ML051440554
| ML051440554 | |
| Person / Time | |
|---|---|
| Site: | Cook  |
| Issue date: | 04/15/2005 |
| From: | Indiana Michigan Power Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| Download: ML051440554 (492) | |
Text
{{#Wiki_filter:Attachment 1, Volume 14, Rev. 1, Page i of i
SUMMARY
OF CHANGES ITS SECTION 3.9 Change Description Affected Pages A self-identified change for ITS 3.9.2 has been made. CTS Pages 26, 27, 28, 29, 30, 31, 32, Amendments 283 (Unit 1) and 267 (Unit 2) have been incorporated 33, and 36 of 188. into the ITS submittal. This CTS change deleted CTS 4.9.2 CHANNEL FUNCTIONAL TEST requirements, modified the CTS 4.9.2 CHANNEL CHECK requirement (new CTS 4.9.2.a), and added a CHANNEL CALIBRATION requirement (new CTS 4.9.2.b). This change does not affect the ITS. However, since the CTS 4.9.2.b Frequency is 18 months, and the proposed ITS SR 3.9.2.2 Frequency is 24 months, a new ITS 3.9.2 DOC L.4 has been provided, adding this item to the scope of Beyond Scope Issue 21. A self-identified change for ITS 3.9.2 has been made. This change Pages 35 and 37 of 188. revises ITS 3.9.2 and Required Action A.2 to be consistent with the wording in TSTF-286 as used in other ITS sections. A self-identified change for ITS 3.9.2 Bases has been made. This Page 41 of 188. change revises the ITS 3.9.2 Bases LCO Section to clarify that audible count rate in the control room is required. The change described in the response to Question 200406041439 Page 90 of 188. for ITS 3.9.4 has been made. This change revises ITS 3.9.4 LCO Note and Required Action A.1 to be consistent with the wording in TSTF-286. The change described in the response to Question 200406041441 Page 115 of 188. for ITS 3.9.5 has been made. This change revises ITS 3.9.5 Required Action B.1 to be consistent with the wording in TSTF-286. Attachment 1, Volume 14, Rev. 1, Page i of i
Attachment 1, Volume 14, Rev. 1, Page 1 of 188 VOLUME 14 CNP UNITS 1 AND 2 IMPROVED TECHNICAL SPECIFICATIONS CONVERSION ITS SECTION 3.9 REFUELING OPERATIONS Revision 1 Attachment 1, Volume 14, Rev. 1, Page 1 of 188
Attachment 1, Volume 14, Rev. 1, Page 2 of 188 LIST OF ATTACHMENTS
- 1. ITS 3.9.1
- 2. ITS 3.9.2
- 3. ITS 3.9.3
- 4. ITS 3.9.4
- 5. ITS 3.9.5
- 6. ITS 3.9.6
- 7. Relocated/Deleted Current Technical Specifications (CTS)
- 8. Improved Standard Technical Specifications (ISTS) not adopted in the CNP ITS Attachment 1, Volume 14, Rev. 1, Page 2 of 188
, Volume 14, Rev. 1, Page 3 of 188 ATTACHMENT I ITS 3.9.1, Boron Concentration , Volume 14, Rev. 1, Page 3 of 188 , Volume 14, Rev. 1, Page 4 of 188 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) , Volume 14, Rev. 1, Page 4 of 188
Attachment 1, Volume 14, Rev. 1, Page 5 of 188 ITS 3.9.1 ITS Y4 LUMXlNG COMNDMONS FOR OPERATION AND SURVMILIANCE REQUIREMNS 3W4.9 REFUEMLNG OPTRAONS
- f IMmmi im E'nmnmc nR nPF MATIAand the refueinhg:ca:vy::
LCO 3.9.1 3.9.1 The boron concentration of all fieafortions oflthe Reactor Coolant Systemt[i the refueling cl+J haIl be maintained unifomranrw icidea to ensure at thore re[stricti of the 1olovwng i ctivity condition3re//// L Eilerafrof 9 or)6 w iefncludes(1 &ktlk conservati nowance for un4inties, or b.
'd MYn boronco n MOE64ihntelm ofgerthy or equal to 2
/ethir dhelmiosecfiordunIertai ppm which incl
/
50 ppm th secitied Inthe COLR_
/A1 APEUCA proposed Applicability LAdd Note SVifi~th the requirerns of the above specific tion not satisfied. I) innedi tely suspend all ~-0 ACTION A th n reIt weev~ntheRWS, toviedthe boron,&ncentaon in tW RWST is "ter p~an the rn~in rquied b fcitktia 3.1.2.7,10. andl2) -iradate aa continue boration asiater 4Mno,,50pmbrccdouo or its o"qatnit tmti I irdued t Icstt,* O.Ho r qusth broneoc,#raon s csor ouato2400
-0a
}-e
*-0 greater th n SIMILAN- U,= Tu2l'm U W U1UT~I-4.9.1.1 Tb tvttive st of the above irntvity conditions sh l be dnned prior to:
,>/ ernvinoruboltngtc~cor~tsthead, nW
/b. Withdrawal of an fD control rod in excess of3 from its fully insened position.
