Action B.3)

Completion Time and associated Bases have been decreased from 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . The decrease in Completion Time to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months is consistent with the CTS.

Both the new Required Action A.1 and A.2 are CLB and have been incorporated based on PI design.

Maintaining CLB was agreed to be acceptable between the industry and NRC during the onset of the ITS conversion project. This change is consistent with that agreement.

184 Not used.

Prairie Island Units 1 and 2 38 4/1/02

Part F Package 3.8 Part F Package 3.8 Difference Difference Category Number Justification for Differences 3.8-PA 185 NUREG-1431 Bases 3.8.7 LCO Section contains an explanation of the Note which allows an instrument bus inverter to be disconnected from its associated DC bus for up to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months while performing an equalizing charge on the battery. The inverters used at PI are not required to be disconnected during equalizing charges.

Therefore, this Note has been deleted consistent with ITS.

186 Not used.

PA 187 NUREG-1431, Rev. 2, LCO 3.8.6 and associated Bases, Completion Time for Required Action B.2 has been revised to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . PI does not have either this Required Action or Completion Time in the CTS.

The 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months is in accordance with PI USAR 8.5.2, which states, "Each of the four battery chargers has been sized to recharge its associated partially discharged battery within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , while carrying its normal load." Therefore, the 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months is consistent with PI licensing basis.

188 Not used.

189 Not used.

Prairie Island Units 1 and 2 39 4/1/02

Part F Package 3.8 Part F Package 3.8 Difference Difference Category Number Justification for Differences 3.8 193-196 Not used.

CL 197 NUREG-1431 LCO SR 3.8.1.5 and associated Bases have been deleted. PI day tanks are not designed with any type of drain in the tank that would allow draining any water. PI operating history has shown that the day tanks have not had any water accumulation problems.

In addition, neither PI CTS or CLB require checking the day tanks for water; therefore, this SR is being deleted.

Prairie Island Units 1 and 2 42 5/01/01

Part F Package 3.8 Part F Package 38 Difference Difference Category Number Justification for Differences 3.8 PA 198 NUREG-1431, Rev. 2, LCO 3.8.5, Condition A, and associated Bases, have been revised to delete the Condition phrase, "The redundant train battery and charger[s] OPERABLE." Per PI design, two trains of DC power are not required to be OPERABLE to support plant DC shutdown requirements as identified by LCO 3.8.10. Therefore, this part of Condition A does not apply to PI and is being deleted. In addition, the word "required" has been added as appropriate. Since PI has two trains, with each train consisting of a battery, battery charger, and interconnecting cable, it is necessary for clarification to state the "required" battery therefore, no mistake can be made on which battery charger is being credited when the plant is in the shutdown condition. ISTS LCO 3.8.5, Required Actions A.1 and A.2 have also been deleted to be consistent with LCO 3.8.4, Required Actions A.1 and A.2. PI design has a portable battery charger which can be used to replace the inoperable battery charger. The ITS allows 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months for the portable charger to be installed or declare the inoperable charger to OPERABLE. In either event, since PI has 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months to get a charger OPERABLE, Required Actions A.1 and A.2 would not be applicable and are therefore deleted.

Prairie Island Units 1 and 2 43 4/1/02

Part F Package 3.8 Difference Difference Category Number Justification for Differences 3.8-PA 199 NUREG-1431,Rev. 2, LCO 3.8.5, Required Action B.1, and associated Bases, have been deleted. This Required Action requires declaring the affected required features(s) inoperable. This action is only applicable if there were more than one DC electrical power subsystems required to be OPERABLE. Since PI design and shutdown operations do not require more than one DC electrical power subsystem to be OPERABLE, this Required Action does not apply.

CL 200 NUREG-1431 Bases 3.8 has been revised to reflect current PI design and operating practices. As an example, Bases 3.8.1, Required Action B.2 states, "This includes motor driven auxiliary feedwater pumps.

Single train systems, such as turbine driven auxiliary feedwater pumps, are not included." PI has two 100%

capacity auxiliary feedwater pumps, a motor and a turbine driven. The turbine driven auxiliary feedwater pump is not supported by the DG. Therefore, this statement is not applicable to P1 design and is deleted.

Prairie Island Units I and 2 44 5/01/01

Part F Package 3.8 Difference Difference Category Number Justification for Differences 3.8 CL 201 NUREG-1431 Bases 3.8.1, Background Section has been revised by adding the statement, "... the Unit 1 DGs meet the intent of Safety Guide 9 and Unit 2 DGs satisfy the intent of Regulatory Guide 1.9, .... " This statement was added to reflect the differences between the two unit DGs. Unit 1 DGs were installed prior to the issuance of Regulatory Guide 1.9. Therefore, Unit 1 DGs rating were consistent with Safety Guide 9. When Unit 2 DGs were installed, Regulatory Guide 1.9 has been issued; however, PI did not adopt this Regulatory Guide in its entirety as discussed in the PI USAR. This change is consistent with the PI CLB.

CL 202 NUREG-1431 Bases 3.8.1 and 3.8.2, LCO Section have been revised by replacing the statement, "This will be accomplished ... ." with "The DG will be ready to load ... following receipt of a start signal." PI design is that each DG is capable of starting, accelerating to the required speed and voltage, and ready to be loaded within 10 seconds. PI DGs are not required to be loaded within 10 seconds. In addition, Bases 3.8.2, LCO statement has been revised by deleting the statement, "This sequence must be accomplished within [10] seconds." As stated above, the PI DGs are required to be ready to load within 10 seconds upon receipt of a start signal. Therefore, the Bases is revised to reflect the PI design and CLB.

Prairie Island Units 1 and 2 45 12/11/00

Part F Package 3.8 Difference Difference Category Number Justification for Differences 3.8 PA 203 NUREG-1431, Rev. 2, Bases 3.8.4 Actions B Section has been revised deleting the following sentence, "In addition the energization transients of any DC loads that are beyond the capability of the battery charger[s] and normally require the assistance of the batter[y][ies] will not be able to be brought online." PI design does not have any energization transients that exceed the battery charger capacity. The PI battery chargers were designed and installed to handle any of the anticipated transients that they would experience. Therefore, this statement is not applicable to Pl.

PA 204 NUREG-1431 Bases 3.8.1, LCO Section has been revised by deleting the subject paragraphs. The subject paragraphs discuss various information about the AC sources in a train and the AC offsite sources being independent and separated to the extent practical. PI USAR provides a detailed discussion about the design of the AC trains and offsite sources; therefore, this redundant information is not needed in the TS and is being deleted.

Also, Bases SR 3.8.1.1 has been revised by editing the sentence discussing preferred power source. PI design does not identify a preferred power source. The correct plant terminology is offsite power source.

Prairie Island Units 1 and 2 46 7/02/01

Part F Package 3.8 Difference Difference Category Number Justification for Differences 3.8 PA 205 NUREG-1431 Bases 3.8.1, 3.8.4, 3.8.7, and 3.8.9 Applicability Sections have been revised by deleting the following, " ... or abnormal transients;" PI considers an abnormal transient as an AOO. Therefore, the specific reference to an abnormal transient is being deleted.

CL 206 NUREG-1431 Bases 3.8.1, Condition C is for two paths inoperable. Required Action C.1 states to declare required feature(s) inoperable when its redundant required feature(s) is inoperable with a Completion Time of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . The ISTS states that the justification for the 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is Regulatory Guide 1.93. PI CTS already has a Completion Time of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . Therefore, any references in the ISTS to the Completion Time being shorter or reduced is deleted.

CL 207 NUREG-1431 Bases 3.8.1, Required Action C.1 and C.2 have been revised by deleting the subject discussions since they are referring to Regulatory Guide 1.93. Since P1 is not committed to Regulatory Guide 1.93, the subject discussions are not applicable to PI.

PA 208 NUREG-1431 Bases 3.8 has been revised deleting redundant information that also appears in the USAR.

209 Not used.

Prairie Island Units 1 and 2 47 12/11/00

Part F Package 3.8 Difference Difference Category Number Justification for Differences 3.8 CL 210 NUREG-1431 Bases 3.8.2, LCO statement has been revised deleting the following, "It is acceptable for trains to be cross tied during shutdown conditions, allowing a single offsite power circuit to supply all required trains."

PI design does not provide a cross tie between the trains. The design, as described in the USAR, provides for each offsite source being capable of supplying both trains, but this not termed a cross tie.

PA 211 NUREG-1431 LCO 3.8.2 and associated Bases has been revised by adding a Note allowing the LCO not being applicable for a period of 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months during the performance of SR 3.8.1.10. Without the Note, the LCO requires that one DG capable of supplying one train of the onsite 4 kV safeguards distribution system required by LCO 3.8.10 be OPERABLE. SR 3.8.2.1 requires the SRs of Specification 3.8.1 be performed at their specified Frequencies for those AC sources that are required to be OPERABLE to support those systems operating during plant shutdown. One of these SRs requires DG testing. At PI, when a DG is being tested, and thus operating, it is considered to be inoperable since during this testing some controls must be placed in manual. SR 3.8.1.10 in particular results in considering both DGs inoperable during test performance. The 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months period is reasonable to allow performance of the required SR.

Prairie Island Units 1 and 2 48 12/11/00

Part F Package 3.8 Difference Difference Category Number Justification for Differences 3.8 PA 212 NUREG-1431, Rev. 2, LCO 3.8.5 and associated Bases Background Section has been revised by adding a NOTE stating, service building DC electrical power subsystem components may be used in lieu of a safeguards DC electrical power subsystem component when the required safeguards DC electrical power subsystem is inoperable due to testing, maintenance, or replacement. PI design comprises of one battery, battery charger, and interconnecting cabling for each train.

Since P1 only has two trains of safeguards DC electrical power, during an outage only one train is required to be OPERABLE to support plant operations. The other train may be inoperable. At times during the outage, one train will be inoperable with the other needing testing, or even replacement.

Based on new shutdown requirements, one train must remain OPERABLE, therefore requiring an extension in the outage schedule in order to accomplish needed maintenance, testing or replacement.

PI design has two service building DC electrical power subsystems from which components may be used in lieu of either safeguards DC electrical power subsystem components. This is acceptable since the service building batteries are maintained in accordance with TS 5.5.15, Battery Testing Program, and the service building DC electrical power components will be maintained the same as the safeguards DC electrical power subsystem components.

Prairie Island Units 1 and 2 49 4/1/02

Part F Package 3.8 Part FPakg38 Difference Difference Category Number Justification for Differences 3.8 PA 212 (continued)

Plant procedures will ensure that the DC electrical power subsystem components will perform their intended safety function. The time in which the service building DC electrical subsystem components can be used in lieu of the safeguards DC electrical power will be limited to the time the safeguards DC electrical power subsystem components are inoperable due to maintenance, testing, or replacement.

CL 213 NUREG-1431 LCO 3.8.9, Required Actions A, B, and C and their associated Bases have been revised by adding the following, "Declare associated required supported feature(s) inoperable, immediately." This Action needed to be added to provide guidance for when a portion of safeguards AC, DC, and Reactor Protection Instrument AC electrical power distribution subsystems are inoperable. This Condition is not covered by the ITS. This change is consistent with P1 practices. In addition, a Note was added stating to enter the applicable Condition and Required Actions of LCO 3.8.4 for DC trains made inoperable by inoperable power distribution subsystems. This Note provides clarification for the operator. This change is consistent with NUREG-1431, Rev. 2.

Prairie Island Units 1 and 2 50 4/1/02

Part F Package 3.8 Difference Difference Category Number Justification for Differences 3.8 CL 214 NUREG-1431 LCO 3.8.9 Required Action E has been revised by adding the following, "Two or more Reactor Protection Instrument AC Panels inoperable, Enter LCO 3.0.3, Immediately." This Required Action has been added to provide specific Actions when two or more Reactor Instrument AC panels are inoperable, since the instrument AC panels are distinct from "Two trains...". The ISTS does not currently specify this condition.

PA 215 NUREG-1431, Rev. 1 LCO 3.8.2, Required Action A.1, LCO 3.8.8, Required Action A.1 and associated Bases have been deleted. The rational for the subject Required Actions A.1 was based on NUREG 1431, Rev.1 which would, in certain conditions, require more than one safeguards bus or inverter required to be OPERABLE. With one of two or more required safeguards bus or inverter inoperable, the remaining safeguards bus(s) or inverter(s) might be able to power all necessary loads. In such a case, it is acceptable to declare inoperable required features associated with the inoperability. However, with only one safeguards bus or inverter required, the above conditions do not exist, and the option to declare required features inoperable is not appropriate.

Therefore, Required Action A.1 is being deleted.

Prairie Island Units 1 and 2 51 4/1/02

Part F Package 3.8 Difference Difference Category Number Justification for Differences 3.8 PA 218 NUREG-1431 LCO 3.8.3, Required Action D, has been revised by a Note stating, "Enter applicable Conditions and Required Actions of LCO 3.7.8, 'CL System' when Condition D is entered as a result of stored fuel oil properties not within limits". The requirements for diesel fuel oil volume have been divided into two separate specifications. ITS 3.7.8 provides diesel fuel oil volume for the diesel driven CL pumps and ITS 3.8.3 provides the requirements for the diesel generators. In addition, ITS 5.5.11 provides the requirements for the Diesel Fuel Oil Testing Program. Both the diesel driven CL pumps and the plant diesel generators share a common storage tank and fuel oil contents. ITS 3.8.3 provides requirements for testing the tank contents and associated Required Actions if the fuel oil properties are not restored to within limits. A Note was added to Condition D instructing the operators that if the diesel fuel oil in the storage tanks is not within limits, to enter the associated Conditions and Required Actions for the diesel driven CL pumps. This change provides consistency between the two systems and is consistent with current plant design and practices.

Prairie Island Units 1 and 2 53 4/1/02

Part G Package 3.8 M - More restrictive (GENERIC NSHD)

(M3.8-04, M3.8-06, M3.8-14, M3.8-18, M3.8-19, M3.8-21, M3.8-24, M3.8-27, M3.8-31, M3.8-41, M3.8-42, M3.8-47, M3.8-49, M3.8-50, M3.8-52, M3.8-55, M3.8-64, M3.8-65, M3.8-66, M3.8-67)

This proposed TS revision involves modifying the CTS to impose more stringent requirements upon plant operations to achieve consistency with the guidance of NUREG-1431, correct discrepancies or remove ambiguities from the specifications.

These more restrictive TSs have been evaluated against the plant design, safety analyses, and other TS requirements to ensure the plant will continue to operate safely with these more stringent specifications.

1. The proposed amendment will not involve a significant increase in the probability or consequences of an accident previously evaluated.

The proposed changes provide more stringent requirements for operation of the plant. These more stringent requirements do not result in operation that will increase the probability of initiating an analyzed event and do not alter assumptions relative to mitigation of an accident or transient event.

These more restrictive requirements continue to ensure process variables, structures, systems, and components are maintained consistent with the safety analyses and licensing basis. Therefore, these changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. The proposed amendment will not create the possibility of a new or different kind of accident from any accident previously analyzed.

The proposed changes do not involve any physical alteration of the plant, that is, no new or different type of equipment will be installed, nor do they change the methods governing normal plant operation.

Prairie Island Units 1 and 2 4 4/1/02

Current Technical Specification Cross-Reference CTS Section CTS Table Section Type ITS Section ITS Table Item Number Item Number 4.6.B.3 Relocated TRM 4.6.B.4 (Partial) Relocated TRM 4.6.B.4 3.8.6.6 4.6.B.5 Deleted SR 3.8.4.2 New New SR 3.8.4.3 New SR 3.8.7.1 4.6.C SR 3.4.9.2 4.6.C (Partial) Relocated Bases Prairie Island Units 1 and 2 4.6-3 4/1/02

Improved Technical Specification Cross-Reference ITS Section ITS Table Section Type CTS Section CTS Table Item Number Item Number 3.8.4 LCO 3.7.B.7 3.8.4 LCO 3.7.B.8 3.8.4.1 SR New 3.8.4.2 SR New 3.8.4.3 SR New 3.8.5 LCO New 3.8.5.1 SR New 3.8.6 LCO New 3.8.6.1 SR New 3.8.6.2 SR 4.6.B.1 3.8.6.3 SR 4.6.8.2 3.8.6.4 SR 4.6.8.1 3.8.6.5 SR New 3.8.6.6 SR 4.6.B.4 3.8.6.6 SR New 3.8.7 LCO 3.7.A 3.8.7 LCO 3.7.A.7 3.8.7 LCO 3.7.B 3.8.7 LCO New Prairie Island Units 1 and 2 3.8-3 4/1/02

Containment Penetrations 3.9.4 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME 4-A. One or more containment A.*1 Suspend movement of Immediately penetrations not in irradiated fuel assemblies required status. within containment.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.9.4.1 Verify each required containment penetration is in 7 days the required status.

SR 3.9.4.2 ---------------------------- NOTE ---------------------------

Not required to be met for containment purge (high flow) and inservice (low flow) purge valve(s) in penetrations closed to comply with LCO 3.9.4.c.1.

Verify each required containment purge (high flow) 24 months and inservice (low flow) purge system valve actuates to the isolation position on an actual or simulated actuation signal.

Prairie Island Units 1 and 2 3.9.4-2 4/1/02

RHR and Coolant Circulation-High Water Level 3.9.5 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME A. (continued) A.3 Initiate action to satisfy Immediately RHR loop requirements.

AND A.4 Close equipment hatch and 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months secure with four bolts.

AND A.5 Close one door in each 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months airlock.

AND A.6.1 Close each penetration 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months providing direct access from the containment atmosphere to the outside atmosphere with a manual or automatic isolation valve, or blind flange.

OR A.6.2 Verify each penetration is 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months capable of being closed by an OPERABLE Containment Ventilation Isolation System.

Prairie Island Units 1 and 2 3.9.5-2 4/1/02

Containment Penetrations B 3.9.4 BASES BACKGROUND The containment air locks, which are also part of the containment (continued) pressure boundary, provide a means for personnel access during MODES 1, 2, 3, and 4 unit operation in accordance with LCO 3.6.2, "Containment Air Locks." Each air lock has a door at both ends.

The doors are normally interlocked to prevent simultaneous opening when containment OPERABILITY is required. During periods of unit shutdown when containment closure is not required, the door interlock mechanism may be disabled, allowing both doors of an air lock to remain open for extended periods when frequent containment entry is necessary.

During movement of irradiated fuel assemblies within containment, containment closure or closure capability is required; therefore, the door interlock mechanism may remain disabled and both doors may be open provided one door can be closed with at least two containment fan coil unit fans capable of operating in high speed.

The requirements for containment penetration closure ensure that a release of fission product radioactivity within containment will restrict fission product radioactivity release from containment to be within regulatory limits.

The Containment Purge and Exhaust System includes two subsystems, Containment Purge and Containment Inservice Purge.

The containment purge subsystem includes a 36 inch purge penetration and a 36 inch exhaust penetration. The second subsystem, a minipurge system referred to as containment inservice purge, includes a 14 inch purge penetration and an 18 inch exhaust penetration.

During MODES 1, 2, 3, and 4, the two valves in each of the normal purge and exhaust penetrations are secured in the closed position, or the penetrations may be blank flanged. The two valves in each of the two containment inservice purge penetrations can be opened intermittently, but are closed automatically by the Containment Ventilation Isolation System.

Prairie Island Units 1 and 2 B 3.9.4-2 4/1/02

Containment Penetrations B 3.9.4 BASES BACKGROUND In MODE 6, sufficient air flow rates are necessary to conduct (continued) refueling operations. The inservice purge system is used for this purpose, and each of the four valves is closed by the radiation monitors associated with the containment inservice purge system in accordance with LCO 3.3.5, "Containment Ventilation Isolation Instrumentation." The 36 inch subsystem is normally blank flanged, although the option for use is allowed during outages, except during movement of irradiated fuel or CORE ALTERATIONS. All four containment purge valves are also closed by the Containment Ventilation Isolation System.

The other containment penetrations that provide direct access from containment atmosphere to outside atmosphere must be isolated on at least one side. Isolation may be achieved by an OPERABLE automatic isolation valve, or by a manual isolation valve, or blind flange.

APPLICABLE During CORE ALTERATIONS or movement of irradiated fuel SAFETY assemblies within containment, the most severe radiological ANALYSES consequences result from a fuel handling accident. The fuel handling accident is a postulated event that involves damage to irradiated fuel (Ref. 1). Fuel handling accidents include dropping a single irradiated fuel assembly and handling tool or a heavy object onto other irradiated fuel assemblies. The requirements of LCO 3.9.2, "Refueling Cavity Water Level," in conjunction with the minimum decay time of 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months prior to irradiated fuel movement with containment closure capability ensure that the release of fission product radioactivity, subsequent to a fuel handling accident, results in doses that are well within the guideline values specified in 10 CFR 100. The acceptance limit for offsite radiation exposure is 25% of 10 CFR 100 values.

Prairie Island Units 1 and 2 B 3.9.4-3 4/1/02

Containment Penetrations B 3.9.4 BASES APPLICABLE The requirements for containment penetration closure ensure that a SAFETY release of fission product radioactivity within containment will ANALYSES restrict fission product release from containment to be well within (continued) regulatory limits. The closure restrictions are sufficient to restrict fission product radioactivity release from containment due to a fuel handling accident during refueling.