SR 3.9.1.1 4.9.1.2 The boron concentration of the reactor coolant system and the refueling canal shall be determinedI SR 3.9.1.2 I cb- !!f!i! at leat onc per 72 hours.
>eal COOK NUCLEAR PLANT.UNrr I PageY4 9-1 AMENDMENT I,t16,30, 243 Page 1 of 2 Attachment 1, Volume 14, Rev. 1, Page 5 of 188
Attachment 1, Volume 14, Rev. 1, Page 6 of 188 ITS 3.9.1 ITS JIM ILJCr OPRAnIWWSFMO_ AND__U__ _ _A_ _ _ __IR _ _ _ LCO 3.9.1 3.9.1 The b4M Io . APPFlUr~An*n! ACTION A a. h Batnn COR ALdW or OF Nonbf
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eF 4.9.1. Tho muicdve of the Ahotwo rA--f'*7 coitc hflb cw x
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SR 3.9.1.1 49.1.2 1M bees ofdoU. ad th _W s___be A SR 3.9.1.2 . [ . ~ ~ slW~a~ pw72 hems COOX NUCLZAR PLANTM P*eY34 94 AMENDMENT94,0213 Page 2 of 2 Attachment 1, Volume 14, Rev. 1, Page 6 of 188
Attachment 1, Volume 14, Rev. 1, Page 7 of 188 DISCUSSION OF CHANGES ITS 3.9.1, BORON CONCENTRATION ADMINISTRATIVE CHANGES A.1 In the conversion of the CNP Current Technical Specifications (CTS) to the plant specific Improved Technical Specifications (ITS), certain changes (wording preferences, editorial changes, reformatting, revised numbering, etc.) are made to obtain consistency with NUREG-1431, Rev. 2, "Standard Technical Specifications-Westinghouse Plants" (ISTS). These changes are designated as administrative changes and are acceptable because they do not result in technical changes to the CTS. A.2 CTS 3.9.1 provides requirements on the boron concentration of all filled portions of the Reactor Coolant System and the refueling canal. ITS 3.9.1 provides requirements on the boron concentration of the Reactor Coolant System, the refueling canal, and the refueling cavity. This changes the CTS by explicitly including the refueling cavity in the volumes required to have boron concentration maintained. This change is acceptable because the technical requirements have not changed. The refueling cavity is considered to be governed by the CTS requirements because the refueling cavity is typically connected to the RCS, the refueling canal, or both. This change is designated as administrative because the technical requirements of the specifications have not changed. A.3 CTS 3.9.1 Action b contains the statement, "The provisions of Specification 3.0.3 are not applicable." ITS 3.9.1 does not contain an equivalent statement. This changes the CTS by deleting the Specification 3.0.3 exception. This change is acceptable because the technical requirements have not changed. ITS LCO 3.0.3 is not applicable in MODE 6. Therefore, the CTS LCO 3.0.3 exception is not needed. This change is designated as administrative because it does not result in a technical change to the CTS. MORE RESTRICTIVE CHANGES M.1 CTS 3.9.1 Action a requires the immediate suspension of positive reactivity changes "except addition of water from the RWST, provided the boron concentration in the RWST is greater than the minimum required by Specification 3.1.2.7.b.2" (i.e., 2400 ppm). ITS 3.9.1 Required Action A.2 requires positive reactivity additions to be suspended, but does not provide any allowance for positive reactivity changes due to the addition of water from the RWST to continue. This changes the CTS by removing the allowance to allow a positive reactivity change from the addition of water from the RWST, provided the boron concentration of the RWST is greater than 2400 ppm. The purpose of CTS 3.9.1 Action a is to provide assurance that an inadvertent criticality will not result when the boron concentration is not within limits in MODE 6. The CTS 3.9.1 Action requires the suspension of all operations involving CORE ALTERATIONS or positive reactivity changes and initiation of activities to restore boron concentration to within its limit. However, allowing a CNP Units 1 and 2 Page 1 of 4 Attachment 1, Volume 14, Rev. 1, Page 7 of 188
Attachment 1, Volume 14, Rev. 1, Page 8 of 188 DISCUSSION OF CHANGES ITS 3.9.1, BORON CONCENTRATION positive reactivity addition conflicts with the requirement to restore boron concentration to its limit. Therefore, this exception is deleted. This change is acceptable because the ITS requires actions that provide assurance that an inadvertent criticality will not result while boron concentration is not within limits in MODE 6, and requires initiation of activities to restore boron concentration to within its limit. This change is designated as more restrictive because it provides more restrictive corrective actions in the ITS than in the CTS. RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES LA.A (Type 5 - Removal of Cycle-Specific Parameter Limits from the Technical Specifications to the Core Operating Limits Report) CTS 3.9.1 states that the boron concentration in MODE 6 shall be the more restrictive reactivity condition of a kff of 0.95 or less or a boron concentration of > 2400 ppm. ITS LCO 3.9.1 states that the boron concentration