A fuel handling accident does not cause containment pressurization; however, with an assumed single failure, the operating purge system supply fan is assumed to continue supplying air to containment. To maintain post-fuel handling accident releases well within the limits of 10 CFR 100, only the inservice purge system is allowed to be operating during fuel movement. Two fan coil unit fans are required to operate in the high speed mode following a fuel handling accident to assure that radioactive material in containment is well mixed and any releases will leave containment at a lower concentration over the duration of the accident. The provision that one air lock door is OPERABLE and under procedural control will assure that at least one door will be closed within 30 minutes as required, thus assuring radioactive releases are well within the limits of 10 CFR 100 (Ref.

1).

Containment penetrations satisfy Criterion 3 of 10 CFR 50.36 (c)

(2)(ii).

LCO This LCO limits the consequences of a fuel handling accident in containment by limiting the potential escape paths for fission product radioactivity released within containment.

The LCO requires containment penetrations to meet the following requirements:

Prairie Island Units 1 and 2 B 3.9.4-4 12/11/00

Containment Penetrations B 3.9.4 BASES LCO a. The equipment hatch is closed and held in place by at least 4 (continued) bolts;

b. One door in each air lock is closed, or both doors in each air lock may be open with:

1. containment (high flow) purge system isolated,

2. one air lock door OPERABLE and capable of being closed within 30 minutes, and

3. at least two containment fan coil unit fans capable of operating in the high speed mode; and

c. At least one isolation valve in each penetration, including the containment (high flow) purge system and inservice (low flow) purge system, providing direct access from the containment atmosphere to the outside atmosphere is either:

1. OPERABLE or closed by a manual valve, or blind flange, or

2. capable of being closed by an OPERABLE Containment Ventilation Isolation System.

A penetration with direct access from the containment atmosphere to the outside atmosphere includes all penetrations that have a flow path that leads anywhere outside containment.

The containment air lock doors may be open during movement of irradiated fuel in the containment provided that the LCO requirements are met. These requirements include one door Prairie Island Units 1 and 2 B 3.9.4-5 1/21/02

Containment Penetrations B 3.9.4 BASES LCO OPERABLE, under procedural control and capable of being closed (continued) within 30 minutes following a fuel handling accident in containment and at least two fan coil unit fans are capable of operating in the high speed mode following a fuel handling accident in containment.

Should a fuel handling accident occur inside containment, the fan coil unit fans will be operated in high speed and one door in each air lock will be closed following an evacuation of containment.

For the OPERABLE containment purge and exhaust penetrations, this LCO ensures that these penetrations are isolable by the Containment Ventilation Isolation System. The OPERABILITY requirements for this LCO require that the automatic purge and exhaust valve closure can be achieved and, therefore, meet the assumptions used in the safety analysis to ensure that releases through the valves are terminated, such that radiological doses are within the acceptance limit.

APPLICABILITY The containment penetration requirements are applicable during movement of irradiated fuel assemblies within containment because this is when there is a potential for the limiting fuel handling accident.

In MODES 1, 2, 3, and 4, containment penetration requirements are addressed by LCO 3.6.1.

In MODES 5 and 6, when movement of irradiated fuel assemblies within containment is not being conducted, the potential for a fuel handling accident does not exist. Therefore, under these conditions no requirements are placed on containment penetration status.

Prairie Island Units 1 and 2 B 3.9.4-6 4/1/02

Containment Penetrations B 3.9.4 BASES (continued)

ACTIONS A. 1 If the containment equipment hatch, air locks, or any containment penetration that provides direct access from the containment atmosphere to the outside atmosphere is not in the required status, including the Containment Ventilation Isolation System not capable of automatic actuation when the purge and exhaust valves are open, the unit must be placed in a condition where the isolation function is not needed. This is accomplished by immediately suspending movement of irradiated fuel assemblies within containment.

Performance of these actions shall not preclude completion of movement of a fuel assembly to a safe position.

SURVEILLANCE SR 3.9.4.1 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 open purge and exhaust valves will demonstrate that the valves will function if required during a fuel handling accident. Also the Surveillance will demonstrate that each valve operator has motive power, which will ensure that each valve is capable of being closed by an OPERABLE automatic Containment Ventilation Isolation signal.

The Surveillance is performed every 7 days during movement of irradiated fuel assemblies within containment. The Surveillance interval is selected to be commensurate with the normal duration of time to complete fuel handling operations. A surveillance is to be conducted before the start of refueling operations and then in accordance with the frequency specified. As such, this Surveillance ensures that a postulated fuel handling accident that releases fission product radioactivity within the containment will not result in a release of significant fission product radioactivity to the environment.

Prairie Island Units 1 and 2 B 3.9.4-7 4/1/02

Containment Penetrations B 3.9.4 BASES SURVEILLANCE SR 3.9.4.2 REQUIREMENTS (continued) This Surveillance demonstrates that each containment purge and exhaust valve actuates to its isolation position on manual initiation or on an actual or simulated high radiation signal. The 24 month Frequency maintains consistency with other similar ESFAS instrumentation and valve testing requirements. In BASES LCO 3.3.5, the Containment Ventilation Isolation instrumentation requires a CHANNEL CHECK every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and a COT every 92 days to ensure the channel OPERABILITY during refueling operations. Every 24 months a CHANNEL CALIBRATION is performed. SR 3.6.3.5 demonstrates that the isolation time of each valve is in accordance with the Inservice Testing Program requirements. These Surveillances, when performed, 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.

The SR is modified by a Note stating that this Surveillance is not required to be met for valves in isolated penetrations. The LCO provides the option to close penetrations in lieu of requiring automatic actuation capability.

REFERENCES 1. USAR, Section 14.5.

Prairie Island Units 1 and 2 B 3.9.4-8 4/l/02

T-S 3; 8--

GOnrf lo; continued 3- If Specification 3.8.A.l.f or 3.8.A.l.g cannot be satisfied, all fuel handling operations in the core c.ntai.nm.nt shall be RAA . 2 1 suspended (3.8.A.I.f (ITS 3.9.5) ,the requirements of LC03I.9.

I JA3.9-47 ]

LC03. 9.5 Specification 3.8.A.l.a.l (Close the equipment hatch and J 9-50 RA A.4, penetrations) shall be satisfied, A.5,A.6.1 LC03.9.6 at least one door in each personnel air lock shall be closed, RA B.3, B.4, B.5.1 LC03. 9.5 a a ad no reduction in reactor coolant boron concentration less than RA A. 1 S-- -*m. T(C h~11 he ma9e.

r ,--I LCO3 .9. 6 RA B.1 LCO3 . 9 . 5 RA A.6. 2 Verify each penetration is capable of being closed by an OPERABLE2 LC03 .9 6 containment ventilatioln isolation system. :R-12 RA B.5.2 PI Current TS 6 of 10 Markup for PI ITS Part C

Part D Package 3.9 NSHD Change Category Number Discussion of Change 3.9 M 16 New SRs, 3.9.4.1 and 3.9.4.2 are included which require verification of containment penetration status every 7 days and verification of containment purge and inservice purge valve actuation every 24 months. The 7 day frequency for containment penetration status verification is commensurate with the normal duration of time to complete fuel handling operations. The 24 month Frequency for verification of containment purge and inservice purge valve actuation is consistent with a 24 month refueling outage interval and will allow this verification to be performed during each refueling outage. ITS SR 3.9.4.2 is modified by a note which does not require the SR to be met when containment purge and inservice purge valves are closed in compliance with LCO 3.9.4 requirements for penetrations to be isolated. This is an acceptable exception since penetrations that are closed in compliance with the LCO do not have to be tested to assure that they can be automatically closed. These are activities which are currently performed under plant procedures; therefore this change does not adversely impact plant operations. Since these will be formal TS required surveillances, this change is considered more restrictive. This change is included to make the PI ITS complete and consistent with the guidance of NUREG-1431.

Prairie Island Units I and 2 8 4/1/02

Containment Penetrations 3.9.4 ArCTTNN CONDITION REQUIRED ACTION COMPLETION TIME A. One or more A.1 Suspend COR Immediately AL TE-"R'A TITONf&l containment penetrations not in I TA3.9-66 required status. AND Suspend movement of Imme diatcy irradiated fuel assemblies within containment.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.9.4.1 Verify each required containment 7 days penetration is in the required status.

SR 3.9.4.2 ------------------- NOTE --------------- ITA3.9-67 Not required to be met for containment purge (high flow) and inservice (low flow) PA.964j purge valve(s) in penetrations closed to comply with LCO 3.9.4.c.I. 1R-12 X3.9-61 L ------

Verify each required containment purge (high flow) and inservice (low flow) purge systeiand exhaus-t valve actuates to the isolation position on an actual or r

1R-12 I1 1 - - - -i simulated actuation signal.

24{1-8 months WOG STS Rev 1, 04/07/95 3.9.4-2 Markup for PI ITS Part E

RHR and Coolant Circulation- High Water Level 3.9.5 ACTIONS CONDITION IREQUIRED ACTION COMPLETION TIME A. (continued) A.4 Close equipment hatch 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months and secure with four boltsal 1 n.nt "ntai penetrati m-s TA3"9-69]

providing CU.direet J l

IIII.I arceess QL..... ,--II4 f-roe.4" UU - L un tainmenf

+. mlllue.r, I;I I e L.

A k- I -

AND A.5 Close one door in 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months each air lock.

AND A.6.1 Close each 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months penetration providing direct access from the containment atmosphere to the outside atmosphere with a manual or automatic isolation valve, or blind flange.

OR A.6.2 Verify each 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months penetration is capable of being R-12 closed by an OPERABLE L----------

Containment Ventilation Isolation System.

______________________________ I __________________________________ L ___________________

WOG STS Rev 1, 04/07/95 3.9.5-3 Markup for PI ITS Part E

Containment Penetrations B 3.9.4 BASES four bolts. Good engineering practice dictates that the bolts required by this LCO be approximately equally spaced.

The containment air locks, which are also part of the containment pressure boundary, provide a means for personnel access during MODES 1, 2, 3, and 4 unit operation in accordance with LCO 3.6.2, "Containment Air Locks." Each air lock has a door at both ends. The doors are normally interlocked to prevent simultaneous opening when containment OPERABILITY is required. During periods of unit shutdown BACKGROUND when containment closure is not required, the door interlock (continued) mechanism may be disabled, allowing both doors of an air lock to remain open for extended periods when frequent containment entry is necessary.

During CRE ALTE-,RATIOS or movement of irradiated TA3.9-66 fuel assemblies within containment, containment closure or closure capability is required; therefore, the door interlock mechanism may remain disabled and both doors may be open provided one CL3.9-62 door can be closed with at least two containment fan coil unit fans capable of operating in high speed but one air lock door must always rem.ain losed.

The requirements for containment penetration closure ensure that a release of fission product radioactivity within containment will be rstricted from escaping

-to the environment. The lsure restrictIOn*s are CL3 .9-62 sufficient to restrict fission product radioactivity release from containment to be within regulatory limitsdtte to a fuel handling accident during refucig PA3.9-101 The Containment Purge and Exhaust System includes two subsystems, Containment Purge and Containment Inservice Purge. The containment purgenormal subsystem includes a 364-2 inch purge penetration and a 3642- inch 1 IR-12 exhaust penetration. The second subsystem, a minipurge (conti nued)

WOG STS Rev 1, 04/07/95 B 3.9.4-2 Markup for PI ITS Part E

Containment Penetrations B 3.9.4 BASES system referred to as containment inservice purge, includes af 148 inch purge penetration and an 18 inch exhaust penetration. JCL3.9_102 During MODES 1, 2, 3, and 4, the two valves in each of the containmentnerma+ purge and exhaust penetrations are secured in the closed positionblank flanged. The two valves in each of the two containment inservice purge mi-rpturge penetrations can be opened intermittently, but are closed automatically by the Containment Ventilation Isolation System. Engineerd S-.fety Features.CL39-102 Actuation System^(ESFAS). Npther of the subsystm.

iS subjeect to a Specification inMODE 5.

In MODE 6, sufficient air flow rates large air exchangers, are necessary to conduct refueling operations. The n-efm&

42 ineh inservice purge system is used for this purpose, and -l-l each of the four valves iseare closed JCL3.9-102 by the radiation monitors associated with the containment inservice purge system in accordance with, LCO 3.3.5, "Containment Ventilation Isolation IPA3.9-11 Instrumentation." [ESAS in a.ccrdance with LCO 3.3.2, "Engineered Safety Feature Actuation System ([SEAS)

Instrumentation." The 36 inch subsystem is normally blank flanged, although the option for use is allowed during outages, except during movement of irradiated fuel or CORE ALTERATIONS. All four containment purge valves are also closed by the Containment Ventilation Isolation System.

TI I I

, efeains o ,rationa, R-12 The minipurge system* L------- I also close d by 1-*,,

in M O*E 6, and all four valves are

[SAS. or The,, 4ipurge system,, is not used in MODE 6.

All four 8 inch valves arc secured in the closed position.

The other containment penetrations that provide direct access from containment atmosphere to outside atmosphere must be isolated on at least one (continued)

WOG STS Rev 1, 04/07/95 B 3.9.4-3 Markup for PI ITS Part E

Containment Penetrations B 3.9.4 BASES

a. The equipment hatch is closed and held in place by at least 4 bolts;

b. One door in each air lock is closed, or both doors in each air lock may be open with: I CL3.9-62

1. containment (high flow) purge system isolated,

2. one air lock door OPERABLE and capable of being closed within 30 minutes, and

3. at least two containment fan coil unit fans capable of operating in the high speed mode; and

c. At least one isolation valve in each penetration, including the containment (high flow) purge system and inservice (low flow) purge system, providing direct access from the containment atmosphere to the outside atmosphere is either:

1. OPERABLE or closed by a manual valve, or blind flange, or

2. capable of being closed by an OPERABLE I PA3.9-64 1 Containment Ventilation Isolation System.

A penetration with direct access from the containment atmosphere to the outside atmosphere includes all penetrations that have a flow path that leads anywhere outside containment. I CL3.9-62 I -----------------

__ _ _ __ _ _ i I- ...- L..

II 1R-10 II (conti nued)

WOG STS Rev 1, 04/07/95 B 3.9.4-6 Markup for PI ITS Part E

Containment Penetrations B 3.9.4 BASES The containment air lock doors may be open cL3"9-62 during movement of irradiated fuel in the containment provided that the LCO requirements are met. These requirements include one door OPERABLE, under procedural control and capable of being closed within 30 minutes following a fuel handling accident in containment and at least two fan coil unit fans are capable of operating in the high speed mode PA3.9-64 following a fuel handling accident in r ,

containment. Should a *~fuel handling accident occur :R-10 1 inside containment, the fan coil unit fans will be L operated in high speed and one door in each air lock will be closed following an evacuation of containment.

For the OPERABLE containment purge and exhaust r penetrations, this LCO ensures that these penetrations R-12 1 Isolation L ------- I are isolable by the Containment Ventilation System. Containmen.. urge and. Exhaust isolation System.

The OPERABILITY requirements for this LCO requireensure that the automatic purge and exhaust valve closure ti-mes specified in the FSAR can be achieved and, therefore, meet the assumptions used in the safety analysis to ensure that R-12 such that L I releases through the valves are terminated, radiological doses are within the acceptance limit.

APPLICABILITY The containment penetration requirements are applicable during COREAL*+/-ERAT+/-IONSor-movement of irradiated fuel assemblies within containment because TA3.9-66 this is when there is a potential for the limiting fuel handling accident.

In MODES 1, 2, 3, and 4, containment penetration requirements are addressed by LCO 3.6.1.

(continued)

WOG STS Rev 1, 04/07/95 B 3.9.4-7 Markup for PI ITS Part E

Containment Penetrations B 3.9.4 BASES In MODES 5 and 6, when CORE ALTERATIONS or movement of irradiated fuel assemblies within containment is a-re-not being conducted, the potential for a fuel handling accident does not exist. Therefore, under these conditions no requirements are placed on containment penetration status. PA3.9-64 ACTIONS A. .

A If the containment equipment hatch, air locks, or any containment penetration that provides direct access from the containment atmosphere to the outside atmosphere is not in the required status, including the Containment Purge-and Exhaust Ventilation Isolation System not capable of automatic actuation when the purge and exhaust valves are open, the unit must be placed in a condition where R-12 the isolation function is not needed. This is R accomplished by immediately suspending CORE ALTERATIGNS eftd-movement of irradiated fuel assemblies within containment. Performance of these actions shall not preclude completion of movement of a fuel assemblye-emponent to a safe position.

SURVEILLANCE SR 3.9.4.1 I PA3.9-64 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 I TA3.9-66 open purge and exhaust valves will demonstrate that the valves will function if required during a fuel handling accidentare not blo*ked froGm clsing. Also the Surveillance will 1R-12 I iL .

II..

(conti nued)

WOG STS Rev 1, 04/07/95 B 3.9.4-8 Markup for PI ITS Part E

Containment Penetrations B 3.9.4 BASES SURVEILLANCE SR~ 3.9.4.1. (continued)

REQUIREMENTS demonstrate that each valve operator has motive power, whi ch will ensure that each valve is capable of being closed by an OPERABLE automatic Ceontainment Ventilationpurge-and ehaust Il+solation signal.

TA3.9-66 The Surveillance is performed every 7 days during PA3. .9-64 CORE-ALTERATINS or-movement of irradiated fuel assemblies within I 9-66 containment. The Surveillance interval is selected TA3.

to be commensurate with the normal duration of time to complete fuel handling operations. A surveillance is to be conducted before the start of refueling operation and then in accordance with the frequency specifiedw-pill ,rovi, two or three surveil .an.e " " during the "fiation applicable period for this LCO. As such, this Surveillance ensures that a postulated fuel handling accident that releases fission product radioactivity within the Tf*3.9-66 j containment will not result in a release of significant fission product radioactivity to the 39-05 environment.

SR 3.9.4.2 This Surveillance demonstrates that each containment purge and exhaust valve actuates to its isolation position on ---

manual initiation or on an actual or simulated high R-12" radiation signal. The 24-18 month Frequency maintains L.------- I consistency with other similar ESFAS instrumentation and valve testing requirements. In LCO 3.3.56, the Containment Ventilation Purge and Exhaust Isolation instrumentation requires a CHANNEL CHECK every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and a COT every 92 days to ensure the channel X3.9-61 OPERABILITY during refueling operations. Every 241-8 months a CHANNEL CALIBRATION is (continued)

WOG STS Rev 1, 04/07/95 B 3.9.4-9 Markup for PI ITS Part E

Part F Package 3.9 Part FPakg39 Difference Difference Category Number Justification for Differences 3.9-63 Not used.

PA 64 The PI name for the instrumentation system which automatically isolates containment ventilation during fuel handling is the "Containment Ventilation Isolation System" and the Specification for this system is 3.3.5, "Containment Ventilation Isolation Instrumentation". The systems which are isolated are the "containment purge (high flow) system" and containment inservice (low flow) purge system", thus these names are used in SR 3.9.4.2 and throughout the Bases as applicable. The parenthetical modifiers

" (high flow)" and "(low flow)" may be included to assure that the operators do not confuse these systems.

CL 65 The "or equivalent" option in NUREG-1431 is not included in the PI ITS. The Specification and Bases have been revised. PI does not currently have this flexibility and the evaluations which support it have not been performed, thus this is not included.