shall be within the limit specified in the COLR. This changes the CTS by relocating the MODE 6 boron concentration limit, which must be confirmed on a cycle-specific basis, to the CORE OPERATING LIMITS REPORT (COLR). The removal of these cycle-specific parameter limits from the Technical Specifications and their relocation into the COLR is acceptable because these limits are developed or utilized under NRC-approved methodologies. The NRC documented in Generic Letter 88-16, "Removal of Cycle-Specific Parameter Limits From Technical Specifications," that this type of information is not necessary to be included in the Technical Specifications to provide adequate protection of public health and safety. The ITS still retains requirements and Surveillances that verify that the cycle-specific parameter limits are being met. ITS 3.9.1 continues to require that boron concentration limit is met. ITS SR 3.9.1.1 requires periodic verification that boron concentration is within the limits provided in the COLR. The method of determining or utilizing the boron concentration limit has not changed. Also, this change is acceptable because the removed information will be adequately controlled in the COLR under the requirements provided in ITS 5.6.5, "Core Operating Limits Report." ITS 5.6.5 ensures that the applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems limits, and nuclear limits such as SDM, transient analysis limits, and accident analysis limits) of the safety analyses are met. This change is designated as a less restrictive removal of detail change because information relating to cycle-specific parameter limits is being removed from the Technical Specifications. LA.2 (Type 3 - Removing Procedural Details for Meeting TS Requirements or Reporting Requirements) CTS 4.9.1.2 requires that the boron concentration of the Reactor Coolant System and the refueling canal be determined "by chemical analysis" at least once per 72 hours. ITS SR 3.9.1.1 and SR 3.9.1.2 require verification that boron concentration is within the limit specified in the COLR. ITS CNP Units 1 and 2 Page 2 of 4 Attachment 1, Volume 14, Rev. 1, Page 8 of 188
Attachment 1, Volume 14, Rev. 1, Page 9 of 188 DISCUSSION OF CHANGES ITS 3.9.1, BORON CONCENTRATION SR 3.9.1.1 and SR 3.9.1.2 do not specify that the boron concentration be determined by chemical analysis. This changes the CTS by moving details of how the boron concentration is determined from the CTS to the Bases. The removal of these details for performing Surveillance Requirements from the Technical Specifications is acceptable because this type of information is not necessary to be included in the Technical Specifications to provide adequate protection of public health and safety. The ITS still retains the requirement that the boron concentration be verified within its limit. Also, this change is acceptable because these types of procedural details will be adequately controlled in the ITS Bases. Changes to the Bases are controlled by the Technical Specification Bases Control Program in Chapter 5. This program provides for the evaluation of changes to ensure the Bases are properly controlled. This change is designated as a less restrictive removal of detail change because procedural details for meeting Technical Specification requirements are being removed from the Technical Specifications. LESS RESTRICTIVE CHANGES L.A (Category2 -Relaxation of Applicability) CTS 3.9.1 provides limits on the boron concentration of all filled portions of the Reactor Coolant System and the refueling canal when in MODE 6. ITS 3.9.1 modifies this requirement with a Note which states "Only applicable to the refueling canal and refueling cavity when connected to the RCS." This changes the CTS by eliminating the applicability of the boron concentration limits on the refueling canal and refueling cavity when those volumes are not connected to the RCS. In addition, ITS SR 3.9.1.2 requires a verification that the boron is within the limit specified in the COLR once within 72 hours prior to connecting the refueling canal and refueling cavity to the RCS. The purpose of CTS 3.9.1 is to ensure the boron concentration of the water surrounding the reactor fuel is sufficient to maintain the required SHUTDOWN MARGIN. This change is acceptable because the requirements continue to ensure that process variables are maintained in the MODES and other specified conditions assumed in the safety analyses and licensing basis. If the refueling canal and refueling cavity are not connected to the RCS (such as when the reactor vessel head is on the reactor vessel), the boron concentration of those volumes cannot affect the SHUTDOWN MARGIN. In addition, prior to connecting the refueling canal and refueling cavity to the RCS, a boron concentration verification will be performed to ensure the newly connected portions cannot decrease the boron concentration below the limit. This change is designated as less restrictive because the LCO requirements are applicable in fewer operating conditions than in the CTS. L.2 (Category 4 - Relaxation of Required Action) CTS 3.9.1 Action a states that when the boron concentration requirement is not met, initiate and continue boration at > 34 gpm of 6,550 ppm boric acid solution or its equivalent until keff is reduced to < 0.95 or the boron concentration is restored to > 2400 ppm, whichever isthe more restrictive. ITS 3.9.1 Required Action A.3 requires initiation of action to restore boron concentration to within limit. This changes the CNP Units I and 2 Page 3 of 4 Attachment 1, Volume 14, Rev. 1, Page 9 of 188