Prairie Island Units 1 and 2 5 4/1/02

Current Technical Specification Cross-Reference CTS Section CTS Table Section Type ITS Section ITS Table Item Number Item Number Table 4.1-1C 11 SR 3.3.4.1 Table 4.1-1C 11 SR 3.3.4.2 Table 4.1-1C 12 Deleted - Boric Acid LAR Table 4.1-1C 13 Relocated TRM Table 4.1-1C 14 CTS Deleted Table 4.1-1C 15 TABLE 3.3.1-1 16.b.2 Table 4.1-1C 15 Relocated TRM Table 4.1-1C 16 Relocated TRM Table 4.1-1C 17 Relocated TRM Table 4.1-1C 18 SR 3.3.1.12 Table 4.1-1C 19 Relocated TRM Table 4.1-1C 20 Relocated TRM Table 4.1-1C 21 SR 3.3.3.1 Table 4.1-1C 21 SR 3.3.3.2 Table 4.1-1C 21 SR 3.3.3.3 Table 4.1-1C 22 CTS Deleted Table 4.1-1C 23 CTS Deleted Table 4.1-1C 24 Relocated TRM Prairie Island Units 1 and 2 Table - 18 4/1/02

Current Technical Specification Cross-Reference CTS Table CTS Section Section Type ITS Section ITS Table Item Number Item Number Table 4.1-1C 24 SR 3.3.6.5 Table 4.1-1C 24 SR 3.3.6.2 Table 4.1-1C 25 SR 3.4.12.4 Table 4.1-1C 25 SR 3.4.12.5 Table 4.1-1C 25 SR 3.4.13.5 Table 4.1-1C 25 SR 3.4.13.6 Table 4.1-1C 26 Relocated TRM Table 4.1-1C 27 Relocated TRM Table 4.1-1C 28 Relocated TRM Table 4.1-1C 29 SR 3.3.3.1 Table 4.1-4C 29 SR 3.3.3.2 Table 4.1-1iC 29 (Partial) Relocated TRM Table 4.1-1C 30 Relocated Bases Table 4.1-1C 31 Relocated TRM Table 4.1-1C Note 30 SR 3.1.7.1 Table 4.1-1C Note 31 Deleted Table 4.1-1C Note 32 Relocated TRM Prairie Island Units 1 and 2 Table - 19 4/1/02

Programs and Manuals 5.5 5.5 Programs and Manuals (continued) 5.5.8 Steam Generator (SG) Tube Surveillance Program Steam generator tubes in each unit shall be determined operable by the following:

a. Steam Generator Sample Selection and Inspection Each steam generator shall be determined operable in accordance with the in-service inspection schedule in Specification 5.5.8.c. The in-service inspection may be limited to one steam generator on a rotating schedule encompassing 6% of the tubes in the single steam generator, provided the previous inspections indicated that the two steam generators are performing in a like manner.

b. Steam Generator Tube Sample Selection and Inspection The steam generator tube minimum sample size, inspection result classification, and the corresponding action required shall be as specified in Tables 5.5.8-1 and 5.5.8-2. The in-service inspection of steam generator tubes shall be performed at the frequencies specified in Specification 5.5.8.c and the inspected tubes shall be verified acceptable per the acceptance criteria of Specification 5.5.8.d. The tubes selected for each in-service inspection shall include at least 3% of the total number of tubes in all steam generators and at least 20% of the total number of sleeves in service in both steam generators; the tubes selected for these inspections shall be selected on a random basis except:

1. Where experience in similar plants with similar water chemistry indicates critical areas to be inspected, then at least 50% of the tubes inspected shall be from these critical areas.

2. The first sample of tubes selected for each in-service inspection (subsequent to the preservice inspection) of each steam generator shall include:

Prairie Island Units 1 and 2 5.0-13 4/1/02

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.8 Steam Generator (SG) Tube Surveillance Program (continued)

(a) All tubes that previously had detectable wall penetrations

(>20%) that have not been plugged or sleeve repaired in the affected area.

(b) Tubes in those areas where experience has indicated potential problems.

(c) A tube inspection (pursuant to Specification 5.5.8.d. .(h)) shall be performed on each selected tube. If any selected tube does not permit the passage of the eddy current probe for a tube inspection, this shall be recorded and an adjacent tube shall be selected and subjected to a tube inspection.

3. In addition to the sample required in Specification 5.5.8.b.2(a) through (c), all tubes which have had the F* or EF* criteria applied will be inspected in the F*

and EF* regions of the roll expanded region. The region of these tubes below the F* and EF* regions may be excluded from the requirements of Specification 5.5.8.b.2(a).

4. The tubes selected as the second and third samples (if required by Tables 5.5.8-1 or 5.5.8-2) during each in-service inspection may be subjected to a partial tube or sleeve inspection provided:

(a) The tubes selected for these samples include the tubes from those areas of the tube sheet array where tubes with imperfections were previously found.

(b) The inspections include those portions of the tubes or sleeves where imperfections were previously found.

Prairie Island Units 1 and 2 5.0-14 4/1/02

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.8 Steam Generator (SG) Tube Surveillance Program (continued)

The results of each sample inspection shall be classified into one of the following three categories:

Category Inspection Results C-1 Less than 5% of the total tubes inspected are degraded tubes and none of the inspected tubes are defective.

C-2 One or more tubes, but not more than 1% of the total tubes inspected are defective, or between 5%

and 10% of the total tubes inspected are degraded tubes.

C-3 More than 10% of the total tubes inspected are degraded tubes or more than 1% of the inspected tubes are defective.

Note: In all inspections, previously degraded tubes must exhibit significant (>10%) further wall penetrations to be included in the above percentage calculations.

5. Indications left in service as a result of application of tube support plate voltage-based repair criteria shall be inspected by bobbin coil probe during all future refueling outages.

6. Implementation of the steam generator tube/tube support plate repair criteria requires a 100 percent bobbin coil inspection for hot leg and cold leg tube support plate intersections down to the lowest cold leg tube support plate with known outside diameter stress corrosion cracking (ODSCC) indications. The determination of the lowest cold leg tube support plate intersections having ODSCC indications shall be based on the performance of at least a 20 percent random sampling of tubes inspected over their full length.

Prairie Island Units 1 and 2 5.0-15 4/1/02

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.8 Steam Generator (SG) Tube Surveillance Program (continued)

c. Inspection Frequencies The above required in-service inspections of steam generator tubes shall be performed at the following frequencies:

1. In-service inspections shall be performed at intervals of not less than 12 nor more than 24 calendar months after the previous inspection. If two consecutive inspections following service under AVT conditions, not including the preservice inspection, result in all inspection results falling into the C-I category or if two consecutive inspections demonstrate that previously observed degradation has not continued and no additional degradation has occurred, the inspection interval may be extended to a maximum of once per 40 months.

2. If the results of the in-service inspection of a steam generator conducted in accordance with Table 5.5.8-1 at 40 month intervals fall in Category C-3, the inspection frequency shall be increased to at least once per 20 months. The increase in inspection frequency shall apply until the subsequent inspections satisfy the criteria of Specification 5.5.8.c. 1; the interval may then be extended to a maximum of once per 40 months.

3. Additional, unscheduled in-service inspections shall be performed on each steam generator in accordance with the first sample inspection specified in Table 5.5.8-1 during the shutdown subsequent to any of the following conditions:

(a) Primary-to-secondary tube leaks (not including leaks originating from tube-to-tube sheet welds) in excess of the limits of Specification 3.4.14.

(b) A seismic occurrence greater than the Operating Basis Earthquake.

Prairie Island Units 1 and 2 5.0-16 4/1/02

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.8 Steam Generator (SG) Tube Surveillance Program (continued)

(c) A loss-of-coolant accident requiring actuation of the engineered safeguards.

(d) A main steam line or feedwater line break.

d. Acceptance Criteria

1. As used in this Specification:

(a) Imperfection means an exception to the dimensions, finish or contour of a tube from that required by fabrication drawings or specifications. Eddy-current testing indications below 20% of the nominal tube wall thickness, if detectable, may be considered as imperfections.

(b) Degradation means a service-induced cracking, wastage, wear or general corrosion occurring on either inside or outside of a tube.

(c) Degraded Tube means a tube containing imperfections

> 20% of the nominal wall thickness caused by degradation.

(d) % Degradation means the percentage of the tube wall thickness affected or removed by degradation.

(e) Defect means an imperfection of such severity that it exceeds the repair limit. A tube containing a defect is defective.

Prairie Island Units 1 and 2 5.0-17 4/1/02

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.8 Steam Generator (SG) Tube Surveillance Program (continued)

(f) Repair Limit means the imperfection depth at or beyond which the tube shall be removed from service by plugging or repaired by sleeving because it may become unserviceable prior to the next inspection and is equal to 50% of the nominal tube wall thickness. If significant general tube thinning occurs, this criteria will be reduced to 40% wall penetration. This definition does not apply to the portion of the tube in the tubesheet below the F* distance provided the tube is not degraded (i.e., no indications of cracks) within the F* or EF* distance for F* or EF* tubes. The repair limit for the pressure boundary region of any sleeve is 25% of the nominal sleeve wall thickness. This definition does not apply to tube support plate intersections for which the voltage-based repair criteria are being applied. Refer to Specification 5.5.8.d.4 for the repair limit applicable to these intersections.

(g) Unserviceable describes the condition of a tube if it leaks or contains a defect large enough to affect its structural integrity in the event of an Operating Basis Earthquake, a loss-of-coolant accident, or a steam line or feedwater line break.

(h) Tube Inspection means an inspection of the steam generator tube from the point of entry (hot leg side) completely around the U-bend to the top support of the cold leg.

(i) Sleeving is the repair of degraded tube regions using a new Alloy 690 tubing sleeve inserted inside the parent tube and sealed at each end by welding or by replacing the lower weld in a full depth tubesheet sleeve with a hard rolled joint. The new sleeve becomes the pressure boundary spanning the original degraded tube region.

Prairie Island Units 1 and 2 5.0-18 4/1/02

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.8 Steam Generator (SG) Tube Surveillance Program (continued) j) F* Distance is the distance from the bottom of the hardroll transition toward the bottom of the tubesheet that has been conservatively determined to be 1.07 inches (not including eddy current uncertainty). The F* distance applies to roll expanded regions below the midplane of the tubesheet.

(k) F* Tube is a tube with degradation, below the F* distance, equal to or greater than 40%, and not degraded (i.e., no indications of cracking) within the F* distance.

(1) EF* Distance is the distance from the bottom of the upper hardroll transition toward the bottom of the tubesheet that has been conservatively determined to be 1.67 inches (not including eddy current uncertainty). EF* distance applies to roll expanded regions when the top of the additional roll expansion is 2.0 inches or greater down from the top of the tubesheet.

(m) EF* Tube is a tube with degradation, below the EF* distance, equal to or greater than 40%, and not degraded (i.e., no indications of cracking) within the EF* distance.

2. The steam generator shall be determined OPERABLE after completing the corresponding actions (plug or repair by sleeving all tubes exceeding the repair limit and all tubes containing through-wall cracks or classify as F* or EF* tubes) required by Tables 5.5.8-1 and 5.5.8-2.

3. Tube repair, after April 1, 1999, using Combustion Engineering welded sleeves shall be in accordance with the methods described in the following:

CEN-629-P, Revision 03-P, "Repair of Westinghouse Series 44 and 51 Steam Generator Tubes Using Leak Tight Sleeves".

Prairie Island Units 1 and 2 5.0-19 4/1/02

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.8 Steam Generator (SG) Tube Surveillance Program (continued)

4. Tube Support Plate Repair Limit is used for the disposition of a steam generator tube for continued service that is experiencing predominantly axially oriented outside diameter stress corrosion cracking confined within the thickness of the tube support plates. At tube support plate intersections, the repair limit is based on maintaining steam generator serviceability as described below:

(a) Steam generator tubes, whose degradation is attributed to outside diameter stress corrosion cracking within the bounds of the tube support plate with bobbin voltages less than or equal to 2.0 volts will be allowed to remain in service.

(b) Steam generator tubes, whose degradation is attributed to outside diameter stress corrosion cracking within the bounds of the tube support plate with a bobbin voltage greater than 2.0 volts, will be repaired or plugged, except as noted in Specification 5.5.8.d.4(c) below.

(c) Steam generator tubes, with indications of potential degradation attributed to outside diameter stress corrosion cracking within the bounds of the tube support plate with a bobbin voltage greater than 2.0 volts but less than or equal to the upper voltage repair limit, may remain in service if a rotating pancake coil (or comparable examination technique) inspection does not detect degradation. Steam generator tubes, with indications of outside diameter stress corrosion cracking degradation with a bobbin voltage greater than the upper voltage repair limit will be plugged or repaired.

Prairie Island Units 1 and 2 5.0-20 4/1/02

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.8 Steam Generator (SG) Tube Surveillance Program (continued)

(d) If an unscheduled mid-cycle inspection is performed, the following mid-cycle repair limits apply instead of the limits in Specifications 5.5.8.d.4(a), (b) and (c). The mid-cycle repair limits are determined from the following equations:

VSL VMURL-

-A 1.0 + NDE + Gr CL 2

VMLRL=VMURL-(VURL- .0) ( CL-Atj Where:

VURL = upper voltage repair limit VLRL = lower voltage repair limit VMURL = mid-cycle upper voltage repair limit based on time into cycle VMLPL = mid-cycle lower voltage repair limit based on VMUpL and time into cycle At = length of time since last scheduled inspection during which VuR1 and VLRL were implemented Prairie Island Units 1 and 2 5.0-21 4/1/02

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.8 Steam Generator (SG) Tube Surveillance Program (continued)

CL = cycle length (time between two scheduled steam generator inspections)

VSL = structural limit voltage Gr = average growth rate per cycle length NDE = 95 percent cumulative probability allowance for nondestructive examination uncertainty (i.e., a value of 20 percent has been approved by the NRC)

Implementation of these mid-cycle repair limits should follow the same approach as described in Specifications 5.5.8.d.4(a),

(b) and (c).

Note: The upper voltage repair limit is calculated according to the methodology in Generic Letter 95-05 as supplemented.

Prairie Island Units 1 and 2 5.0-22 4/1/02

Programs and Manuals 5.5 5.5 Programs and Manuals (continued) 5.5.9 Ventilation Filter Testing Program (VFTP)

A program shall be established to implement the following required testing of Engineered Safety Feature (ESF) filter ventilation systems and the Spent Fuel Pool Special and Inservice Purge Ventilation System each operating cycle (18 months for shared systems).

Demonstrate for the Auxiliary Building Special Ventilation, Shield Building Ventilation, Control Room Special Ventilation, and Spent Fuel Pool Special and Inservice Purge Ventilation Systems that:

a. An inplace DOP test of the high efficiency particulate air (HEPA) filters shows a penetration and system bypass < 1% (for DOP, particles having a mean diameter of 0.7 microns);

b. A halogenated hydrocarbon test of the inplace charcoal adsorber shows a penetration and system bypass < 1% (for DOP, particles having a mean diameter of 0.7 microns);

c. A laboratory test of a sample of the charcoal adsorber, when obtained as described in Regulatory Guide 1.52, Revision 2, shows the methyl iodide penetration less than 15% penetration (less than 5%

penetration for the Control Room Special Ventilation System) when tested in accordance with ASTM D3803-1989 at a temperature of 30'C and 95% relative humidity (RH) (or 70% RH with humidity controls if the humidity controls are capable of maintaining the humidity of the air entering the charcoal less than or equal to 70% RH under worst-case design-basis conditions); and

d. The pressure drop across the combined HEPA filters and the charcoal adsorbers is less than 6 inches of water at the system flowrate +/- 10%.

The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the VFTP test frequencies.

Prairie Island Units 1 and 2 5.0-23 11/1/01

Programs and Manuals 5.5 5.5 Programs and Manuals (continued) 5.5.10 Explosive Gas and Storage Tank Radioactivity Monitoring Program This program provides controls for potentially explosive gas mixtures contained in the waste gas holdup system, the quantity of radioactivity contained in gas storage tanks, and the quantity of radioactivity contained in unprotected outdoor liquid storage tanks.

The program shall include:

a. The limits for concentrations of oxygen in the waste gas holdup system and a surveillance program to ensure the limits are maintained. Such limits shall be appropriate to the system's design criteria;

b. A surveillance program to ensure that the quantity of radioactivity contained in each gas storage tank is less than or equal to 78,800 curies of noble gas (considered as dose equivalent Xe-133); and

c. A surveillance program to ensure that the quantity of radioactivity contained in each of the following tanks shall be limited to 10 curies, excluding tritium and dissolved or entrained noble gases:

Condensate storage tanks Outside temporary tanks The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Explosive Gas and Storage Tank Radioactivity Monitoring Program surveillance frequencies.

Prairie Island Units l and 2 5.0-24 12/11/00

Programs and Manuals 5.5 5.5 Programs and Manuals (continued) 5.5.11 Diesel Fuel Oil Testing Program A diesel fuel oil testing program to implement required testing of both new fuel oil and stored fuel oil shall be established. The program shall include sampling and testing requirements, and acceptance criteria, all in accordance with the limits specified in Table 1 of ASTM D975-77 when checked for viscosity, water, and sediment. Acceptability of new fuel oil shall be determined prior to addition to the safeguards storage tanks.

Testing of diesel fuel oil stored in the safeguards storage tanks shall be performed at least every 31 days.

The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Diesel Fuel Oil Testing Program test frequencies.

5.5.12 Technical Specifications (TS) Bases Control Program This program provides a means for processing changes to the Bases of these Technical Specifications.

a. Changes to the Bases of the TS shall be made under appropriate administrative controls and reviews.

b. Licensees may make changes to Bases without prior NRC approval provided the changes do not require either of the following:

1. a change in the TS incorporated in the license; or

2. a change to the USAR or Bases that requires NRC approval pursuant to 10 CFR 50.59.

c. The Bases Control Program shall contain provisions to ensure that the Bases are maintained consistent with the USAR.

Prairie Island Units 1 and 2 5.0-25 5/1/01

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.12 Technical Specifications (TS) Bases Control Program (continued)

d. Proposed changes that meet the criteria of Specification 5.5.12 b above shall be reviewed and approved by the NRC prior to implementation.

Changes to the Bases implemented without prior NRC approval shall be provided to the NRC on a frequency consistent with USAR updates.

5.5.13 Safety Function Determination Program (SFDP)

This program ensures loss of safety function is detected and appropriate actions taken. Upon entry into LCO 3.0.6, an evaluation shall be made to determine if loss of safety function exists. Additionally, other appropriate actions may be taken as a result of the support system inoperability and corresponding exception to entering supported system Conditions and Required Actions. This program implements the requirements of LCO 3.0.6. The SFDP shall contain the following:

a. Provisions for cross train checks to ensure a loss of the capability to perform the safety function assumed in the accident analysis does not go undetected;

b. Provisions for ensuring the plant is maintained in a safe condition if a loss of function condition exists;

c. Provisions to ensure that an inoperable supported system's Completion Time is not inappropriately extended as a result of multiple support system inoperabilities; and

d. Other appropriate limitations and remedial or compensatory actions.

Prairie Island Units 1 and 2 5.0-26 12/11/00

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.13 Safety Function Determination Program (SFDP) (continued)

A loss of safety function exists when, assuming no concurrent single failure, no concurrent loss of offsite power or no concurrent loss of onsite diesel generator(s), a safety function assumed in the accident analysis cannot be performed. For the purpose of this program, a loss of safety function may exist when a support system is inoperable, and:

a. A required system redundant to the system(s) supported by the inoperable support system is also inoperable; or

b. A required system redundant to the system(s) in turn supported by the inoperable supported system is also inoperable; or

c. A required system redundant to the inoperable support system(s) for the supported systems (a) and (b) above is also inoperable.

The SFDP identifies where a loss of safety function exists. If a loss of safety function is determined to exist by this program, the appropriate Conditions and Required Actions of the LCO in which the loss of safety function exists are required to be entered. When a loss of safety function is caused by the inoperability of a single Technical Specification support system, the appropriate Conditions and Required Actions to enter are those of the support system.

5.5.14 Containment Leakage Rate Testing Program

a. A program shall be established to implement the leakage rate testing of the containment as required by 10 CFR 50.54(o) and 10 CFR 50, Appendix J, Option B, as modified by approved exemptions. This program shall be in accordance with the guidelines contained in Regulatory Guide 1.163, "Performance-Based Containment Leak-Test Program," dated September 1995.

Prairie Island Units 1 and 2 5.0-27 5/1/01

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.14 Containment Leakage Rate Testing Program (continued)

b. The peak calculated containment internal pressure for the design basis loss of coolant accident is less than the containment internal design pressure, Pa, of 46 psig.

c. The maximum allowable primary containment leakage rate, La, at Pa, shall be 0.25% of primary containment air weight per day. For pipes connected to systems that are in the auxiliary building special ventilation zone, the total leakage shall be less than 0.1% of primary containment air weight per day at pressure Pa. For pipes connected to systems that are exterior to both the shield building and the auxiliary building special ventilation zone, the total leakage past isolation valves shall be less than 0.0 1% of primary containment air weight per day at pressure Pa

d. Leakage Rate acceptance criteria are:

1. Primary containment leakage rate acceptance criterion is *< 1.0 La.

Prior to unit startup, following testing in accordance with the program, the combined leakage rate acceptance criteria are *< 0.60 La for all components subject to Type B and Type C tests and

< 0.75 La for Type A tests.

2. Air lock testing acceptance criteria are:

a) Overall air lock leakage rate is

0.05 La when tested at >_46 psig.

b) For each door intergasket test, leakage rate is _<0.01 L, when pressurized to _ 10 psig.

Prairie Island Units 1 and 2 5.0-28 12/11/00

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.14 Containment Leakage Rate Testing Program (continued)

e. The provisions of SR 3.0.3 are applicable to the Containment Leakage Rate Testing Program.

f. Nothing in these Technical Specifications shall be construed to modify the testing Frequencies required by 10 CFR 50, Appendix J.

5.5.15 Battery Monitoring and Maintenance Program This Program provides for restoration and maintenance of the 125V plant safeguards batteries and service building batteries, which may be used instead of the safeguards batteries during shutdown conditions in accordance with manufacturer's recommendations, as follows:

a. Actions to restore battery cells with float voltage < 2.13 V will be in accordance with manufacturer's recommendations, and

b. Actions to equalize and test battery cells that had been discovered with electrolyte level below the minimum established design limit.

Prairie Island Units 1 and 2 5.0-29 5/l/01

Programs and ivManuals 5.5 Table 5.5.8-1 STEAM GENERATOR TUBE INSPECTION 1st SAMPLE INSPECTION 2 nd SAMPLE INSPECTION 3 rd SAMPLE INSPECTION Sample Size Result Action Required Result Action Required Result Action Required A minimum C- I None N/A N/A N/A N/A of S Tubes C-2 Repair defective tubes C-i None N/A N/A per S.G. and inspect additional 2S Repair defective C-1 None tubesC-2 tubes and inspect C-2 Repair defective additional 4S tubes tubes in this S.G.

C-3 Perform action for C-3 result of first sample C-3 Perform action for N/A N/A C-3 result of first sample All other S.G.s are None N/A N/A C-3 Inspect all tubes in this S.G., Repair defective C-1 tubes and inspect 2S Some S.G.s C-2 but Perform action for N/A N/A tubes in each other S.G. no additional S.G. C-2 result of second Prompt notification to are C-3 sample NRC Additional S.G. is Inspect all tubes in N/A N/A C-3 each S.G. and repair defective tubes.

Prompt notification to NRC.

S=3%; When two steam generators are inspected during that outage.

S=6%; When one steam generator is inspected during that outage.

Prairie Island Units 1 and 2 5.0-30 4/1/02

Programs and ivianuals 5.5 Table 5.5.8-2 STEAM GENERATOR TUBE SLEEVE INSPECTION 1 " Sample Inspection 2nd Sample Inspection Sample Size Result Action Required Result Action Required A minimum C- I None N/A N/A of 20% of Tb SleevesC-2 Inspect all remaining tube C-i None (1) sleeves in this S.G. and plug or C-2 Plug or repair defective sleeved repair defective sleeved tubes. tubes C-3 Perform action for C-3 result of first sample C-3 Inspect all tube sleeves in this The other S.G. None S.G., inspect 20% of the tube is C-1 sleeves in the other S.G., and The other S.G. Perform action for C-2 result of first plug or repair defective sleeved is C-2 sample tubes The other S.G. Inspect all tube sleeves in each S.G.

is C-3 and plug or repair defective sleeved tubes (1) Each type of sleeve is considered a separate population for determination of scope expansion Prairie Island Units 1 and 2 5.0-31 4/1/02

Reporting Requirements 5.6 5.0 ADMINISTRATIVE CONTROLS 5.6 Reporting Requirements The following reports shall be submitted in accordance with 10 CFR 50.4.

5.6.1 Occupational Exposure Report

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

A single submittal may be made for the plant. The submittal should combine sections common to both units.

A tabulation on an annual basis of the number of station, utility, and other personnel (including contractors) for whom monitoring was performed, receiving an annual deep dose equivalent > 100 mrem and the associated collective deep dose equivalent (reported in person-rem) according to work and job functions, e.g., reactor operations and surveillance, inservice inspection, routine maintenance, special maintenance (describe maintenance), waste processing, and refueling. This tabulation supplements the requirements of 10 CFR 20.2206. The dose assignments to various duty functions may be estimated based on pocket ionization chamber, thermoluminescent dosimeter (TLD), electronic dosimeter, or film badge measurements. Small exposures totaling < 20% of the individual total dose need not be accounted for. In the aggregate, at least 80% of the total deep dose equivalent received from external sources should be assigned to specific major work functions. The report covering the previous calendar year shall be submitted by April 30 of each year.

Prairie Island Units 1 and 2 5.0-32 12/11/00

Reporting Requirements 5.6 5.6 Reporting Requirements (continued) 5.6.2 Annual Radiological Environmental Monitoring Report

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

A single submittal may be made for the plant. The submittal should combine sections common to both units.

The Annual Radiological Environmental Monitoring Report covering the operation of the plant during the previous calendar year shall be submitted by May 15 of each year. The report shall include summaries, interpretations, and analyses of trends of the results of the radiological environmental monitoring program for the reporting period. The material provided shall be consistent with the objectives outlined in the Offsite Dose Calculation Manual (ODCM), and in 10 CFR 50, Appendix I, Sections IV.B.2, IV.B.3, and IV.C.

The Annual Radiological Environmental Monitoring Report shall include summarized and tabulated results, in the format of Regulatory Guide 4.8, December 1975, of all radiological environmental samples taken during the report period. In the event that some individual results are not available for inclusion with the report, the report shall be submitted noting and explaining the reasons for the missing results. The missing data shall be submitted in a supplementary report as soon as possible.

The report shall also include the following: a summary description of the radiological environmental monitoring program; a map of sampling locations keyed to a table giving distances and directions from the reactor site; and the results of licensees participation in the Interlaboratory Comparison Program defined in the ODCM.

Prairie Island Units l and 2 5.0-33 12/11/00

Reporting Requirements 5.6 5.6 Reporting Requirements (continued) 5.6.3 Radioactive Effluent Report

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

A single submittal may be made for the plant. The submittal shall combine sections common to both units.

The Radioactive Effluent Report covering the operation of the plant during the previous calendar year shall be submitted by May 15 of each year. The report shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the plant. The material provided shall be consistent with the objectives outlined in the ODCM and in conformance with 10 CFR 50.36a and 10 CFR 50, Appendix I, Section IV.B. 1.

5.6.4 Monthly Operating Reports Routine reports of operating statistics and shutdown experience shall be submitted on a monthly basis no later than the 15th of each month following the calendar month covered by the report.

5.6.5 CORE OPERATING LIMITS REPORT (COLR)

a. Core operating limits shall be established prior to each reload cycle, or prior to any remaining portion of a reload cycle, and shall be documented in the COLR for the following:

LCO 3.1.1, "SHUTDOWN MARGIN (SDM)";

LCO 3.1.3, "Isothermal Temperature Coefficient (ITC)";

LCO 3.1.5, "Shutdown Bank Insertion Limits";

LCO 3.1.6, "Control Bank Insertion Limits";

LCO 3.1.8, "PHYSICS TESTS Exceptions - MODE 2";

Prairie Island Units 1 and 2 5.0-34 12/11/00

Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued)

LCO 3.2.1, "Heat Flux Hot Channel Factor (FQ(Z))";

LCO 3.2.2, "Nuclear Enthalpy Rise Hot Channel Factor (FZ4);

LCO 3.2.3, "AXIAL FLUX DIFFERENCE (AFD)";

LCO 3.4.1, "RCS Pressure, Temperature, and Flow - Departure from Nucleate Boiling (DNB) Limits"; and LCO 3.9.1, "Boron Concentration".

b. The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:

1. NSPNAD-8 10 1-PA, "Qualification of Reactor Physics Methods for Application to PI Units" (latest approved version);

2. NSPNAD-8102-PA, "Prairie Island Nuclear Power Plant Reload Safety Evaluation Methods for Application to PI Units"(latest approved version);

3. NSPNAD-97002-PA, "Northern States Power Company's "Steam Line Break Methodology", (latest approved version);

4. WCAP-9272-P-A, "Westinghouse Reload Safety Evaluation Methodology", July, 1985;

5. WCAP-10054-P-A, "Westinghouse Small Break ECCS Evaluation Model using the NOTRUMP Code", August, 1985;

6. WCAP-10924-P-A, "Westinghouse Large Break LOCA Best-Estimate Methodology", December, 1988;

7. WCAP-10924-P-A, Volume 1, Addendum 4, "Westinghouse Large Break LOCA Best Estimate Methodology", August, 1990; Prairie Island Units 1 and 2 5.0-35 12/11/00

Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued)

8. XN-NF-77-57 (A), XN-NF-77-57, Supplement 1 (A), "Exxon Nuclear Power Distribution Control for Pressurized Water Reactors Phase II", May, 1981;

9. WCAP- 13677, "10 CFR 50.46 Evaluation Model Report:

W-COBRA/TRAC 2-Loop Upper Plenum Injection Model Update to Support ZIRLOTM Cladding Options", April 1993 (approved by NRC SE dated November 26, 1993);

10. NSPNAD-93003-A, "Transient Power Distribution Methodology",

(latest approved version).

c. The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal-mechanical limits, core thermal-hydraulic limits, Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as SDM, transient analysis limits, and accident analysis limits) of the safety analysis are met.

d. The COLR, including any midcycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.

5.6.6 Reactor Coolant System (RCS) PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)

a. RCS pressure and temperature limits for heat-up, cooldown, low temperature operation, criticality, and hydrostatic testing, LTOP arming, PORV lift settings and Safety Injection Pump Disable Temperature as well as heatup and cooldown rates shall be established and documented in the PTLR for the following:

LCO 3.4.3, "RCS Pressure and Temperature (P/T) Limits";

LCO 3.4.6, "RCS Loops - MODE 4";

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

LCO 3.4.10, "Pressurizer Safety Valves";

Prairie Island Units 1 and 2 5.0-36 12/11/00

Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.6 Reactor Coolant System (RCS) PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR) (continued)

LCO 3.4.12, "Low Temperature Overpressure Protection (LTOP)

Reactor Coolant System Cold Leg Temperature (RCSCLT) > Safety Injection (SI) Pump Disable Temperature";

LCO 3.4.13, "Low Temperature Overpressure Protection (LTOP)

Reactor Coolant System Cold Leg Temperature (RCSCLT) _<Safety Injection (SI) Pump Disable Temperature"; and LCO 3.5.3, "ECCS - Shutdown".

b. The analytical methods used to determine the RCS pressure and temperature limits and Cold Overpressure Mitigation System setpoints shall be those previously reviewed and approved by the NRC, specifically those described in the following document:

WCAP-14040-NP-A, Revision 2, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves" (includes any exemption granted by NRC to ASME Code Case N-514).

c. The PTLR shall be provided to the NRC upon issuance for each reactor vessel fluence period and for any revision or supplement thereto.

Changes to the curves, setpoints, or parameters in the PTLR resulting from new or additional analysis of beltline material properties shall be submitted to the NRC prior to issuance of an updated PTLR.

Prairie Island Units 1 and 2 5.0-37 4/1/02

Reporting Requirements 5.6 5.6 Reporting Requirements (continued) 5.6.7 Steam Generator Tube Inspection Report

1. Following each in-service inspection of steam generator tubes, if there are any tubes requiring plugging or sleeving, the number of tubes plugged or sleeved in each steam generator shall be reported to the Commission within 15 days.

2. The results of steam generator tube in-service inspections shall be included with the summary reports of ASME Code Section XI inspections submitted within 90 days of the end of each refueling outage. Results of steam generator tube in-service inspections not associated with a refueling outage shall be submitted within 90 days of the completion of the inspection. These reports shall include: (1) number and extent of tubes inspected, (2) location and percent of wall-thickness penetration for each indication of an imperfection, and (3) identification of tubes plugged or sleeved.

3. Results of steam generator tube inspections which fall into Category C-3 require notification to the Commission prior to resumption of plant operation, and reporting as a special report to the Commission within 30 days. This special report shall provide a description of investigations conducted to determine cause of the tube degradation and corrective measures taken to prevent recurrence.

4. The results of inspections performed under Specification 5.5.8.b for all tubes that have defects below the F* or EF* distance, and were not plugged, shall be reported to the Commission within 15 days following the inspection. The report shall include:

a. Identification of F* and EF* tubes, and

b. Location and extent of degradation.

Prairie Island Units 1 and 2 5.0-38 4/1/02

Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.7 Steam Generator Tube Inspection Report (continued)

5. For implementation of the voltage-based repair criteria to tube support plate intersections, notify the NRC staff prior to returning the steam generators to service should any of the following conditions arise:

a. If estimated leakage based on the projected end-of-cycle (or if not practical, using the actual measured end-of-cycle) voltage distribution exceeds the leak limit (determined from the licensing basis dose calculation for the postulated main steamline break) for the next operating cycle.

b. If circumferential crack-like indications are detected at the tube support plate intersections.

c. If indications are identified that extend beyond the confines of the tube support plate.

d. If indications are identified at the tube support plate elevations that are attributable to primary water stress corrosion cracking.

e. If the calculated conditional burst probability based on the projected end-of-cycle (or if not practical, using the actual measured end-of-cycle) voltage distribution exceeds 1 x 10-2, notify the NRC and provide an assessment of the safety significance of the occurrence.

Prairie Island Units 1 and 2 5.0-39 4/1/02

Reporting Requirements 5.6 5.6 Reporting Requirements (continued) 5.6.8 EM Report When a report is required by Condition C or J of LCO 3.3.3, "Event Monitoring (EM) Instrumentation," a report shall be submitted within the following 14 days.

The report shall outline the preplanned alternate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels of the Function to OPERABLE status.

Prairie Island Units 1 and 2 5.0-40 1/2/02

High Radiation Area 5.7 5.0 ADMINISTRATIVE CONTROLS 5.7 High Radiation Area As provided in paragraph 20.1601(c) of 10 CFR Part 20, the following controls shall be applied in place of the controls required by paragraph 10 CFR 20.1601 (a) and (b) of 10 CFR 20:

5.7.1 High Radiation Areas accessible to personnel in which radiation levels could result in an individual receiving a deep dose equivalent less than 1.0 rem in one hour at 30 centimeters from the radiation source or from any surface that the radiation penetrates

a. Each entryway to such an area shall be barricaded and conspicuously posted as a high radiation area. Such barricades may be opened as necessary to permit entry or exit of personnel or equipment.

b. Access to, and activities in each such area shall be controlled by means of a Radiation Work Permit (RWP) or equivalent that includes specification of radiation dose rates in the immediate work area(s) and other appropriate radiation protection equipment and measures.

c. Individuals qualified in radiation protection procedures and personnel continuously escorted by such individuals may be exempted from the requirement for an RWP or equivalent while performing their assigned duties provided they are otherwise following plant radiation protection procedures for entry to, exit from, and work in such areas.

d. Each individual or group entering such an area shall possess:

1. A radiation monitoring device that continuously displays radiation dose rates in the area; or

2. A radiation monitoring device that continuously integrates the radiation dose rates in the area and alarms when the device's dose alarm setpoint is reached, with an appropriate alarm setpoint; or Prairie Island Units l and 2 5.0-41 12/11/00

High Radiation Area 5.7 5.7 High Radiation Area 5.7.1 High Radiation Areas accessible to personnel in which radiation levels could result in an individual receiving a deep dose equivalent less than 1.0 rem in one hour at 30 centimeters from the radiation source or from any surface that the radiation penetrates (continued)

3. A radiation monitoring device that continuously transmits dose rate and cumulative dose information to a remote receiver monitored by radiation protection personnel responsible for controlling personnel radiation exposure within the area; or

4. A self-reading dosimeter (e.g., pocket ionization chamber or electronic dosimeter) and, (i) Be under the surveillance, as specified in the RWP or equivalent, while in the area, of an individual qualified in radiation protection procedures, equipped with a radiation monitoring device that continuously displays radiation dose rates in the area, who is responsible for controlling personnel exposure within the area; or (ii) Be under the surveillance as specified in the RWP or equivalent, while in the area, by means of closed circuit television, of personnel qualified in radiation protection procedures, responsible for controlling personnel radiation exposure in the area, and with the means to communicate with individuals in the area who are covered by such surveillance.

e. Except for individuals qualified in radiation protection procedures, or personnel continuously escorted by such individuals, entry into such areas shall be made only after dose rates in the area have been determined and entry personnel are knowledgeable of them. These continuously escorted personnel will receive a pre-job briefing prior to entry into such areas. This dose rate determination, knowledge, and pre-job briefing does not require documentation prior to initial entry.

Prairie Island Units 1 and 2 5.0-42 12/11/00

High Radiation Area 5.7 5.7 High Radiation Area (continued) 5.7.2 High Radiation Areas accessible to personnel in which radiation levels could result in an individual receiving a deep dose equivalent in excess of 1.0 rem in one hour at 30 centimeters from the radiation source or from any surface that the radiation penetrates, but less than 500 rad in one hour at one meter from the source

a. Each entryway to such an area shall be conspicuously posted as a high radiation area and shall be provided with a locked or continuously guarded door or gate that prevents unauthorized entry, and, in addition:

1. All such door and gate keys shall be maintained under the administrative control of the shift supervisor, radiation protection manager, or their designee.

2. Doors and gates shall remain locked except during periods of personnel or equipment entry or exit.

b. Access to, and activities in, each such area shall be controlled by means of an RWP or equivalent that includes specification of radiation dose rates in the immediate work area(s) and other appropriate radiation protection equipment and measures.

c. Individuals qualified in radiation protection procedures may be exempted from the requirement for an RWP or equivalent while performing radiation surveys in such areas provided they are otherwise following plant radiation protection procedures for entry to, exit from, and work in such areas.

d. Each individual or group entering such an area shall possess:

1. A radiation monitoring device that continuously integrates the radiation dose rates in the area and alarms when the device's dose alarm setpoint is reached, with an appropriate alarm setpoint; or Prairie Island Units 1 and 2 5.0-43 12/11/00

High Radiation Area 5.7 5.7 High Radiation Area 5.7.2 High Radiation Areas accessible to personnel in which radiation levels could result in an individual receiving a deep dose equivalent in excess of 1.0 rem in one hour at 30 centimeters from the radiation source or from any surface that the radiation penetrates, but less than 500 rad in one hour at one meter from the source (continued)

2. A radiation monitoring device that continuously transmits dose rate and cumulative dose information to a remote receiver monitored by radiation protection personnel responsible for controlling personnel radiation exposure within the area with the means to communicate with and control every individual in the area; or

3. A self-reading dosimeter (e.g., pocket ionization chamber or electronic dosimeter) and, (i) Be under the surveillance, as specified in the RWP or equivalent, while in the area, of an individual qualified in radiation protection procedures, equipped with a radiation monitoring device that continuously displays radiation dose rates in the area, who is responsible for controlling personnel exposure within the area; or (ii) Be under the surveillance as specified in the RWP or equivalent, while in the area, by means of closed circuit television, of personnel qualified in radiation protection procedures, responsible for controlling personnel radiation exposure in the area, and with the means to communicate with and control every individual in the area.

4. In those cases where options (2) and (3), above are impractical or determined to be inconsistent with the "As Low As is Reasonably Achievable" principle, a radiation monitoring device shall be used that continuously displays radiation dose rates in the area.

Prairie Island Units 1 and 2 5.0-44 12/11/00

High Radiation Area 5.7 5.7 High Radiation Area 5.7.2 High Radiation Areas accessible to personnel in which radiation levels could result in an individual receiving a deep dose equivalent in excess of 1.0 rem in one hour at 30 centimeters from the radiation source or from any surface that the radiation penetrates, but less than 500 rad in one hour at one meter from the source (continued)

e. Except for individuals qualified in radiation protection procedures, or personnel continuously escorted by such individuals, entry into such areas shall be made only after dose rates in the area have been determined and entry personnel are knowledgeable of them. These continuously escorted personnel will receive a pre-job briefing prior to entry into such areas. This dose rate determination, knowledge, and pre-job briefing does not require documentation prior to initial entry.

f. Such individual areas that are located within a larger area where no enclosure exists for the purpose of locking and where no enclosure can be reasonably constructed around the individual area, that individual area need not be controlled by a locked door or gate, nor continuously guarded, but shall be barricaded, conspicuously posted, and a flashing light shall be activated at the area as a warning device.

Prairie Island Units 1 and 2 5.0-45 12/11/00

I . 4.t 2-i REV 132 11/4/97 4, 1 S F 6 E -NF A5. 0-04 A . nIm' A A_ -T f m C n yr_

R-12

-A- tu iu"ci vn e trc. tlre--st--a-i gerrn b ---. --- I-n-uTii TU Ub1_1 tir in e i t f h,, LIIIUil*ie tu u c h---r -a pn it of the pi-iliUary CUo-o litTa- Ifl1T- bU. .. iidc. i Y 5..8eam generator tubes in each unit shall be determined operable by the following:

at-. Steam Generator Sample Selection and Inspection-Each steam generator shall be determined operable in accordance with the in-service inspection schedule in Specification 4 -. 2--2T5.5.8.&c. The in-service inspection may be limited to one steam generator on a rotating schedule encompassing 6% of the tubes in the single steam generator, provided the previous inspections indicated that the two steam generators are performing in a like manner.

bB-. Steam Generator Tuhe Sample Selection and Inspection-The steam generator tube minimum sample size, inspection result classification, and the corresponding action required shall be as specified in Tables T-S-.4-.. 5.5.8.-1 and T S.4 .- 125.5.8S-2. The in-service inspection of steam generator tubes shall be performed at the frequencies specified in Specification 4.r-2. 5.5.8.&c and the inspected tubes shall be verified acceptable per the acceptance criteria of Specification 4+/-22--5.5.8.f-d.

The tubes selected for each in-service inspection shall include at least 3% of the total number of tubes in all steam generators and at least 20%

of the total number of sleeves in service in both steam generators; the tubes selected for these inspections shall be selected on a random basis except:

1. Where experience in similar plants with similar water chemistry indicates critical areas to be inspected, then at least 50% of the tubes inspected shall be from these critical areas.

R-12 PI Current TS Page 5 of 41 Markup for PI ITS Part C

TS.4.12-2 REV 137*J 8/13/98..

L

.5.8 2. The first sample of tubes selected for each in-service inspection A5.0-04 (subsequent to the preservice inspection) of each steam generator shall include:

(a) All tubes that previously had detectable wall penetrations (>20%)

that have not been plugged or sleeve repaired in the affected area.