Attachment 1, Volume 14, Rev. 1, Page 10 of 188 DISCUSSION OF CHANGES ITS 3.9.1, BORON CONCENTRATION CTS by eliminating the specific requirements for the boric acid solution to be used to restore compliance with the LCO. The purpose of CTS 3.9.1 Action a is to restore the required SHUTDOWN MARGIN in a timely manner. This change is acceptable because the Required. Actions are used to establish remedial measures that must be taken in response to the degraded conditions in order to minimize risk associated with continued operation while providing time to repair inoperable features. Specifying the boric acid solution requirements in the Action is not necessary, since the ITS requires that action to restore the boron concentration be initiated immediately. This prompt action will result in the boron concentration being restored as quickly, or more quickly, than the CTS requirement. This change is designated as less restrictive because less stringent Required Actions are being applied in the ITS than were applied in the CTS. L.3 (Category 5 - Deletion of Surveillance Requirement) CTS 4.9.1.1 requires the LCO reactivity condition to be determined prior to removing or unbolting the reactor vessel head, and prior to withdrawal of any full length control rod in excess of 3 feet from its fully inserted position. ITS 3.9.1 does not contain this Surveillance Requirement. The purpose of CTS 4.9.1.1 is to ensure that the LCO requirements are met prior to entering MODE 6 and that the reactor has sufficient SHUTDOWN MARGIN prior to withdrawing any control rods. This change is acceptable because the deleted Surveillance Requirement is not necessary to verify that the values used to meet the LCO are consistent with the safety analyses. Thus, appropriate values continue to be tested in a manner and at a frequency necessary to give confidence that the assumptions in the safety analyses are protected. ITS 3.9.1 requires that the boron concentration be met in MODE 6 or that action be immediately initiated to restore the boron concentration and that all positive reactivity additions be suspended. Therefore, verification that the boron concentration requirement is met must be performed prior to entering MODE 6 in order to avoid immediately entering into an Action and withdrawal of control rods is prohibited when the boron concentration requirement is not met. While the CTS Surveillance is not required, the level of protection provided is appropriate. This change is designated as less restrictive because Surveillances required in the CTS will not be required in the ITS. CNP Units I and 2 Page 4 of 4 Attachment 1, Volume 14, Rev. 1, Page 10 of 188
Attachment 1, Volume 14, Rev. 1, Page 11 of 188 Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs) Attachment 1, Volume 14, Rev. 1, Page 11 of 188
Attachment 1, Volume 14, Rev. 1, Page 12 of 188 CTS Boron Concentration 3.9.1 3.9 REFUELING OPERATIONS 3.9.1 Boron Concentration (ic) 0D LCO 3.9.1 Boron concentrations of the Reactor Coolant Syste the refueling canal, and the refueling cavity shall be maintained within the limit specified In the COLR. .9 APPLICABILITY: MODE 6.
- NOTE -
Only applicable to the refueling canal cavity when connected to the RCS. elm na nd refueling 0D ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME
,^i A. Boron concentration not A.1 Suspend CORE Immediately within lmit. ALTERATIONS.
IjAc~ioy a-AND A.2 Suspend positive reactivity Immediately additions. AND A.3 Initiate action to restore Immediately boron concentration to within limit. SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY .I -/7. SR 3.9.1.1 Verify boron concentration is within the limit specified 72 hours In the COLR. P 1-11 0 WOG STS 3.9.1 - 1 Rev. 2, 04/30/01 Attachment 1, Volume 14, Rev. 1, Page 12 of 188
Attachment 1, Volume 14, Rev. 1, Page 13 of 188 3.9.1 INSERT I SR 3.9.1.2 Verify boron concentration of refueling canal and Once within 72 refueling cavity is within the limit specified in the hours prior to COLR. connecting the refueling canal and refueling cavity to the RCS Insert Page 3.9.1-1 Attachment 1, Volume 14, Rev. 1, Page 13 of 188
Attachment 1, Volume 14, Rev. 1, Page 14 of 188 JUSTIFICATION FOR DEVIATIONS ITS 3.9.1, BORON CONCENTRATION
- 1. Typographical/grammatical error corrected.