(b) Tubes in those areas where experience has indicated potential problems.

(c) A tube inspection (pursuant to Specification 5.5,8dd,.12.D-.1 .(h)) shall be performed on each selected tube.

If any selected tube does not permit the passage of the eddy current probe for a tube inspection, this shall be recorded and an adjacent tube shall be selected and subjected to a tube inspection.

3. In addition to the sample required in Specification .5.8.b4H .zrB.2.a through c, all tubes which have had the F* or EF* criteria applied will be inspected in the F* and EF* regions of the roll expanded region. The region of these tubes below the F* and EF* regions may be excluded from the requirements of 5,5.8.b 4.12.B.2.a.

4. The tubes selected as the second and third samples (if required by Tables tSw4.r25.5,8-1 or T-S4r12 5.5.8- 2 ) during each inservice inspection may be subjected to a partial tube or sleeve inspection provided:

(a) The tubes selected for these samples include the tubes from those areas of the tube sheet array where tubes with imperfections were previously found.

(b) The inspections include those portions of the tubes or sleeves where imperfections were previously found.

The results of each sample inspection shall be classified into one of the following three categories:

Category Inspection Results C-1 Less than 5% of the total tubes inspected are degraded tubes and none of the inspected tubes are defective.

C-2 One or more tubes, but not more than 1% of the total tubes inspected are defective, or between 5% and 10%

of the total tubes inspected are degraded tubes.

C-3 More than 10% of the total tubes inspected are degraded tubes or more than 1% of the inspected tubes are defective.

Note: In all inspections, previously degraded tubes must exhibit significant (>10%) further wall penetrations to be included in the above percentage calculations. R-12 PI Current TS Page 6 of 41 Markup for P1 ITS Part C

TS.4.12-3 REV 133 11/18/97 IA5.0-04 87 I5. Indications left in service as a result of application of tube support plate voltage-based repair criteria shall be inspected by bobbin coil probe during all future refueling outages.

6. Implementation of the steam generator tube/tube support plate repair criteria requires a 100 percent bobbin coil inspection for hot leg and cold leg tube support plate intersections down to the lowest cold leg tube support plate with known outside diameter stress corrosion cracking (ODSCC) indications. The determination of the lowest cold leg tube support plate intersections having ODSCC indications shall be based on the performance of at least a 20 percent random sampling of tubes inspected over their full length.

ce. Inspection Frequencies-The above required in-service inspections of steam generator tubes shall be performed at the following frequencies:

1. In-service inspections shall be performed at intervals of not less than 12 nor more than 24 calendar months after the previous inspection. If two consecutive inspections following service under AVT conditions, not including the preservice inspection, result in all inspection results falling into the C-1 category or if two consecutive inspections demonstrate that previously observed degradation has not continued and no additional degradation has occurred, the inspection interval may be extended to a maximum of once per 40 months.

2. If the results of the inservice inspection of a steam generator conducted in accordance with Table 5.5.SFTS.4.12-1 at 40 month intervals fall in Category C-3, the inspection frequency shall be increased to at least once per 20 months. The increase in inspection frequency shall apply until the subsequent inspections satisfy the criteria of Specification 5,5.8,c4*.12..1; the interval may then be extended to a maximum of once per 40 months.

3. Additional, unscheduled inservice inspections shall be performed on each steam generator in accordance with the first sample inspection specified in Table 5.5.8,TS.4.12-1 during the shutdown subsequent to any of the following conditions.

(a) Primary-to-secondary tube leaks (not including leaks originating from tube-to-tube sheet welds) in excess of the limits of Specification 3.414 4.4C 3. .6.

(b) A seismic occurrence greater than the Operating Basis Earthquake.

(c) A loss-of-coolant accident requiring actuation of the engineered safeguards.

(d) A main steam line or feedwater line break.

'R-12 PI Current TS Page 7 of 41 Markup for PI ITS Part C

"TS. 4.12 -4 REV 144 4/15/99 dD-. Acceptance Criteria Specification:

1. As used in this (a) Imperfection means an exception to the dimensions, finish or contour of a tube from that required by fabrication drawings or specifications. Eddy-current testing indications below 20% of the nominal tube wall thickness, if detectable, may be considered as imperfections, (b) Degradation means a service-induced cracking, wastage, wear or general corrosion occurring on either inside or outside of a tube.

(c) Degraded Tube means a tube containing imperfections >20% of the nominal wall thickness caused by degradation.

(d) % Degradation means the percentage of the tube wall thickness affected or removed by degradation.

(e) Defect means an imperfection of such severity that it exceeds the repair limit. A tube containing a defect is defective.

(f) Repair Limtml means the imperfection depth at or beyond which the tube shall be removed from service by plugging or repaired by sleeving because it may become unserviceable prior to the next inspection and is equal to 50% of the nominal tube wall thickness.

If significant general tube thinning occurs, this criteria will be reduced to 40% wall penetration. This definition does not apply to the portion of the tube in the tubesheet below the F* or EF*

distance provided the tube is not degraded (i.e., no indications of cracks) within the F* or EF* distance for F* or EF* tubes. The repair limit for the pressure boundary region of any sleeve is 25%

of the nominal sleeve wall thickness. This definition does not apply to tube support plate intersections for which the voltage based repair criteria are being applied. Refer to Specification 5.5.8.d4+/-,.tD.4 for the repair limit applicable to these intersections.

(g) Ulnserviceable describes the condition of a tube if it leaks or contains a defect large enough to affect its structural integrity in the event of an Operating Basis Earthquake, a loss-of-coolant accident, or a steam line or feedwater line break.

(h) Tube Inspection means an inspection of the steam generator tube from the point of entry (hot leg side) completely around the U-bend to the top support of the cold leg.

(i) Sleeving is the repair of degraded tube regions using a new Alloy 690 tubing sleeve inserted inside the parent tube and sealed at each end by welding or by replacing the lower weld in a full depth tubesheet sleeve with a hard rolled joint. The new sleeve becomes the pressure boundary spanning the original degraded tube region.

R-12 PI Current TS Page 8 of 41 Markup for PI ITS Part C

TS.4.12-5 REV 153 4/19/00 (j) F* Distance is the distance from the bottom of the hardroll trans As.O-04 toward the bottom of the tubesheet that has been conservatively determined to be 1.07 inches (not including eddy current uncertainty).

The F* distance applies to roll expanded regions below the midplane of the tubesheet (k) F* Tube is a tube with degradation, below the F* distance, equal to or greater than 40%, and not degraded (i.e., no indications of cracking) within the F* distance.

(1) RF* Distance is the distance from the bottom of the upper hardroll transition toward the bottom of the tubesheet that has been conservatively determined to be 1.67 inches (not including eddy current uncertainty). EF* distance applies to roll expanded regions when the top of the additional roll expansion is 2.0 inches or greater down from the top of the tubesheet (m) FF* Tube is a tube with degradation, below the EF* distance, equal to or greater than 40%, and not degraded (i.e., no indications of cracking) within the EF* distance.

2. The steam generator shall be determined OPERABLE after completing the corresponding actions (plug or repair by sleeving all tubes exceeding the repair limit and all tubes containing through-wall cracks or classify as F* or EF* tubes) required by Tables 5.5.8S*.-4.19-1 and 5,5.8TS.4.t2-2.

3. Tube repair, after April 1, 1999, using Combustion Engineering welded sleeves shall be in accordance with the methods described in the following:

CEN-629-P, Revision 03-P, "Repair of Westinghouse Series 44 and 51 Steam Generator Tubes Using Leak Tight Sleeves";

4. Tube Support Plate Repair Limit is used for the disposition of a steam generator tube for continued service that is experiencing predominantly axially oriented outside diameter stress corrosion cracking confined within the thickness of the tube support plates. At tube support plate intersections, the repair limit is based on maintaining steam generator serviceability as described below:

a. Steam generator tubes, whose degradation is attributed to outside diameter stress corrosion cracking within the bounds of the tube support plate with bobbin voltages less than or equal to 2.0 volts will be allowed to remain in service.

b. Steam generator tubes, whose degradation is attributed to outside diameter stress corrosion cracking within the bounds of the tube support plate with a bobbin voltage greater than 2.0 volts, will be repaired or plugged, except as noted in Specification 4.12.D5.5.8.d.4.c below.

R-12:

PI Current TS Page 9 of 41 Markup for P1 ITS Part C

REV 137 8/13/98 A5.0-04

c. Steam generator tubes, with indications of potential degradation attributed to outside diameter stress corrosion cracking within the bounds of the tube support plate with a bobbin voltage greater than 2.0 volts but less than or equal to the upper voltage repair limit, may remain in service if a rotating pancake coil (or comparable examination technique) inspection does not detect degradation. Steam generator tubes, with indications of outside diameter stress corrosion cracking degradation with a bobbin voltage greater than the upper voltage repair limit will be plugged or repaired.

d. If an unscheduled mid-cycle inspection is performed, the following mid-cycle repair limits apply instead of the limits in Specifications 5.58.d.4-5l-2--.D.4.('a), (b) and (,c). The mid-cycle repair limits are determined from the following equations:

VSTL VS Vmurl =

1.0+ NDE+Gr yCL-ATj

\ _CL (Vurl -2.0) rCL-AT; Vmiri = Vmurl -

( CL where:

VURL = upper voltage repair limit VLRL= lower voltage repair limit VMURL = mid-cycle upper voltage repair limit based on time into cycle and VMLRL = mid-cycle lower voltage repair limit based on VMURL time into cycle At = length of time since last scheduled inspection during which VURL and VLRL were implemented CL = cycle length (time between two scheduled steam generator inspections)

VSL = structural limit voltage Gr = average growth rate per cycle length NDE= 95 percent cumulative probability allowance for nondestructive examination uncertainty (i.e., a value of 20 percent has been approved by the NRC) same Implementation of these mid-cycle repair limits should follow the (b) and approach as described in Specifications 5.5.8.d.4+/-2wIY4j(a),

(c).

Note: The upper voltage repair limit is calculated according to the methodology in Generic Letter 95-05 as supplemented.

R-12:

Page 10 of 41 Markup for PI ITS Part C PI Current TS

T-S.4-.1-2--7 REV 137 8/H13/-'9,8 Er. Steam Generator Tube Inspection e port

1. Following each in-service inspection of steam generator tubes, if there are any tubes requiring plugging or sleeving, the number of tubes plugged or sleeved in each steam generator shall be reported to the Commission within 15 days.

2. The results of steam generator tube inservice inspections shall be included with the summary reports of ASME Code Section XI inspections submitted within 90 days of the end of each refueling outage. Results of steam generator tube inservice inspections not associated with a refueling outage shall be submitted within 90 days of the completion of the inspection. These reports shall include: (1) number and extent of tubes inspected, (2) location and percent of wall-thickness penetration for each indication of an imperfection and (3) identification of tubes plugged or sleeved.

3. Results of steam generator tube inspections which fall into Category C-3 require notification to the Commission prior to resumption of plant operation, and reporting as a special report to the Commission within 30 days. This special report shall provide a description of investigations conducted to determine cause of the tube degradation and corrective measures taken to prevent recurrence.

4. The results of inspections performed under Specification 5.5.8.b. 4- for all tubes that have defects below the F* or EF* distance, and were not plugged, shall be reported to the Commission within 15 days following the inspection. The report shall include:

a. Identification of F* and EF* tubes, and

b. Location and extent of degradation.

5. For implementation of the voltage-based repair criteria to tube support plate intersections, notify the NRC staff prior to returning the steam generators to service should any of the following conditions arise:

a. If estimated leakage based on the projected end-of-cycle (or if not practical, using the actual measured end-of-cycle) voltage distribution exceeds the leak limit (determined from the licensing basis dose calculation for the postulated main steamline break) for the next operating cycle.

b. If circumferential crack-like indications are detected at the tube support plate intersections.

c. If indications are identified that extend beyond the confines of the tube support plate.

d. If indications are identified at the tube support plate elevations that are attributable to primary water stress corrosion cracking.

e. If the calculated conditional burst probability based on the projected end-of-cycle (or if not practical, using the actual measured end-of cycle) voltage distribution exceeds I x 10-2, notify the NRC and provide an assessment of the safety significance of the occurrence. -------

PRf124I PI Current TS Page 11 of 41 Markup for PI ITS Part C

TABLE 5.5.8'f-S.4.12-1 STEAM GENERATOR TUBE INSPECTION 1ST SAMPLE INSPECTION 2ND SAMPLE INSPECTION 3RD SAMPLE INSPECTION Sample Size Result Action Required Result Action Required Result Action Required A minimum of S C-1 None N/A N/A N/A N/A Tubes per S.G. C-2 Repair defective tubes and C-1 None N/A N/A inspect additional 2S tubes in C-2 Repairdefective tubes and C-i None this S.G. inspect additional 4S tbbes in C-2 Repair defective tubes this S.G.

C-3 Perform action for C-3 result of first sample C-3 Perform action for C-3 result of N/A N/A first sample C-3 Inspect all tubes in this S.G., All other S.G.s are None N/A N/A Repair defective tubes and C-1 inspect 2S tubes in each other Some S.G.s C-2 Perform action for C-2 result of N/A N/A NRC. but no additional second sample S.G. are C-3 _

Additional S.G. is Inspect all tubes in each S.G. and N/A N/A C-3 repair defective tubes.

Prompt notification to NRC.

S=3%; When two steam generators are inspected during that outage.

S=6%; When one steam generator is inspected during that outage.

(.

C, FA V N

h 0

'R-12 PI Current TS Page 12 of 41 Markup for PI ITS Part C

TABLE 5.5.8TS-4ct--2 A5.0-4 Steam Generator Tube Sleeve Inspection 1St S Sample Inspection 2n Sa le Inspection Sample Size Result Action Required Result Action Required A minimum of 20% of Tube C-1 None N/A N/A Sleeves (1) C-2 Inspect all remaining tube sleeves in this S.G. C-1 None and plug or repair defective sleeved tubes. C-2 Plug or repair defective sleeved tubes C-3 Perform action for C-3 result of first

-- sample C-3 Inspect all tube sleeves in this S.G., inspect 20% The other S.G. is None of the tube sleeves in the other S.G., and plug or C-1 repair defective sleeved tubes The other S.G. is Perform action for C-2 results of first C-2 sample The other S.G. is Inspect all tube sleeves in each S.G. and C-3 plug or repair defective sleeved tubes (1) Each type of sleeve is considered a separate population for detennination of scope expansion F-P

<2

R-12 PI Current TS Page 13 of 41 Markup for PI ITS Part C

OGverflo1.W LCcD 3.4.12, "L1ow Temi-perature 0Ove rpres sure Protection (LTOP),

Reactor Coolant System Cold Leg Temperature 1 5.6.6 (RCSCIJT) > Safety Injection (SI) Pump Disable R-12 Temperatlure"; I .1 LCO 3.4.1L3, "Low., TIemperature Overpres sure Protection (LTOP)

Reactor Coolant System Cold Leg Temperature

(.RCSCLT) *< Safety Injection (SI) PupDisable Temperature"; and ,R-12 LCO 3.53,-, "71CCS - Shutdown.

b2-.The analytical methods used to determine the RCS pressure and temperature limits and Cold Overpressure Mitigation System setpoints shall be those previously reviewed and approved by the NRC, specifically those described in the following document:

WCAP-14040-NP-A, Revision 2, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves" (4-includes any exemption granted by NRC to ASME Code Case N-514)

A5.0-38 R-7 PI Current TS Page 35 of 41 Markup for PI ITS Part C

Part D Package 5.0 NSHD Change category number Discussion Of Change 5.0 LR 02 4.2.A.2. The CTS requirements for inservice testing have been relocated to the Inservice Testing (IST) Program in accordance with the guidance of NUREG-1431. This change is acceptable since the IST is required by the Administrative Controls Section 5.5. Since the program definition has been moved to the Administrative Controls section of the ITS this is a less restrictive change.

LR 03 Table 4.2-1 and 6.5.F. The CTS requirements for reactor coolant pump flywheel inspection have been relocated to the Reactor Coolant Pump Flywheel Inspection Program which is required by ITS Administrative Controls Section 5.5. This change is acceptable since reactor coolant pump flywheel inspection continues to be required by ITS Section 5.5.

Since the program definition has been moved to the Administrative Controls section of the ITS, this is a less restrictive change. This change is consistent with the guidance of NUREG-1431.

A 04 CTS 4.12. The CTS requirements for Steam Generator (SG) tube surveillance in CTS 4.12.A through D have been included in the SG Tube Surveillance Program in the ITS Administrative Controls Section 5.5.8. CTS 4.12.E has been included in the Steam Generator Tube Inspection Report in ITS Administrative Controls Section 5.6.7. This change is acceptable since SG tube surveillance will continue to be required in accordance with the new program and report.

Since there are no changes in technical requirements, this is an administrative change.

Prairie Island Units 1 and 2 2 4/1/02

Part D Package 5.0 NSHD Change category number Discussion Of Change 5.0 A 05 CTS 4.12.E. The CTS requirements for steam generator (SG) tube surveillance reports in CTS 4.12.E have been included in the SG Tube Inspection Report which is required by the ITS Administrative Controls Section 5.6.7. This change is acceptable since SG tube surveillance reports will continue to be required in accordance with the new program and report. Since there are no changes in technical requirements, this is an administrative change. This change is consistent with the guidance of NUREG-1431.

A 06 Throughout the CTS 6.0 markup, the section and paragraph numbering, punctuation and paragraph references have been revised to correspond to the NUREG-1431 format, numbering and punctuation. Since these changes do not introduce any substantive requirement changes, these changes are administrative.

A 07 CTS 6.5.L.2.b The use of "involves an unreviewed safety question as defined in" has been replaced by "requires NRC approval pursuant to" to be consistent with the most recent issuance of 10CFR50.59. Since this does not involve any substantive changes, this is an administrative change. This change is consistent with NUREG-1431 as modified by approved TSTF-364.

08 Not used.

09 Not used.

10 Not used.

Prairie Island Units 1 and 2 3 4/1/02

Part D Package 5.0 NSHD Change category number Discussion Of Change 5.0 A 26 6.6.A, B and C. In conformance with the guidance of NUREG-1431 as modified by TSTF-152, a note is included to clarify the ITS reporting requirements. Since this does not change the reporting requirements for PI this change is administrative.

A 27 CTS 6.6.D. The CTS requirement for monthly reporting of challenges to the pressurizer power operated relief valves or pressurizer safety valves is not included in the ITS. In accordance with NRC GL 97-02, "Revised Contents of the Monthly Operating Report", the NRC has requested less information in the monthly operating report. This generic leter identifies what needs to be reported to support the NRC Performance Indicator Program, and availability and capacity statistics. The generic letter does not identify the need to report these valve challenges and thus in conformance with the guidance of NUREG-1431 as modified by TSTF-258, Rev. 4, these challenges are not included in the ITS. Since these changes only involve reporting requirements and do not affect the safe operation of the plant this is an administrative change.

A 28 CTS 6.6.E. A new COLR reference to the latest Prairie Island approved steam line break methodology is included.

Since the change is just a new reference which was previously reviewed and approved by the NRC in a letter dated January 21, 2000, this is an administrative change in this submittal. The status designator for reports NSPNAD 8101 and 8102 have been corrected to "PA" to indicate that these are proprietary documents. These changes do not materially change these reports, thus these are also administrative changes.

Prairie Island Units I and 2 8 4/1/02

Programs and Manuals 5.5 5.5 Programs and Manuals (continued)

Weekly At least once per 7 days Monthly At least once per 31 days Semiquarterly At least once per 46 days Quarterly or every 3 months At least once per I PA5.0 92 days Semiannually or every 6 months At least once per 184 days Every 9 months At least once per 276 days Yearly or annually At least once per 366 days Biennially or every 2 years At least once per 731 days

b. The provisions of SR 3.0.2 are applicable to the above required Frequencies for performing inservice testing activities;

c. The provisions of SR 3.0.3 are applicable to inservice testing activities; and

d. Nothing in the ASME Boiler and Pressure Vessel Code shall be construed to supersede the requirements of any TS.

5.5.89 Steam Generator (SG) Tube CL5.0-64 II Surveillance Proqram s

is II I Reviewer's Note: The Licensce's eurr-ent licensing ba generator tube surveillancek euicents shall be rcl IV ,L V -

from the LG0 and included here. An appropriate admir Steam qenerator tubes in each unit shall be determined operable by the following:

a. Steam Generator Sample Selection and Inspection Each steam generator shall be determined operable in I *-----I--- i R-12 I

-- -I (conti nued)

WOG STS Rev 1, 04/07/95 5.0-16 Markup for PI ITS Part E

Programs and Manuals 5.5 5.5 Programs and Manuals (continued) accordance with the in-service inspection schedule in Specification 5.5.8.c. The in-service inspection may be limited to one steam qenerator on a rotatinq schedule encompassinq 6% of the tubes in the sinqle steam qenerator, provided the previous inspections indicated that the two steam generators are performing in a like manner.

b. Steam Generator Tube Sample Selection and Inspection The steam qenerator tube minimum sample size, inspection result classification, and the correspondinq action required shall be as specified in Tables 5.5.8-1 and 5.5.8-2. The in-service inspection of steam qenerator tubes shall be performed at the frequencies specified in Specification 5.5.8.c and the inspected tubes shall be verified acceptable per the acceptance criteria of Specification 5.5.8.d. The tubes selected for each in service inspection shall include at least 3% of the total number of tubes in all steam qenerators and at least 20%

of the total number of sleeves in service in both steam generators; the tubes selected for these inspections shall be selected on a random basis except:

1. Where experience in similar plants with similar water chemistry indicates critical areas to be inspected, then at least 50% of the tubes inspected shall be from these critical areas.