- 2. ISTS SR 3.9.1.1 requires a verification that the boron concentration is within limit every 72 hours. The Bases for the SR states that prior to re-connecting portions of the refueling canal or the refueling cavity to the RCS, this SR must be met per SR 3.0.4. SR 3.0.4 requires the SR to be met prior to entering a MODE or other specified condition in the Applicability. However, SR 3.0.4 is only applicable in MODES 1, 2,3, and 4; it is not applicable in MODE 6, the MODE in which ISTS 3.9.1 is applicable. Therefore, to meet the intent of the Bases requirement, a new SR has been added, SR 3.9.1.2, which requires a verification that the boron concentration of the refueling canal and refueling cavity is within the limit specified in the COLR once within 72 hours prior to connecting the refueling canal and refueling cavity to the RCS.
CNP Units 1 and 2 Page 1 of 1 Attachment 1, Volume 14, Rev. 1, Page 14 of 188
Attachment 1, Volume 14, Rev. 1, Page 15 of 188 Improved Standard Technical Specifications (ISTS) Bases Markup and Justification for Deviations (JFDs) Attachment 1, Volume 14; Rev. 1, Page 15 of 188
Attachment 1, Volume 14, Rev. 1, Page 16 of 188 Boron Concentration B 3.9.1 B 3.9 REFUELING OPERATIONS B 3.9.1 Boron Concentration BASES BACKGROUND The limit on the boron concentrations of the Reactor Coolant System (RCS), the refueling canal, and the refueling cavity during refueling ensures that the reactor remains subcritical during MODE 6. Refueling boron concentration Is the soluble boron concentration In the coolant In each of these volumes having direct access to the reactor core during refueling. I i
.1 The soluble boron concentration offsets the core reactivity and is "I
measured by chemical analysis of a representative sample of the coolant In each of the volumes. The refueling boron concentration limit Is j . 14 specified In the COLR. Plant procedures ensure the specified boron concentration In order to maintain an overall core reactivity of k, s 0.95 11 during fuel handling, with control rods and fuel assemblies assumed to be In the most adverse configuration (least negative reactivity) allowed by plant procedures. _ _ ~GDC 26 of 10 CFIR,, AppendixA, requyye that two Inde t A reactivi~ty wcn stems of differae~sgn principeoie (Rf 1) o tee systemzust be capabl oiithreco 0 COMIcar under cold anditions. FThe Chemical and Volume Control System (CVCS) Is the system capable of maintaining the reactor subcritical In cold conditions by maintaining the boron concentration. The reactor Is brought to shutdown conditions before beginning operations to open the reactor vessel for refueling. After the RCS is cooled and depressurized and the vessel head Is unbolted, the head Is j- emovedtf0rm be r~luelci cnviF. The refuelIng cana e Nie reKueting cavity are then flooded with borated water from the refueling water storage tank through the open reactor vessel by gravity feeding or by the use of the Residual Heat Removal (RHR) System pumps. The pumping action of the RHR System in the RCS and the natural circulation due to thermal driving heads in the reactor vessel and refueling cavity mix the added concentrated boric acid with the water In the refuelin canal. The RHR System IsIn operation during refueling (see LCO 3 Residual Heat Removal (RHR) and Coolant Circulation - High Water Level," and LCO 3.9 Residual Heat Removal (RHR) and Coolant Circulation - Low Water Lever') to provide forced circulation in w6 the RCS and assist in maintaining the boron concentrations in the RCS, the refueling canal, and the refueling cavity above the COLR limit. WOG STS B 3.9.1 - 1 Rev. 2, 04/30101 Attachment 1, Volume 14, Rev. 1, Page 16 of 188
Attachment 1, Volume 14, Rev. 1, Page 17 of 188 B 3.9.1 0 INSERT I Plant Specific Design Criterion (PSDC) 27 requires that two independent reactivity control systems, preferably of different design principles, be provided. According to PSDC 28 (Ref. 1), the reactivity controls must be capable of making and holding the core subcritical from any hot standby or hot operating condition. Insert Page B 3.9.1-1 Attachment 1, Volume 14, Rev. 1, Page 17 of 188