2. The first sample of tubes selected for each in-service inspection (subsequent to the preservice inspection) of each steam generator shall include:

(a) All tubes that previously had detectable wall penetrations (>20%) that have not been plugged or sleeve repaired in the affected area.

(b) Tubes in those areas where experience has indicated potential problems.

(continued)!LR-2 I R-12 WOG STS Rev 1, 04/07/95 5.0-17 Markup for PI ITS Part E

Programs and Manuals 5.5 5.5 Programs and Manuals (continued)

(c) A tube inspection (pursuant to Specification 5.5.8.d.l(h)) shall be performed on each selected tube. If any selected tube does not permit the passage of the eddy current probe for a tube inspection, this shall be recorded and an adjacent tube shall be selected and subjected to a tube inspection.

3. In addition to the sample required in Specification 5.5.8.b.2(a) through (c),

all tubes which have had the F* or EF*

criteria applied will be inspected in the F* and EF* regions of the roll expanded region. The region of these tubes below the F* and EF* regions may be excluded from the requirements of Specification 5.5.8.b.2(a).

4. The tubes selected as the second and third samples (if required by Tables 5.5.8-1 or 5.5.8-2) during each in-service inspection may be subjected to a partial tube or sleeve inspection provided:

(a) The tubes selected for these samples include the tubes from those areas of the tube sheet array where tubes with imperfections were previously found.

(b) The inspections include those portions of the tubes or sleeves where imperfections were previously found.

The results of each sample inspection shall be classified into one of the following three categories:

Category Inspection Results C-I Less than 5% of the total tubes inspected are degraded tubes and none of the inspected tubes are defective. r ------- I 1R-12 I (continued) L-------

WOG STS Rev 1, 04/07/95 5.0-18 Markup for PI ITS Part E

Programs and Manuals 5.5 5.5 Programs and Manuals (continued)

C-2 One or more tubes, but not more than 1% of the total tubes inspected are defective, or between 5% and 10% of the total tubes inspected are degraded tubes.

C-3 More than 10% of the total tubes inspected are degraded tubes or more than 1% of the inspected tubes are defective.

Note: In all inspections, previously degraded tubes must exhibit significant (>10%)

further wall penetrations to be included in the above percentage calculations.

5. Indications left in service as a result of application of tube support plate voltage-based repair criteria shall be inspected by bobbin coil probe during all future refueling outages.

6. Implementation of the steam generator tube/tube support plate repair criteria requires a 100 percent bobbin coil inspection for hot leg and cold leg tube support plate intersections down to the lowest cold leg tube support plate with known outside diameter stress corrosion cracking (ODSCC) indications. The determination of the lowest cold leg tube support plate intersections having ODSCC indications shall be based on the performance of at least a 20 percent random sampling of tubes inspected over their full length.

c. Inspection Frequencies The above required in-service inspections of steam generator tubes shall be performed at the following frequencies:

1. In-service inspections shall be performed at intervals of not less than 12 nor more than 24 calendar months after the previous inspection.

I R-12 (conti nued) 1L-------I WOG STS Rev 1, 04/07/95 5.0-19 Markup for PI ITS Part E

Programs and Manuals 5.5 5.5 Programs and Manuals (continued)

If two consecutive inspections following service under AVT conditions, not including the preservice inspection, result in all inspection results falling into the C-1 category or if two consecutive inspections demonstrate that previously observed degradation has not continued and no additional degradation has occurred, the inspection interval may be extended to a maximum of once per 40 months.

2. If the results of the in-service inspection of a steam generator conducted in accordance with Table 5.5.8-1 at 40 month intervals fall in Category C-3, the inspection frequency shall be increased to at least once per 20 months. The increase in inspection frequency shall apply until the subsequent inspections satisfy the criteria of Specification 5.5.8.c.1; the interval may then be extended to a maximum of once per 40 months.

3. Additional, unscheduled in-service inspections shall be performed on each steam generator in accordance with the first sample inspection specified in Table 5.5.8-1 during the shutdown subsequent to any of the following conditions:

(a) Primary-to-secondary tube leaks (not including leaks originating from tube-to-tube sheet welds) in excess of the limits of Specification 3.4.14.

(b) A seismic occurrence greater than the Operating Basis Earthquake.

(c) A loss-of-coolant accident requiring actuation

- engineered of the - - safeguards.

I R-12 I (continued} ------

WOG STS Rev 1, 04/07/95 5.0-20 Markup for PI ITS Part E

Programs and Manuals 5.5 5.5 Programs and Manuals (continued)

(d) A main steam line or feedwater line break.

d. Acceptance Criteria

1. As used in this Specification:

(a) Imperfection means an exception to the dimensions, finish or contour of a tube from that required by fabrication drawings or specifications.

Eddy-current testing indications below 20% of the nominal tube wall thickness, if detectable, may be considered as imperfections.

(b) Degradation means a service-induced cracking, wastage, wear or general corrosion occurring on either inside or outside of a tube.

(c) Degraded Tube means a tube containing imperfections

Ž20% of the nominal wall thickness caused by degradation.

(d) % Degradation means the percentage of the tube wall thickness affected or removed by degradation.

(e) Defect means an imperfection of such severity that it exceeds the repair limit. A tube containing a defect is defective.

(f) Repair Limit means the imperfection depth at or beyond which the tube shall be removed from service by plugging or repaired by sleeving because it may become unserviceable prior to the next inspection and is equal to 50% of the nominal tube wall thickness. If significant general F ....... I IR-12 (continud WOG STS Rev 1, 04/07/95 5.0-21 Markup for PI ITS Part E

Programs and Manuals 5.5 5.5 Programs and Manuals (continued) tube thinning occurs, this criteria will be reduced to 40% wall penetration. This definition does not apply to the portion of the tube in the tubesheet below the F*

distance provided the tube is not degraded (i.e., no indications of cracks) within the F* or EF* distance for F* or EF* tubes.

The repair limit for the pressure boundary region of any sleeve is 25% of the nominal sleeve wall thickness. This definition does not apply to tube support plate intersections for which the voltage-based repair criteria are being applied. Refer to Specification 5.5.8.d.4 for the repair limit applicable to these intersections.

(g) Unserviceable describes the condition of a tube if it leaks or contains a defect large enough to affect its structural integrity in the event of an Operating Basis Earthquake, a loss-of-coolant accident, or a steam line or feedwater line break.

(h) Tube Inspection means an inspection of the steam generator tube from the point of entry (hot leg side) completely around the U-bend to the top support of the cold leg.

(i) Sleeving is the repair of degraded tube regions using a new Alloy 690 tubing sleeve inserted inside the parent tube and sealed at each end by welding or by replacing the lower weld in a full depth tubesheet sleeve with a hard rolled joint. The new sleeve becomes the pressure boundary spanning the original degraded tube region.

(j) F* Distance is the distance from the bottom 1 R-12 I (continued)L ------- i WOG STS Rev 1, 04/07/95 5.0-22 Markup for PI ITS Part E

Programs and Manuals 5.5 5.5 Programs and Manuals (continued) of the hardroll transition toward the bottom of the tubesheet that has been conservatively determined to be 1.07 inches (not including eddy current uncertainty).

The F* distance applies to roll expanded regions below the midplane of the tubesheet.

(k) F* Tube is a tube with degradation, below the F* distance, equal to or greater than 40%, and not degraded (i.e., no indications of cracking) within the F* distance.

(1) EF* Distance is the distance from the bottom of the upper hardroll transition toward the bottom of the tubesheet that has been conservatively determined to be 1.67 inches (not including eddy current uncertainty).

EF* distance applies to roll expanded regions when the top of the additional roll expansion is 2.0 inches or greater down from the top of the tubesheet.

(m) EF* Tube is a tube with degradation, below the EF*

distance, equal to or greater than 40%, and not degraded (i.e., no indications of cracking) within the EF* distance.

2. The steam generator shall be determined OPERABLE after completing the corresponding actions (plug or repair by sleeving all tubes exceeding the repair limit and all tubes containing through-wall cracks or classify as F* or EF* tubes) required by Tables 5.5.8-1 and 5.5.8-2.

3. Tube repair, after April 1, 1999, using Combustion Engineering welded sleeves shall be in accordance with the methods described in the following:

"Repair of Westinghouse r CEN-629-P, Revision 03-P, SR-12 i

(conti nued)(con WOG STS Rev 1, 04/07/95 5.0-23 Markup for PI ITS Part E

Programs and Manuals 5.5 5.5 Programs and Manuals (continued)

Series 44 and 51 Steam Generator Tubes Using Leak Tight Sleeves";

4. Tube Support Plate Repair Limit is used for the disposition of a steam generator tube for continued service that is experiencing predominantly axially oriented outside diameter stress corrosion cracking confined within the thickness of the tube support plates. At tube support plate intersections, the repair limit is based on maintaining steam generator serviceability as described below:

(a) Steam generator tubes, whose degradation is attributed to outside diameter stress corrosion cracking within the bounds of the tube support plate with bobbin voltages less than or equal to 2.0 volts will be allowed to remain in service.

(b) Steam generator tubes, whose degradation is attributed to outside diameter stress corrosion cracking within the bounds of the tube support plate with a bobbin voltage greater than 2.0 volts, will be repaired or plugged, except as noted in Specification 5.5.8.d.4(c) below.

(c) Steam generator tubes, with indications of potential degradation attributed to outside diameter stress corrosion cracking within the bounds of the tube support plate with a bobbin voltage greater than 2.0 volts but less than or equal to the upper voltage repair limit, may remain in service if a rotating pancake coil (or comparable examination R-12 ,

(continued)L WOG STS Rev 1, 04/07/95 5.0-24 Markup for PI ITS Part E

Programs and Manuals 5.5 5.5 Programs and Manuals (continued) technique) inspection does not detect degradation. Steam generator tubes, with indications of outside diameter stress corrosion cracking degradation with a bobbin voltage greater than the upper voltage repair limit will be plugged or repaired.

(d) If an unscheduled mid-cycle inspection is performed, the following mid-cycle repair limits apply instead of the limits in Specifications 5.5.8.d.4(a), (b) and (c).

The mid-cycle repair limits are determined from the following equations:

VL VMURL - SL 1.0+NDE+Gr( CL-At VMLRL VMURL (V URL -2.0)( CL -At)

CL where:

VURL upper voltage repair limit VLRL lower voltage repair limit VMmJL = mid-cycle upper voltage repair limit based on time into cycle r- . . .. .

IR-12 L-(conti nued)

WOG STS Rev 1, 04/07/95 5.0-25 Markup for PI ITS Part E

Programs and Manuals 5.5 5.5 Programs and Manuals (continued)

VMLRL = mid-cycle lower voltage repair limit based on V,,L and time into cycle At = length of time since last scheduled inspection during which VRL and VRL were implemented CL = cycle length (time between two scheduled steam generator inspections)

V = structural limit voltage Gr = average growth rate per cycle length NDE = 95 percent cumulative probability allowance for nondestructive examination uncertainty (i.e., a value of 20 percent has been approved by the NRC)

Implementation of these mid-cycle repair limits should follow the same approach as described in Specifications 5.5.8.d.4(a), (b) and (c).

Note: The upper voltage repair limit is calculated according to the methodology in Generic Letter 95-05 as supplemented.

5.5.10 * =,* GheEfistrv Seendary Water '°'

,,* **y - '*Proeram

,,=.... JR-12 1

L-------.

This progra provides ontrols for onitorin seondary CL5.0-56 (continued)

WOG STS Rev 1, 04/07/95 5.0-26 Markup for PI ITS Part E

Programs and Manuals 5.5 5.5 Programs and Manuals (continued) chemistry to inhibit SG tube degradation and low pressure turbine disc stress corrosion cracking .The programf shall-i nclude:.

a. identification of a sampling shedule for the critical vari.ables and control points for these variables;

b. Identification of the procedures used to measure the val ues of the cri ti cal variableCs;,

e. identification of process sampling points, whieh shall include monitoring the discharge of the condensate pumps for evidence of condenser in leakage;

d. Procedures for the recording and management of data;

e. Procedures defining corrective actions for all off control point chcmistry conditions; and

f. A procedurc identifying the authority responsible for the interpretation of the data and the sequence and timning of admi ni strativye events, whi ch is requi red. to initiat corrective action.

5.5.9-1 Ventilation Filter Testing Program (VFTP)

A program shall be established to implement the following required testing of Engineered Safety Feature (ESF) filter ventilation systems and the Spent Fuel Pool Special and I CL5.0-66 I Inservice Purge Ventilation System each operating cycle (18 months for shared systems)at the frequencies specified in

[Regulatory Guide ], and in accordance with [Regulatory Guide 1.52, Revision 2, ASME N510 1989, and AG I].

(conti nued)

WOG STS Rev 1, 04/07/95 5.0-27 Markup for PI ITS Part E

Programs and Manuals 5.5 5.5 Programs and Manuals (continued)

Demonstrate for the Auxiliary Building Special Ventilation, Shield Building Ventilation, Control Room Special Ventilation, and Spent Fuel Pool Special and Inservice Purge Ventilation Systems that:

a. Demnstratc for eah of the [SF systems ,A-an inpiace DOP test of the high efficiency particulate air (HEPA) filters shows a penetration and system bypass < 1 *1% (for DOP, particles having a mean diameter of 0.7 microns); whe-n tested in accordance with [Regulatory -Uie 1-52, Revision 2, and ASME NSI -n989] at thc.. ystc flow...e specified below .. 10*ý].

ESF Ventilation Syst"emf FIawae SI

-74 I 5.5.9-1 Ventilation Filter Testing Program (VFTP) (continued)

b. Demonstrate for cah of the SF systems thatA CL5.0-66 inplaee halogenated hydrocarbon test of the inplace charcoal adsorber shows a penetration and system bypass < 1-[O-.*-d-% (for DOP, particles having a mean diameter of 0.7 microns); when tested in accordance 1 r7--) n ,, ' 4^' ") 4

,,r *4-* ,!*4 * , P ,44*

with [Regulatory Guijde 1.5, evision 2,and ASM[ N510 1989] at the system flowrate specified below E+/-J 'I lm ] . - . -) J l . * -II 'JV 1 .I, . ,.i* .I I .. ,J i. w

[SF Ventilation Syst-em Florat I SCL5.0-66 F

I I (conti nued)

WOG STS Rev 1, 04/07/95 5.0-28 Markup for PI ITS Part E

Programs and Manuals 5.5 5.5 Programs and Manuals (continued)

C. . for each of the [SF systems that aA laboratory test

.monstratc of a sample of the charcoal adsorber, when obtained as described in [Regulatory Guide 1.52, Revision 2+, shows the methyl iodide penetration less than 15% penetration (less than 5% penetration for the Control Room Special Ventilation System)the value specified b,4.ow when tested in accordance with

-ASTM D3803-1989- at a temperature of _--30°C land 95%-greate-r than or equal to the relative humidity (RH) (or 70% RH with humidity controls if the humidity controls are capable of maintaining the humidity of the air entering the charcoal less than or equal to 70% RH under worst-case design-basis conditions); andsdeeifie belo..

[SF Ventil ation Sstem Pnetratio l Rif R-5 F L ------- -

[~~~~F SI 1I I I nevielwm'lr'3 Nltu,.., rAl L,.*owa-le peln '.-t . "-i L Fin oo° lu*.

evaluation]' 's ft y-f*.... o.^'

Safety factor - [5] for systems with h-aters

d. monstrate*for each of the

.. SF systems-thatT-the pressure drop across the combined HEPA filtersj the prefil*^',, and the charcoal adsorbers is less than 6 inches of CL5.0-66]

water at the val ....... below when t e "

in ......

with [Regulatory Quic 1.52 5 .5 .ii.

Ventilatiorn, Fil t er, ,Pe t n o. * ,r , ,

r,,_ v ,f,,

Revision 2, an ,v-. at the system flowrate specified below i+/- 1 0%}.

(continued)

WOG STS Rev 1, 04/07/95 5.0-29 Markup for PI ITS Part E

Programs and Manuals 5.5 5.5 Programs and Manuals (continued)

ESF Ventilation Syst*m Delta P Flowrate, I I

e. Demonstrate that,SF the heaters for each of the systefs dissioate thc val ue spe.i f4ied below [+/- 10%] when tested +

in accordance with [ASME N519 1989]. t ESF Ventilation System Wattage I I I i The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the VFTP test frequencies.

5.5.102 Explosive Gas and Storage Tank Radioactivity Monitoring Program This program provides controls for potentially explosive gas mixtures contained in the waste gas holdup system EWaste-G CL5.0-56 HodupSyse*"-, fthe quantity of radioactivity contained in gas storage tanks or fed into the o,,a treatment syst,,, and the quantity of radioactivity contained in unprotected outdoor liquid storage tanks3. The gaseous radioactivity quantities s be deterniined following the mfethodology in [Branch Teehnical Position (BTP) [TSB 11 5, "Postulated Radioactive Release due to Waste Gas Systecm Leak or Failure"]. The liquid radwaste quantities shall be determined in accordance with [Standard Review Plant, Section 15.73, "Postulated Radioactive Release due to Tank Failures"].

(continued)

WOG STS Rev 1, 04/07/95 5.0-30 Markup for PI ITS Part E

Programs and Manuals 5.5 5.5 Programs and Manuals (continued)

The program shall include:

a. The limits for concentrations of hyd"rog.ena oxygen in the waste gas holdup system [Waste Gas Holdup System] and a surveillance program to ensure the limits are maintained. Such limits shall be 5.5.10Z Explosive Gas and Storage Tank Radioactivity Monitoring Program (continued) appropriate to the system's design criteria CL5.0-5671 (i.e., whether or not the system is designed to withstand a hydrogen explosion);

b. A surveillance program to ensure that the quantity of radioactivity contained in -each gas storage tank and fed into the offgas treatment system] is less than or equal to 78,800 curies of noble gas (considered as dose equivalent Xe-133) the amount that would result in a whole body exposure of > 0.5 rem to any individua -an in unrestricted area, in the event of [an uncontrolled release of the tanks' contents]; and

c. A surveillance program to ensure that the quantity of radioactivity contained in each of the following tanks shall be limited to 10 curies, excluding tritium and dissolved or entrained noble gases:

Condensate storage tanks Outside temporary tanks "alloutdoor liquid radwaste tanks that are not surrounded by liners, dikes, or walls, capable of holding the tanks' contents and that do not have tank overflows and surrounding area drains connected to the

[Liquid Radwaste Treatment System] is less than the amlount th would result in **n*entrations less than the limits of 10 CFR 20, Appendix B, Table 2, Column 2, at the nearest potable water supply (continued)

WOG STS Rev 1, 04/07/95 5.0-31 Markup for PI ITS Part E

Programs and Manuals 5.5 5.5 Programs and Manuals (continued) and the nlearest surfaee water supply in an unrestricted area, in the event of an uncontrolled release of the tanks' contents.

The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Explosive Gas and Storage Tank Radioactivity Monitoring Program surveillance frequencies.

5.5.113 Diesel Fuel Oil Testing Program A diesel fuel oil testing program to implement required testing of both new fuel oil and stored fuel oil shall be established. The CL5.0-56 program shall include sampling and testing requirements, and acceptance criteria, all in accordance with the limits specified in Table 1 of ASTM D975-77 when checked for viscosity, water, and sediment applicable A,,M Standa . Th purpose of the program is to establis the foilloing1:

a7--Acceptability of new fuel oil shall be determined for use prior to addition to the safeguards storage tanks. Testing of diesel fuel oil stored in the safeguards storage tanks shall be performed at least every 31 days. by determining that the fuel oil has;

i. an API gravity or an absolute specifi gravity wi limits, I i

i%

5J. 5J. .i3, ^C IIC W A921 11 -'~-~---

'~

UiE.*E I ,L.Ll!. I IIHU-U

2. a flash point and kinemfatic viscosity within limnitsfo

3. a clear and bright appearance with proper color; ICL5.0-56 I (conti nued)

WOG STS Rev 1, 04/07/95 5.0-32 for PI ITSnMarkup Part E

Programs and Manuals 5.5 5.5 Programs and Manuals (continued)

b. Other Pryperti* s for ASTM 2D fuel oil are within limits within 31 days following sami,,,,,ng and, addition to storage tanks; and

c. ,

Tota parti.ulate conentration of the fuel ail is < 10 mlg/i Awen tested every 31 days in acAord3an. with ASIM D-2276, Method A 2 or A3.

The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the IFTA5 .0-67 Diesel Fuel Oil Testing Program test frequencies.