Attachment 1, Volume 14, Rev. 1, Page 18 of 188 Boron Concentration B 3.9.1 BASES APPLICABLE During refueling operations, the reactivity condition of the core Is SAFETY consistent with the Initial conditions assumed for the boron dilution ANALYSES accident in the accident analysis and is conservative for MODE 6. The boron concentration limit specified in the COLR is based on the core reactivity at the beginning of each fuel cycle (the end of refueling) and indudes an uncertainty allowance. The required boron concentration and the plant refueling procedures that verify the correct fuel loading plan (including full core mapping) ensure that the k.w of the core will remain s 0.95 during the refueling operation. Hence, at least a 5% tk/lk margin of safety Is established during refueling. During refueling, the water volume Inthe spent fuel pool. the transfer canal, the refueling canal, the refueling cavity, and the reactor vessel form a single mass. As a result, the soluble boron concentration Is relatively the same in each of these volumes. {The limitingoron dilution adent analyzetccurs In MO 5 (Ref. A detailediscusson of thisvent Is provi d In Ba "SHUTDyWN MARGIN (S M).' The RCS boron concentration satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii). LCO The LCO requires that a minimum boron concentration be maintained In \ the RCS, the refueling canalfand te refueling wnu'e in MODE 6. ' @ The boron concentration limit specified in the COLR ensures that a core ro~ s cacJ k.,f of s 0.95 is maintained during fuel handling operations. Violation of the LCO could lead to an inadvertent criticality during MODE 6. APPLICABILITY This LCO is applicable in MODE 6 to ensure that the fuel in the reactor vessel will remain subcritical. The required boron concentration ensures a k. s 0.95. Above MODE 6, LCO 3.1.1, SHUTDOWN MARGIN (SDM)," ensures that an adequate amount of negative reactivity Is available to shut down the reactor and maintain It subcritical. The Applicability is modified by a Note. The Note states that the limits on boron concentration are only applicable to the refueling canal and the refuelingcaity when those volumes are connected toM (C) (L~b: il When the refueling canal and the refuelin avity are isoateafromthe RCS, no potential path for boron dilution exists. (6-.C are ,wi CDA *";'J, r, A 4) WOG STS B 3.9.1 - 2 Rev. 2, 04130101 Attachment 1, Volume 14, Rev. 1, Page 18 of 188
Attachment 1, Volume 14, Rev. 1, Page 19 of 188 Boron Concentration B 3.9.1 BASES ACTIONS A.1 and LA2 Continuation of CORE ALTERATIONS or positive reactivity additions (including actions to reduce boron concentration) is contingent upon maintaining the unit in compliance with the LCO. If the boron concentration of any coolant volume in the RCS, the refueling canal, or the refueling cavity Isless than its limit, all operations Involving CORE ALTERATIONS or positive reactivity additions must be suspended immediately. Suspension of CORE ALTERATIONS and positive reactivity additions shall not preclude moving a component to a safe position. Operations that Individually add limited positive reactivity (e.g temperature fluctuations from Inventory addition or temperature control uctuations). but when combined with all other operations affecting core reactvity (e.g., Intentional boration) result in overall net negative reactivity addition, are not precluded by this action. In addition to Immediately suspending CORE ALTERATIONS and positive reactivity additions, boration to restore the concentration must be initiated immediately. In determining the required combination of boration flow rate and concentration, no unique Design Basis Event must be satisfied. The only requirement Isto restore the boron concentration to Its required value as soon as possible. In order to raise the boron concentration as soon as
-possible, the operator should begin boration with the best source available for unit conditions.
Once actions have been Initiated, they must be continued until the boron concentration Isrestored. The restoration time depends on the amount of boron that must be Injected to reach the required concentration. '4 SURVEILLANCE 3 REQUIREMENTS Hi i@ TtSRtensurthatoroconcentration in the RCS. and connected portions of the refuelingpnal and the reueling within the COLR limits. The bor In em )U nwcenrtonQnKoiit ~4
- ale p~ is determined perihe ical analys A an u~thion ac vitv a e e y nro m the RCS, WOG STS x B 3.9.1-3 Rev. 2, 04/30101 lq4 .M . .
Attachment 1, Volume 14, Rev.-1, Page 19 of 188
Attachment 1, Volume 14, Rev. 1, Page 20 of 188
. . .I .
Boron Concentration II B 3.9.1 .1I d2WolfmS~AD BASES I SURVEILLANCEREQUIREMENTS. (connued) .1 (Sg eis 66orect oron concentratJ npalor tlo co`mmuiato wt1 0
- 1
(~-- ni Freuenc ofonce every 72 hours Is a reasonable amount of 1ime to verify the boron concentration of representative samples. The Frequency Is based on operating experience, which has shown 72 hours to be adequate. 4 - REFEERENCES 1. ppendiC 2 >'J¶" ^ ((DCFR ha.RCh ap15]-
;O 4O O. c; Qgr 'reQIL: CQ; I. aad- refuel$ }
cmjorh 4o C*IRt tcs ~rS I "I i I0 WOG STS. B 3.9.1 - 4 Rev. 2, 04/30101 Attachment 1, Volume 14, Rev. 1, Page 20 of 188
Attachment 1, Volume 14, Rev. 1, Page 21 of 188 JUSTIFICATION FOR DEVIATIONS ITS 3.9.1 BASES, BORON CONCENTRATION
- 1. CNP Units 1 and 2 were designed and under construction prior to the promulgation of 10 CFR 50, Appendix A. CNP Units I and 2 were designed and constructed to meet the intent of the proposed General Design Criteria, published in 1967.
However, the CNP UFSAR contains discussions of the Plant Specific Design Criteria (PSDCs) used in the design of CNP Units 1 and 2. Bases references to the 10 CFR 50, Appendix A, criteria have been replaced with references to the appropriate section of the UFSAR.
- 2. The brackets have been removed and the proper plant specific information/value has been provided.
- 3. Changes are made (additions, deletions, and/or changes) to the ISTS Bases which reflect the plant specific nomenclature, number, reference, system description, analysis, or licensing basis description.
- 4. Changes are made to be consistent with changes made to the Specification.
- 5. Editorial change made for clarity.
- 6. Changes have been made to be consistent with similar words in other places in the ITS Bases.
- 7. Changes made to be consistent with the Specification.
- 8. Typographical/grammatical error corrected.
- 9. The paragraph and associated reference have been deleted since it is discussing a MODE 5 analysis, and this Specification is applicable in MODE 6.
CNP Units I and 2 Page 1 of I Attachment 1, Volume 14, Rev. 1, Page 21 of 188
Attachment 1, Volume 14, Rev. 1, Page 22 of 188 Specific No Significant Hazards Considerations (NSHCs) Attachment 1, Volume 14, Rev. 1, Page 22 of 188
Attachment 1, Volume 14, Rev. 1, Page 23 of 188 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.9.1, BORON CONCENTRATION There are no specific NSHC discussions for this Specification. CNP Units 1 and 2 Page 1 of 1 Attachment 1, Volume 14, Rev. 1, Page 23 of 188
, Volume 14, Rev. 1, Page 24 of 188 ATTACHMENT 2 ITS 3.9.2, Nuclear Instrumentation , Volume 14, Rev. 1, Page 24 of 188 , Volume 14, Rev. 1, Page 25 of 188 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) , Volume 14, Rev. 1, Page 25 of 188
Attachment 1, Volume 14, Rev. 1, Page 26 of 188 ITS 3.9.2 ITS 3/14 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4.9 REFUELING OPERATIONS
- NSTRUMENTAIM .
LIMITING CONDTION FOR OPERATION LCO 3.9.2 3.9.2 . As a minimum, two source range neutron flux monitors shall be i each )ith continu vi a udicati n the cortrol roo!rand one w audia ecount rateccrcuitto be OPERABLE. APPLICA81LITY: MODE 6. ACTION A a. With the requirernents of the above specification not satisfied, immediately suspend all operations involfung CORE ALTERATiONS orloskive reactivity hangelexcept addlition olf atromh M. SpeCification A.1 .712l - (
- b. The provisioraf-pea cation 3.0.3 are not appicable. I 14 lJAdd proposed ACTION B l-- _i SURVEILLANCE REOUTREMENTS [Add proposed ACTO 4.92 Each source range neutron flux monitor shall be demonstrated OPERABLE by performance of-SR 3.9.2.1 a. A CHANNEL CHECK at kast once per 12 hours.
SR 3.9.2.2 b. A CHANNEL CALIBRATION at least every Wl$honth. Note to Neutron detectors may be excluded from CHANNEL CALIBRATION SR 3.9.2.2 COOK NUCLEAR PLANT-UNIT I Page 3/4 9-2 ANIENDMENTO, 1a0, 283 Page 1 of 2 Attachment 1, Volume 14, Rev. 1, Page 26 of 188
Attachment 1, Volume 14, Rev. 1, Page 27 of 188 ITS 3.9.2 ITS 314 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS
.314.9 REFUELING OPERATIONS ISTRUMIENTATION LIMITING CONDITION FOR OPERATION O L LCO 3.9.2 3.9.2 As a minimum, two source range neutron flux monitors shall be l tin each %)thcontnonusn dus vu lindct ion In th~ecootrolmorromand one with nudible count rate circuit to beOPERABLE..Jt I
APPL1ABILITY: MODE 6. ACTION: j .t ACTION A a. With the requirements of the above specification not satified, immediately suspend all operations involving CORE ALTERATiONS or jositivc reactivt chiangalxCept aaditjo tvae rmte RWST, provided the boron concentration ta thlr RWST is greater than the imureiedy Specification 3,A2..2b.1 ) i b. TThe iprovlsiii, - tion 3.0.3 are not Fsdd popose ACTI Z ~ pAdd prozposed ACTION B SURVEtLLANCE REOUTREMENTM 4.92 Each source range neutron flux monitor shall be demonstrated OPERABLE by performance or: SR 3.9.2.1 a A CHANNEL CHECK at least once per 12 hours. (7') crD~~~~ a ny . et 't~ttr I~ PA nff1 NAtnt.. -. I...............It& L.......L.. *". orf O.U.Z.Z U. 1A L-lAttANINiL .. 11.fll¶J I IUji at IC831 e"rYC ILyUJ1U0-I Note to
- Neutron detectors may be excluded from CHANNEL CALIBRATION SR 3.9.2.2 COOK NUCLEAR PLANT-UNIT 2 Page 3l49-2 AMENDMENT 64, 4W, Z9, 267 Page 2 of 2 Attachment 1, Volume 14, Rev. 1, Page 27 of 188