5.5.124 Technical Specifications (TS) Bases Control Program This program provides a means for processing changes to the Bases of these Technical Specifications.

a. Changes to the Bases of the TS shall be made under appropriate administrative controls and reviews.

b. Licensees may make changes to Bases without prior NRC approval L-I1 provided the changes do not requireinvolve either of the following: IR-2 r---

R- II I

1. a change in the TS incorporated in the license; or

2. a change to the updated FSAR or Bases that requires NRC approval pursuant toinvolves an I TA5.0-58 1 unrvie,,. safety question as defined i SR-2 "

I I 10 CFR 50.59.

c. The Bases Control Program shall contain provisions to ensure that the Bases are maintained consistent with the UFSAR.

d. Proposed changes that meet the criteria of Specification 5.5.124 1 CL5.0-56 b above shall be reviewed and approved by the NRC prior to implementation. Changes to the Bases implemented without prior NRC approval shall be provided to the NRC on a frequency CL5.0-56 1 consistent with USAR updates 10- R, 5*.71(e).

WOG STS Rev 1, 04/07/95 5.0-33 Markup for PI ITS Part E

Programs and Manuals 5.5 5.5 Programs and Manuals (continued) 5.5.135 Safety Function Determination Program (SFDP)

This program ensures loss of safety function is detected and appropriate actions taken. Upon entry into LCO 3.0.6, an evaluation shall be made to determine if loss of safety function exists.

Additionally, other appropriate actions may be taken as a result of the support system inoperability and corresponding exception to entering supported system Conditions and Required Actions. This program implements the requirements of LCO 3.0.6. The SFDP shall contain the following:

a. Provisions for cross train checks to ensure a loss of the capability to perform the safety function assumed in the accident analysis does not go undetected;

b. Provisions for ensuring the plant is maintained in a safe condition if a loss of function condition exists;

c. Provisions to ensure that an inoperable supported system's Completion Time is not inappropriately extended as a result of multiple support system inoperabilities; and

d. Other appropriate limitations and remedial or compensatory actions.

A loss of safety function exists when, assuming no concurrent single failure, no concurrent loss of offsite power or no I concurrent loss of onsite diesel generator(s), a safety TA5.0-71 function assumed in the accident analysis cannot be ,-

performed. For the purpose of this program, a loss of safety function !R-2 ,

may exist when a support system is inoperable, and:

a. A required system redundant to the system(s) supported by the inoperable support system is also inoperable; or

b. A required system redundant to the system(s) in turn supported by the inoperable supported system is also inoperable; or WOG STS Rev 1, 04/07/95 5.0-34 Markup for PI ITS Part E

Programs and Manuals 5.5 5.5 Programs and Manuals (continued)

I PA5.0-72

c. A required system redundant to the inoperable support system(s) for the supported systems (a) and (b) above is also inoperable.

loss The SFDP identifies where a loss of safety function exists. If a of safety function is determined to exist by this program, the appropriate Conditions and Required Actions of the LCO in which the loss of safety function exists are required to be entered. When a loss of safety function is caused by the I TA5.-71 inoperability of a single Technical Specification support system, the appropriate Conditions and Required Actions to enter are those of the support system.

5.5.14 Containment Leakage Rate Testing Program

.0-73 i rate CL5

a. A program shall be established to implement the leakage testing of the containment as required by 10 CFR 50.54(o) and 10 CFR 50, Appendix J, Option B, as modified by approved exemptions.

in This program shall be in accordance with the guidelines contained Regulatory Guide 1.163, "Performance-Based Containment Leak-Test Program," dated September 1995.

b. The peak calculated containment internal pressure for the design basis loss of coolant accident is less than the containment internal design pressure, Pa, of 46 psig.

c. The maximum allowable primary containment leakage rate, La, at P,,

shall be 0.25% of primary containment air weight per day. For pipes connected to systems that are in the auxiliary building special ventilation zone, the total leakage shall be less than 0.1% of primary containment air weight per day at pressure P,. For pipes connected to systems that are exterior to I CL!5.0-73 ]

both the shield building and the auxiliary building special be ventilation zone, the total leakage past isolation valves shall less than 0.01% of primary containment air weight per day at 5.0-35 Markup for PI ITS Part E WOG STS Rev 1, 04/07/95

Programs and Manuals 5.5 5.5 Programs and Manuals (continued) pressure Pa" 5.0-73 J I CL5

d. Leakage Rate acceptance criteria are:

1. Primary contaiment leakage rate acceptance criterion is < 1.0 La. Prior to unit startup, following testing in accordance with the program, the combined leakage rate subject acceptance criteria are <_ 0.60 La for all components A tests.

to Type B and Type C tests and _<0.75 La for Type

2. Air lock testing acceptance criteria are:

at a)Overall air lock leakage rate is *< 0.05 La when tested

Ž 46 psig.

_ 0.01 b)For each door intergasket test, leakage rate is La when pressurized to Ž 10 psig.

Leakage

e. The provisions of SR 3.0.3 are applicable to the Containment Rate Testing Program.

to

f. Nothing in these Technical Specifications shall be construed modify the testing Frequencies required by 10 CFR 50, Appendix J. iTA5"0-86 5.5.15 Battery Monitoring and Maintenance Program 125V plant This Program provides for restoration and maintenance of the be used safeguards batteries and service building batteries, which may in instead of the safeguards batteries during shutdown conditions accordance with manufacturer's recommendations, as follows:

V will be

a. Actions to restore battery cells with float voltage < 2.13 in accordance with manufacturer's recommendations, and

b. Actions to equalize and test battery cells that had been discovered with electrolyte level below the minimum established design limit.

1R---R-2L-2I1 L - -- - - -!

5.0-36 Markup for PI ITS Part E WOG STS Rev 1, 04/07/95

Reporting Requirements 5.6 5.0 ADMINISTRATIVE CONTROLS 5.6 Reporting Requirements The following reports shall be submitted in accordance with 10 CFR 50.4.

5.6.1 Occupational R-a&eExposure Report

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

A si gle submittal may be made for the plant a multiple unit s-taton. The submittal should combine sections common to both--a-+-

unit6-at the station.

A tabulation on an annual basis of the number of plantstation, utility, and other personnel (including contractors) for whom CL5.0-56 monitoring was performed, receiving an annual deep dose TA5.0-74 equivalentexpesures > 100 mremjLyr and the4-r associated collective deep dose equivalent (reported in personmnia-rem)expasure according to work and job functions, -e.g., reactor operations and surveillance, inservice inspection, routine maintenance, special maintenance

(-describe maintenance-), waste processing, and refueling3-. This tabulation supplements the requirements of 10 CFR 20.2206. The dose assignments to various duty functions may be estimated based on pocket ionization chamberdosme-i-e, thermoluminescent dosimeter (TLD),

electronic dosimeter, or film badge measurements. Small exposures totalling < 20% of the individual total dose need not be accounted for.

In the aggregate, at least 80% of the total deep whole body-dose equivalent received from external sources should be assigned to specific major work functions. The report covering the previous calendar year shall be submitted by April 30 of each year. {The initial report shall be submitted by April 30 of the year following th-e initial critieality-.]

(conti nued)

WOG STS Rev 1, 04/07/95 5.0-37 Markup for PI ITS Part E

Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.2 Annual Radiological Environmental Monitoring fiperatift-Report

NOTE.

A si gle submittal may be made for the plant a multiple unit stati-on. The submittal should combine sections common to both-a-l-l uni t The Annual Radiological Environmental Monitoring Operating Report L5.0-56 I covering the operation of the plant unit-during the previous U calendar year shall be submitted by May 15 of each year. The report shall include summaries, interpretations, and analyses of trends of the results of the radiological environmental monitoring program for the reporting period. The material provided shall be consistent with the objectives outlined in the Offsite Dose Calculation Manual 5.6.2 Annual Radiological Environmental Monitoring*ptfl-q Report (continued)

CL5.0-56 (ODCM), and in 10 CFR 50, Appendix I, Sections IV.B.2, IV.B.3, and IV.C.

The Annual Radiological Environmental Monitoring Gpeart-ng-Report shall include the results of analyses of all radiological envirn,,ntaI samples and of all enviro..ental radiation measurem,,ents taken, dur*i the period pursuant to the locations speified in the table and fiures ,

"in the 0*,.,*, as well as. summarized and tabulated results, of t-the-se analyses and mneasuremfents [in the format of Regulatory Guide 4.8, December 1975, of all radiological environmental samples taken during the report period the table in the Radilogieal Assessment Braneh Technical Psit*i*n, Revision 1, November 1979]. [The report shall identify the TL- results that represent1 olloated dosimeters in relation to the NRC TILDU program and the eposurc period associated with (conti nued)

WOG STS Rev 1, 04/07/95 5.0-38 Markup for PI ITS Part E

Reporting Requirements 5.6 5.6 Reporting Requirements h result.] In the event that some individual results are not available for inclusion with the report, the report shall be submitted noting and explaining the reasons for the missing results. The missing data shall be submitted in a supplementary report as soon as possible.

The report shall also include the following: a summary description I CL5.0-56 of the radiological environmental monitoring program; a map of sampling locations keyed to a table giving distances and directions from the reactor site; and the results of licensees participation in the Interlaboratory Comparison Program defined in the ODCM.

5.6.3 Radioactive Effluent-Re-l-ease Reoort

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

A si igle submittal may be made for the plant a multiple unit stat -en. The submittal shalls-heu-ld combine sections common to both inits at the station; however, for units with separate radwast.

II PA .0-68 F ---

'iI~ii-'- -

AJII

-aJ E II Q R IQ r-VH..

J. HJ I Lr.

- P-H.

-r.0c A v

I ------.

A I r-Lr LI I iJ* *..',,... I t,.J, I l I L.,* II %.*'* "*,,I I *4*I I I., m The Radioactive Effluent-Release Report covering the operation of the plant during the previous calendar year--uni-t shall be submitted -ft aceerdanee-by May 15 of each year-w-ih FR-50.36a. The report [ CL5.0-56 shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the plant--n-i.

The material provided shall be consistent with the objectives outlined in the ODCM and Process Control Program and in conformance with 10 CFR 50.36a and 10 CFR 50, Appendix I, Section IV.B.1.

(conti nued)

WOG STS Rev 1, 04/07/95 5.0-39 Markup for PI ITS Part E

Reporting Requirements 5.6 5.6 Reporting Requirements (continued) 5.6.4 Monthly Operating Reports I TA5.0-54 I Routine reports of operating statistics and shutdown experience{-

including documentation of all challenges to the pressurizer power operated relief valves or pressurizer safety va.iv shall be submitted on a monthly basis no later than the 15th of each month following the calendar month covered by the report.

5.6.5 CORE OPERATING LIMITS REPORT (COLR)

a. Core operating limits shall be established prior to each reload cycle, or prior to any remaining portion of a reload cycle, and shall be documented in the COLR for the following:

LCO 3.1.1, "SHUTDOWN MARGIN (SDM)";

LCO 3.1.3, "Isothermal Temperature Coefficient (ITC)"; PA5.0-76l LCO 3.1.5, "Shutdown Bank Insertion Limits";

LCO 3.1.6, "Control Bank Insertion Limits";

LCO 3.1.8, "PHYSICS TESTS Exceptions - MODE 2";

LCO 3.2.1, "Heat Flux Hot Channel Factor (FQ(Z))";

LCO 3.2.2, "Nuclear Enthalpy Rise Hot Channel Factor (FAH)";

LCO 3.2.3, "AXIAL FLUX DIFFERENCE (AFD)";

from LCO 3.4.1, "RCS Pressure, Temperature, and Flow - Departure Nucleate Boiling (DNB) Limits"; and LCO 3.9.1, "Boron Concentration".

The must ninviul specifications tha bE!==,ef-ene here.

addr q FArp operating limit J

b. The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, (continued)

WOG STS Rev 1, 04/07/95 5.0-40 Markup for PI ITS Part E

Reporting Requirements 5.6 5.6 Reporting Requirements (continued) specifically those described in the following documents: PA5.0-76

1. NSPNAD-8101-PA, "Qualification of Reactor Physics Methods for Application to PI Units" (latest approved version);

2. NSPNAD-8102-PA, "Prairie Island Nuclear Power Plant Reload Safety Evaluation Methods for Application to PI Units"(latest approved version);

3. NSPNAD-97002-PA, "Northern States Power Company's Steam Line Break Methodology" (latest approved version);

4. WCAP-9272-P-A, "Westinghouse Reload Safety Evaluation Methodology", July, 1985;

5. WCAP-10054-P-A, "Westinghouse Small Break ECCS Evaluation Model using the NOTRUMP Code", August, 1985;

6. WCAP-10924-P-A, "Westinghouse Large Break LOCA Best-Estimate Methodology", December, 1988;

7. WCAP-10924-P-A, Volume 1, Addendum 4, "Westinghouse Large Break LOCA Best Estimate Methodology", August, 1990;

8. XN-NF-77-57 (A), XN-NF-77-57, Supplement 1 (A), "Exxon Nuclear Power Distribution Control for Pressurized Water Reactors Phase II", May, 1981;

9. WCAP-13677, "10 CFR 50.46 Evaluation Model Report: W COBRA/TRAC 2-Loop Upper Plenum Injection Model Update to Support ZIRLOT, Cladding Options", April 1993 (approved by NRC SE dated November 26, 1993);

(continued)

WOG STS Rev 1, 04/07/95 5.0-41 Markup for PI ITS Part E

Reporting Requirements 5.6 5.6 Reporting Requirements

10. NSPNAD-93003-A, "Transient Power Distribution Methodology",

(latest approved version).

c. The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal-mechanical limits, core thermal-hydraulic limits, Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as SDM, transient analysis limits, and accident analysis limits) of the safety analysis are met.

d. The COLR, including any midcycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.

ITA5'0-77 5.6.6 Reactor Coolant System (RCS) PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)

a. RCS pressure and temperature limits for heat-up, cooldown, low temperature operation, criticality, and hydrostatic testing, LTOP arming, PORV lift settings and Safety Injection Pump Disable Temperature as well as heatup and cooldown rates shall be established and documented in the PTLR for the following:

PA5.0-76 LCO 3.4.3, "RCS Pressure and Temperature (P/T) Limits";

LCO 3.4.6, "RCS Loops - MODE 4";

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

LCO 3.4.10, "Pressurizer Safety Valves";

LCO 3.4.12, "Low Temperature Overpressure Protection (LTOP)

Reactor Coolant System Cold Leg Temperature (RCSCLT) - ---

R-12 1 L ------- I (continued)

WOG STS Rev 1, 04/07/95 5.0-42 Markup for PI ITS Part E

Reporting Requirements 5.6 5.6 Reporting Requirements Safety Injection (SI) Pump Disable Temperature";

LCO 3.4.13, "Low Temperature Overpressure Protection (LTOP)

Reactor Coolant System Cold Leg Temperature (RCSCLT) < Safety _I IR -i r- -I Injection (SI) Pump Disable Temperature'; and L ---------

LCO 3.5.3, "ECCS - Shutdown".

[Th, individual specifications that address temperature limits mfust be rcfereneed here-.]-

I CL5.0-56

b. The analytical methods used to determine the RCS pressure and temperature limits and Cold Overpressure Mitigation System setpoints shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:

WCAP-14040-NP-A, Revision 2, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves" (includes any exemption granted by NRC to ASME Code Case N-514).

E[ldntify the NRC staff approval document y . CL5.0-56

c. The PTLR shall be provided to the NRC upon issuance for each if-ctor vessel fluence period and for any revision or supplement 4hereto. Changes to the curves, setpoints, or parameters in the PTLR resulting from new or additional analysis of beltline material properties shall be submitted to the NRC prior to issuance of an updated PTLR.

i P~ iQN p 1 15 :- The methodology for the calculation of the P T 3 Li IP.76 is3*i5 ;:

i4-fti-jts for NIRC approval should in lude the toi..owln. pov I. he mneth logyshall describe how the neutron fluence is ealculat"d

A

,rM P

-A,,,

~ H II P, AF ZI- Ai U

- vn r, .*

VHI IC4F.r- I I

Fy~l*

l i 1.)J. U %

!ý :1h.- QAAP*Ar vessel mater4ai Surve i l IanIe Program shall co1me; ply wIi th (continued)

WOG STS R~v 1, 04/07/95 5.0-43 Markup for PI ITS Part E

Reporting Requirements 5.6 5.6 Reporting Requirements W R.,. - .f. Requirements 5.6. IH" :F Ieotn ;

r .. ; Hi: 'N'. n'. .r +Frr I

r 'c - :ssel material irradiation I PA5.0-76

'W'-

kttl.Aý .

I I Surveillance specimen removal schedule shall be provided, along with how the specimen examinations shall be used to update th PTLPX curve-s-.

3. Low Tefm-perature uverpressure -rotection .L...) vsze*

m it setting II I L*

I111 I V1 LI1.. I U/V/l UVF kL I C LL, %-U IXL. I I i, I V" I V I.. ,.). Il \

,Jl J -,A L.. I LJpLLJ L.. IuI

- .-LIL J J-1 -- -.. .... L- ;---12...J--J -- .tL.. lIn i n I I1 X LHJ--uI V IId LII UCAA U I U .)I"I%-I , AL I I1L C/ -LU III 'lIiL I Iýl\,

A 1 - - J-!..-- --- J r -- - L -.-.- .- -*. L ^- .I-'I!

iC III Clujl U3 LCU I U i F ICL- dikIId LL LLAIU I I-%1\IJ I UI UI L I L II CA I LLU L Illi-I lIlTiQaCU I I L-l CIL-oUIL CU C1L, ULL*I- I,[ (U IAU lIUklUII

/

I - *i - -- - -- - - - - - - - - --.... tI*1 I X .. 1 --L .... #'.!- A-.. l r _.;

n1 CHI/IIL~lILIL~ inI aCALL-C/I U CAILIt U IH CyUIULUIJY AL, ,*

ICC .. J I\C;V IiIUII L,.

I *

i I 'I ,1 I

r" .I-rý - 4 4imMn- i4 R -

nr1 nrnnrrR4F rcin:AF inrFR. FR-r PR4I II-Ir12fl 2Rm

,4 4-L NIT rc- npon~r\

I l AIq -L.11P.-1 LL114IL 11-V A PLLI V- R. API'- -

SLt d arII d ReX-vii Il :..0.7. , I ,Ci CiLiiiIT*I ULUIC Lilmli Ls.

r r C nI.- , .. Ilt i ý ft - --- ,(nrc- nnri-ciirr ARMl -rrlnrn iinr I T4TTTC nrnnrn CA. U

U l\ (ACkL- LLII kL*U I luI L 0AV. L.L%.III Ik L,..) I I\LC ,.AUIXL- I'LJ ILl LIXCI IL.

I, L-iIl.

- - ILl 'IAI f6-.- he m,,inimum temperature requirements of Appendix G to 10 CFR Part 50 ihall be incorporated into the pressure and temperature limnit 7.cPensees who have remolved two or more capsules should compare for 4 ach surveillance material the measured e in referen Iemper; ure-.R.. .. ) to the predicted.

.T,, w-hre the predicted increase in RTon mean

£ -hifin-,,*T plus the two standard dcviati*,n value (2a-s-pecified

~~~~IXL~~~~~~~~~U~~

A 3 Iy XVI3 /l C. II LIC~ ILJA X UICL - CLIJ the predicted value (increase RT _2 + J..., the licensee shou pir V i I), inpAl l, IIIIeII LUo liL I LX LU L demon UIl. I tratI L HOUII tLIIL eU I

-r- *- - - J-----------------------J_ ...

QICI L-L L*l.lu l. I lVC-U IIICLIIUUC1IUýJy.

(continued)

WOG STS Rev 1, 04/07/95 5.0-44 Markup for PI ITS Part E

Reporting Requirements 5.6 5.6 Reporting Requirements 5.6."7 [*r Failure-Report 1f- n individua or fore emergency diesel generator ([DG) exerenes four valid failures in the last 25 demands, these failures and I cLJ.0-81 I

,,ynonvalid failures experienedby n t period sh71Wbe re-ported within 30 days. Reports-on LOG Guide failures shall

.9,

-- nelc e the +n,,,flton o eorlne roc dIn Regulat....

i*.*,,,,,,

n ai Fn the ifm ude

! Qo Revision- 3, egu iat .. ry.. V U 1-El.l U ! CT Z 3lllU I UL reporting requiremnent-.

5.6.8 PAEM Report When a report is required by Condition BC or GJ of LCO 3.3.f3r, "P-o-s+Event A*eident Monitoring (PAEM) Instrumentation," a report shall be submitted within the following 14 days. The report shall outline the preplanned alternate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels of the Function to OPERABLE status. I R 1R7I I

- - - I i-f C fr)f~nA" '<"'"-'- r" l-(dim ~~(-,lk-L ~ \,.,.V ~ . ~ ~ 4 41 +*

I A5.61 CýWWA=WIAP RP:RRPT-RF+

[I1Y il~

-- ~~~

L~t~

L1..4'.J I

~Ii k-1Q I k,,~..

LJ 4-1.- n fl 4-'--'n fl -'~' r-n A-:Fp P nn-t-inm~rnt (2(21 "C' -I21I, "

T ....

I IA Jul vl 1 l1IlI..L IrUhl g a ul J~~ I beiC IpC I bed to the NRC within 30 days. The report shall include a description of the tendon condition, the condition of the oncrete (especiall, at tendon anchorages), the inspection procedures, the tolerances on cracking, and the corrective action taken-.

(continued)

WOG STS Rev 1, 04/07/95 5.0-45 Markup for PI ITS Part E

Reporting Requirements 5.6 5.6 Reporting Requirements (continued) 5.6.71- team Generator Tube Inspection inspector Report CL5.0-83

1. Following each in-service inspection of steam

_ generator tubes, if there are any tubes requiring plugging or__

sleeving, the number of tubes plugged or sleeved in each steam generator shall be reported to the Commission within 15 days.

2. The results of steam generator tube in-service inspections shall be included with the summary reports of ASME Code Section XI inspections submitted within 90 days of the end of each refueling outage. Results of steam generator tube in service inspections not associated with a refueling outage shall be submitted within 90 days of the completion of the inspection. These reports shall include: (1) number and extent of tubes inspected, (2) location and percent of wall-thickness penetration for each indication of an imperfection, and (3) identification of tubes plugged or sleeved.

3. Results of steam generator tube inspections which fall into Category C-3 require notification to the Commission prior to resumption of plant operation, and reporting as a special report to the Commission within 30 days. This special report shall provide a description of investigations conducted to determine cause of the tube degradation and corrective measures taken to prevent recurrence.

4. The results of inspections performed under Specification 5.5.8.b for all tubes that have defects below the F* or EF* distance, and were not plugged, shall be reported to the Commission within 15 days following the inspection. The report shall rI~R-12, or T TS ar L- ....

WOGSTSRe 1,0407/5 .0-6 arkp WOG STS Rev 1, 04/07/95 5.0-46 Markup for PI ITS Part E

Reporting Requirements 5.6 5.6 Reporting Requirements (continued) include:

a. Identification of F* and EF* tubes, and

b. Location and extent of degradation.

5. For implementation of the voltage-based repair criteria to tube support plate intersections, notify the NRC staff prior to returning the steam generators to service should any of the following conditions arise:

a. If estimated leakage based on the projected end-of-cycle (or if not practical, using the actual measured end of-cycle) voltage distribution exceeds the leak limit (determined from the licensing basis dose calculation for the postulated main steamline break) for the next operating cycle.

b. If circumferential crack-like indications are detected at the tube support plate intersections.

C. If indications are identified that extend beyond the confines of the tube support plate.

d. If indications are identified at the tube support plate elevations that are attributable to primary water stress corrosion cracking.

e. If the calculated conditional burst probability based on the projected end of-cycle (or if not practical, using the actual measured end-of-cycle) voltage distribution exceeds 1 x 10-2, notify the NRC and provide an assessment of the safety significance of the occurrence. r........

1 R-12 L -------

WOG STS Rev 1, 04/07/95 5.0-47 Markup for PI ITS Part E

Reporting Requirements 5.6 5.6 Reporting Requirements (continued)

Reviewer's Note: Reports required by the Licensec 's currcnt licens-ing basis regarding stcam generator tube surveillance requiremfents shall be included herc. An appropriate admfinistrativc controls formfat should be used.

Rcviev or's Note: These reports mfay be requircd covering -inspection,

-te-s-t,--and m-faintenance activitics. Thcse reports arc determfined on an in-d-i-vdual basis for each unit and their preparation and submfittal are dei~ ated in the Technical Specifications.

WOG STS Rev 1, 04/07/95 5.0-48 Markup for PI ITS Part E

fHigh Radiation Area]

f5.71 TA5.0-54 5.0 ADMINISTRATIVE CONTROLS f5.7 High Radiation Area provided in Pursuant

.7.As 10 CFR 20- paragraph 20.1601(c) of 10 CFR Part 20, the following controls shall be applied in place of the controls required by paragraphin lieu of the requirements of 10 CFR 20.1601(a) and (b) of, each high radiation area, as defined in 10 CFR 20:

5.7.1 High Radiation Areas accessible to personnel in which radiation levels could result in an individual receiving a deep dose equivalent less than 1.0 rem in one hour at 30 centimeters from the radiation in w,-iet, source or from any surface that the radiation penetrates, the intensit,,,y of radiation is :100mrem/hr &bu 1000 m-rem"hr,

a. Each entryway to such an area shall be barricaded and conspicuousl posted as a high radiation area. Such barricades may be opened as necessary to permit entry or exit of personnel or equipment.

b. Access to, and activities in each such area and entrance thereto shall be controlled by means requiring issuance of a Radiation Wolk Permit (RWP) or equivalent that includes specification of radiation dose rates in the immediate work area(s) and other appropriate radiation protection equipment and measures.

c. Individuals qualified in radiation protection procedures and (e.g., [leal',h hysics Technicians]) or personnel continuously escorted by such individuals may be exempted from the RWP issuance requirement for an RWP or equivalent while performing during the perforfflane of their assigned duties in high radiation areas with exposure rates !1000 ,re/hr, provided they are otherwise following plant radiation protection procedures for entry to, exit from,d. Eah An-indvidul and orgrou work in-into entringofndivieals.

such high -radiationareas. pritdt

d. Each Afty-individual or group entering of i,,diviuas permfitted to, enter-such an areas- shall possess be provided with or accompanied (continujd)

I WOG STS Rev 1, 04/07/95 5.0-49 Markup for PI ITS Part E

{High Radiation Area]

f5.71 by one or more of the following:

1a. A radiation monitoring device that continuously displays indicates the radiation dose rates in the area; or_

2b. A radiation monitoring device that continuously integrates the radiation dose rates in the area and alarms when the device's a preset integrated dose alarm setpoint is reached, with an appropriate alarm setpoint; or is rece;ived. [,,nrY into such areas with this monitorlilng ddevice~mjnay be made after.

the dose rate levels in the area have been established and personnel are aware of ther.

3. A radiation monitoring device that continuously transmits dose rate and cumulative dose information to a remote receiver monitored by radiation protection personnel responsible for controlling personnel radiation exposure within the area; or

4. A self-reading dosimeter (e.g., pocket ionization chamber or electronic dosimeter) and, (i) Be under the surveillance, as specified in the RWP or equivalent, while in the area, of an individual qualified in radiation protection procedures, equipped with a radiation monitoring device that continuously displays radiation dose rates in the area, who is responsible for controlling personnel exposure within the area; or (ii) Be under the surveillance as specified in the RWP or equivalent, while in the area, by means of closed circuit television, of personnel qualified in radiation protection procedures, responsible for controlling personnel radiation exposure in the area, and with the means to communicate with individuals in the area who are covered by such surveillance.

(continued)

WOG STS Rev 1, 04/07/95 5.0-50 Markup for PI ITS Part E

{High Radiation Area]

f5.71

e. Except for individuals qualified in radiation protection procedures, or personnel continuously escorted by such TA5"0-54 individuals, entry into such areas shall be made only after dose rates in the area have been determined and entry personnel are knowledgeable of them. These continuously escorted personnel will receive a pre-job briefing prior to entry into such areas.

This dose rate determination, knowledge, and pre-job briefing does not require documentation prior to initial entry.

e. An individual qualified in radiation protectin*

arcedurs with a radiation dose rate monitoring device, "s who respons-ibl for providing positive control over the activities within the area and shall perform periodic radia-tion surveillance at thiefeunc specified by the [radiation Protection Manger i t RWP.

levels 5.7.2 High Radiation Areas accessible to personnel in which radiation could result in an individual receiving a deep dose equivalent in excess or of 1.0 rem in one hour at 30 centimeters from the radiation source rad in from any surface that the radiation penetrates, but less than 500 one hour at one meter from the source In addition to the requtHr*ments *of Specification 5.7.1, areas with radiation levels ý, 1000mrem/+r

a. Each entryway to such an area shall be conspicuously posted as ITA5.0-547 a high radiation area and shall be provided with a locked or continuously guarded door& or gate that +/-o-prevents unauthorized entry, and, in addition:

1. All such door and gate-the keys shall be maintained under .0-84 ý the administrative control of the sShift supervisor, [PA5 radiation protection manager, or their designee-f-em on duty or health physics supervision.

- 2. Doors and gates shall remain locked except during periods of aeee*s--by-personnel or equipment entry or exit.d-er a approved RWP that shall speci fy the dosc rate level s in

.7.2 (continued)

b. Access to, and activities in, each such area shall be controlled by means of an RWP or equivalent that includes specification of radiation dose rates in the immediate work area(s) and other appropriate radiation protection equipment and measures.

5.0-51 Markup for PI ITS Part E WOG STS Rev 1, 04/07/95

{High Radiation Area]

E5.71

c. Individuals qualified in radiation protection procedures may be exempted from the requirement for an RWP or equivalent while TA5"0-54 performing radiation surveys in such areas provided they are otherwise following plant radiation protection procedures for entry to, exit from, and work in such areas.

d. Each individual or group entering such an area shall possess:

1. A radiation monitoring device that continuously integrates the radiation dose rates in the area and alarms when the device's dose alarm setpoint is reached, with an appropriate alarm setpoint; or

2. A radiation monitoring device that continuously transmits dose rate and cumulative dose information to a remote receiver monitored by radiation protection personnel responsible for controlling personnel radiation exposure within the area with the means to communicate with and control every individual in the area; or

3. A self-reading dosimeter (e.g., pocket ionization chamber or electronic dosimeter) and, (i) Be under the surveillance, as specified in the RWP or equivalent, while in the area, of an individual qualified in radiation protection procedures, equipped with a radiation monitoring device that continuously displays radiation dose rates in the area, who is responsible for controlling personnel exposure within the area; or (ii) Be under the surveillance as specified in the RWP or equivalent, while in the area, by means of closed circuit television, of personnel qualified in radiation protection procedures, responsible for controlling personnel radiation exposure in the area, and with the means to communicate with and control every individual in the area.

4. In those cases where options (2) and (3), above are impractical or determined to be inconsistent with the "As Low As is Reasonably Achievable" principle, a radiation WOG STS Rev 1, 04/07/95 5.0-52 Markup for PI ITS Part E

{high Radiation Area]

f5 .71 displays monitoring device shall be used that continuously radiation dose rates in the area.

e. Except for individuals qualified in radiation protectionindividuals, such procedures, or personnel continuously escorted by rates in the entry into such areas shall be made only after dose are knowledgeable of area have been determined and entry personnel receive a pre-job them. These continuously escorted personnel will dose rate briefing prior to entry into such areas. This does not require determination, knowledge, and pre-job briefing documentation prior to initial entry.

stay times for the im,,,ediate work areas and the maximum allowable tf individuals in thse areas In lieu of the stay as closed circuit TV caeras of the RU,, ir or rfmote (**uh to.p...ide positive exposure control over radiat pr. du.

the activitiese-in performed itn te area.

radiation levels of 5.7.3 f. Such Fo-r--individual hig-h-radýationareas with a

> lo mr,,/,,, accessible to prsonel that are located within no enclosure exists larger area-. such as reactr ontinment, where for the purposes- of locking, or that cano b continuously where no enclosure can be Sdand reasonably constructed need not be around the individual area, that individual area guarded, but controlled by a locked door or gate, nor continuously flashing light and a shall be barricaded, ornd-conspicuously posted, shall be activated at the area as a warning device.

TABLE 5.5.8-1 5.0-53 Markup for PI ITS Part E WOG STS Rev 1, 04/07/95

TABLE 5.5.8-1 STEAM GENERATOR TUBE INSPECTION 2ND SAMPLE INSPECTION 3RD SAMPLE INSPECTION 1ST SAMPLE INSPECTION Result Action Required Result Action Required Result Action Sample Required Size C-1 None N/A N/A N/A N/A A minimum of S Tubes C-2 Repair defective C-1 None N/A N/A per S.G. tubes and inspect C-2 Repair defective tubes additional 2S tubes and inspect additional in this S.G. 4S tubes in this S.G.

C-3 Perform action for C-3 result C-1 None of first sample C-2 Repair defective tubes C-3 Perform action for C-3 result of first sample C-3 Inspect all tubes in this All other S.G.s are None N/A N/A S.G., Repair defective C-1 tubes and inspect 2S tubes in each other S.G. N/A N/A Some S.G.s C-2 Perform action for C-2 but no result of second sample NRC. additional S.G.

are C-3 Additional S.G. Inspect all tubes in N/A N/A is C-3 each S.G. and repair defective tubes.

Prompt notification to NRC.

S=3%; When two steam generators are inspected during that outage. II S=6%; When one steam generator is inspected during that outage. ' R-12 L --------.

WOG STS Rev 1, 04/07/95 5.0-54 Markup for PI ITS Part E

TABLh j.5.8-2 Steam Generator Tube Sleeve Inspection 15t Sample Inspection 2 nd Sample Inspection Sample Size Result Action Required Result Action Required None A minimum of 20% of C-1 N/A N/A Tube Sleeves (1)

C-2 C-l None Inspect all remaining tube sleeves in this S.G. and plug or repair defective sleeved tubes.

C-2 Plug or repair defective sleeved tubes C-3 Perform action for C-3 result of first sample C-3 Inspect all tube sleeves in The other None this S.G., inspect 20% of the S.G. is C-1 tube sleeves in the other S.G.,

and plug or repair defective sleeved tubes The other Perform action for C-2 S.G. is C-2 results of first sample The other Inspect all tube sleeves S.G. is C-3 in each S.G. and plug or repair defective sleeved tubes (1) Eacn type or sleeve is cons-lcered a separate populaioun, -LUL aLt oLnLoL'J scope epa o Ri1i 1

R-12 I L------

WOG STS Rev 1, 04/07/95 5.0-55 Markup for PI ITS Part E

Part F Package 5.0 P rt F Difference Difference Category Number Justification for Differences 5.0-TA 63 This change incorporates TSTF-279.

CL 64 The CTS SG program requirements are provided as required by the Reviewer's Note in NUREG-1431. The CTS requirements from 4.12.A through D have been included in ITS.

65 Not used.

CL 66 In conformance with the guidance of NUREG-1431, program definition for the VFTP is provided. The format and contents of the Program requirements have been changed to incorporate CTS requirements for these systems and incorporate the requirements of NRC Generic Letter 99-02.

TA 67 This change incorporates TSTF-1 18.

PA 68 The Note in brackets has been modified to correctly apply to Pl.

PA 69 A new test interval of "Semiquarterly" has been included to allow accelerated testing of equipment that fails a quarterly test as required by the ASME test program.

70 Not used.

Prairie Island Units 1 and 2 4 4/1/02

Part F Package 5.0 Part F Package 5.0 Difference Difference Category Number Justification for Differences 5.0-80 Not used.

CL 81 CTS do not require this report; therefore it is not included in the ITS. This change is also consistent with approved TSTF-37, Revision 2.

82 Not used.

CL 83 The CTS report requirements are provided in the ITS as required by the Reviewer's Notes in NUREG-1431.

The CTS requirements from 4.12.E have been included in ITS.

PA 84 The titles of ITS 5.7.1 and 5.7.2 have been revised to be consistent with the guidance of Regulatory Guide 8.38. This change is beneficial in that overall it may reduce plant radiation exposure.

Prairie Island Units 1 and 2 6 4/1/02

Part G PACKAGE 5.0 ADMINISTRATIVE CONTROLS NO SIGNIFICANT HA7ARDS DETERMINATION AND ENVIRONMENTAL ASSESSMENT NO SIGNIFICANT HAZARDS DETERMINATION The proposed changes to the Operating License have been evaluated to determine whether they constitute a significant hazards consideration as required by 10CFR Part 50, Section 50.91 using the standards provided in Section 50.92.

For ease of review, the changes are evaluated in groupings according to the type of change involved. A single generic evaluation may suffice for some of the changes while others may require specific evaluation in which case the appropriate reference change numbers are provided.

A - Administrative (GENERIC NSHD)

(A5.0-00, A5.0-04, A5.0-06, A5.0-07, A5.0-11, A5.0-12, A5.0-13, A5.0-14, A5.0-16, A5.0-24, A5.0-26, A5.0-27, A5.0-28, A5.0-31, A5.0-32, A5.0-33, A5.0-34, A5.0-36, A5.0-38)

Most administrative changes have not been marked-up in the Current Technical Specifications, and may not be specifically referenced to a discussion of change. This No Significant Hazards Determination (NSHD) may be referenced in a discussion of change by the prefix "A" if the change is not obviously an administrative change and requires an explanation.

These proposed changes are editorial in nature. They involve reformatting, renaming, renumbering, or rewording of existing Technical Specifications to provide consistency with NUREG-1431 or conformance with the Writer's Guide, or change of current plant terminology to conform to NUREG-1431. Some administrative changes involve relocation of requirements within the Technical Specifications without affecting their technical content. Clarifications within the new Prairie Island Improved Technical Specifications which do not impose new requirements on plant operation are also considered administrative.

Prairie Island Units 1 and 2 1 4/1/02

P~rtG Packaae 5.0 LR - Less restrictive, Relocated details (GENERIC NSHD)

(LR5.0-01, LR5.0-02, LR5.0-03, LR5.0-05, LR5.0-22)

Some information in the Prairie Island Current Technical Specifications that is descriptive in nature regarding the equipment, system(s), actions or surveillances identified by the specification has been removed from the proposed specification and relocated to the proposed Bases, Updated Safety Analysis Report or licensee controlled procedures. The relocation of this descriptive information to the Bases of the Improved Technical Specifications, Updated Safety Analysis Report or licensee controlled procedures is acceptable because these documents will be controlled by the Improved Technical Specifications required programs, procedures or 10CFR50.59.

Therefore, the descriptive information that has been moved continues to be maintained in an appropriately controlled manner.

1. The proposed amendment will not involve a significant increase in the probability or consequences of an accident previously evaluated, The proposed changes relocate detailed, descriptive requirements from the Technical Specifications to the Bases, Updated Safety Analysis Report or licensee controlled procedures. These documents containing the relocated requirements will be maintained under the provisions of 10CFR50.59, a program or procedure based on 10CFR50.59 evaluation of changes, or NRC approved methodologies.

Since these documents to which the Technical Specifications requirements have been relocated are evaluated under 10CFR50.59 or its guidance, or in accordance with NRC approved methodologies, no increase in the probability or consequences of an accident previously evaluate will be allowed without prior NRC approval.

Therefore, these changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. The proposed amendment will not create the possibility of a new or different kind of accident from any accident previously analyzed, These proposed changes do not necessitate physical alteration of the plant, that is, no new or different type of equipment will be installed, or change parameters governing normal plant operation. The proposed changes will not impose any different requirements and adequate control of the information will be maintained.

Thus, these changes do not create the possibility of a new or different kind of accident from any accident previously evaluated.

Prairie Island Units 1 and 2 7 4/1/02

Current Technical Specification Cross-Reference CTS Table CTS Section Section Type ITS Section ITS Table Item Number Item Number Table 4.1-1C 11 SR 3.3.4.1 Table 4.1-1C 11 SR 3.3.4.2 Table 4.1-1C 12 Deleted - Boric Acid LAR Table 4.1-1C 13 Relocated TRM Table 4.1-1C 14 CTS Deleted Table 4.1-1C 15 TABLE 3.3.1-1 16.b.2 Table 4.1-1C 15 Relocated TRM Table 4.1-1C 16 Relocated TRM Table 4.1-1C 17 Relocated TRM Table 4.1-1C 18 SR 3.3.1.12 Table 4.1-1C 19 Relocated TRM Table 4.1-1C 20 Relocated TRM Table 4.1-1C 21 SR 3.3.3.1 Table 4.1-1C 21 SR 3.3.3.2 Table 4.1-1C 21 SR 3.3.3.3 Table 4.1-1C 22 CTS Deleted Table 4.1-1C 23 CTS Deleted Table 4.1-1C 24 Relocated TRM Prairie Island Units 1 and 2 Table - 18 4/1/02

Current Technical Specification Cross-Reference CTS Table CTS Section Section Type ITS Section ITS Table Item Number Item Number Table 4.1-1C 24 SR 3.3.6.5 Table 4.1-1C 24 SR 3.3.6.2 Table 4.1-1C 25 SR 3.4.12.4 Table 4.1-1C 25 SR 3.4.12.5 Table 4.1-1C 25 SR 3.4.13.5 Table 4.1-1C 25 SR 3.4.13.6 Table 4.1-1C 26 Relocated TRM Table 4.1-1C 27 Relocated TRM Table 4.1-1C 28 Relocated TRM Table 4.1-1C 29 SR 3.3.3.1 Table 4.1-1C 29 SR 3.3.3.2 Table 4.1-1C 29 (Partial) Relocated TRM Table 4.1-1C 30 Relocated Bases Table 4.1-1C 31 Relocated TRM Table 4.1-1C Note 30 SR 3.1.7.1 Table 4.1-1C Note 31 Deleted Table 4.1-1C Note 32 Relocated TRM Prairie Island Units 1 and 2 Table - 19 4/1/02

ML021060604

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