Attachment 1, Volume 14, Rev. 1, Page 28 of 188 DISCUSSION OF CHANGES ITS 3.9.2, NUCLEAR INSTRUMENTATION ADMINISTRATIVE CHANGES A.1 In the conversion of the CNP Current Technical Specifications (CTS) to the plant specific Improved Technical Specifications (ITS), certain changes (wording preferences, editorial changes, reformatting, revised numbering, etc.) are made to obtain consistency with NUREG-1431, Rev. 2, "Standard Technical Specifications-Westinghouse Plants" (ISTS). These changes are designated as administrative changes and are acceptable because they do not result in technical changes to the CTS. A.2 CTS 3.9.2 Action b contains the statement, "The provisions of Specification 3.0.3 are not applicable." ITS 3.9.2 does not contain an equivalent statement. This changes the CTS by deleting the Specification 3.0.3 exception. This change is acceptable because the technical requirements have not changed. ITS LCO 3.0.3 is not applicable in MODE 6. Therefore, the CTS LCO 3.0.3 exception is not needed. This change is designated as administrative because it does not result in a technical change to the CTS. MORE RESTRICTIVE CHANGES M.1 CTS 3.9.2 states, in part, that two source range neutron flux monitors shall be "operating." ITS 3.9.2 states, in part, that two source range neutron flux monitors shall be "OPERABLE." This changes the CTS by requiring the source range neutron flux monitors to be OPERABLE, instead of just operating. The purpose of CTS 3.9.2 is to ensure that the source range neutron flux monitors are capable of performing the safety functions assumed in the accident analysis. However, as written, the CTS LCO wording could be interpreted to allow the source range neutron flux monitors to be operating in a location or condition that would prevent them from performing the assumed safety function. The ITS wording eliminates this possible misinterpretation. This change is acceptable because the source range neutron flux monitors must be OPERABLE (i.e., capable of performing their safety function) instead of just operating. This change is designated as more restrictive because the ITS contains more specific requirements for a specific component. M.2 CTS 3.9.1 Action a requires the immediate suspension of positive reactivity changes except for the addition of water from the RWST, provided the boron concentration in the RWST is greater than the minimum required by Specification 3.1.2.7.b.2 (i.e., 2400 ppm). ITS 3.9.2 Required Action A.2 requires suspension of operations that would cause introduction into the RCS, coolant with boron concentration less than required to meet the boron concentration of LCO 3.9.1. This changes the CTS by replacing the allowance to allow a positive reactivity change from the addition of water from the RWST, provided the boron concentration of the RWST is greater than 2400 ppm with a requirement that the boron concentration must meet the boron concentration of LCO 3.9.1. CNP Units 1 and 2 Page 1of 6 Attachment 1, Volume 14, Rev. 1, Page 28 of 188
Attachment 1, Volume 14, Rev. 1, Page 29 of 188 DISCUSSION OF CHANGES ITS 3.9.2, NUCLEAR INSTRUMENTATION The purpose of CTS 3.9.2 Action a is to provide assurance that activities that could result in reducing boron concentration such that the required SHUTDOWN MARGIN is not met will not occur when any source range neutron flux monitor is inoperable in MODE 6. Allowing positive reactivity additions from sources with boron concentrations meeting the requirements of ITS 3.9.1 preserves the required SHUTDOWN MARGIN. This change is acceptable because the ITS requires actions that prohibit activities that could result in reducing boron concentration such that the required SHUTDOWN MARGIN is not met. This change is designated as more restrictive because it provides more restrictive corrective actions in the ITS than in the CTS. M.3 CTS 3.9.2 Action a states that with fewer than two source range channels operating, immediately suspend all operations involving CORE ALTERATIONS or positive reactivity changes except addition of water from the RWST, provided the boron concentration in the RWST is greater than the minimum required by Specification 3.1.2.7.b.2 (i.e., 2400 ppm). The ITS provides similar ACTIONS as the CTS (except where changed as described in DOCs M.2 and L.2). In addition, ITS 3.9.2 ACTION B requires additional actions when two source range neutron flux monitors are inoperable. The ITS requires immediate initiation of action to restore one source range neutron flux monitor to OPERABLE status and to perform a verification of boron concentration (per ITS SR 3.9.1.1) once per 12 hours. This changes the CTS requirements by requiring an additional verification of boron concentration every 12 hours when both source ranges are inoperable and by requiring an additional action to initiate immediate action to restore one source range neutron flux monitor to OPERABLE status. The purpose of this change is to provide necessary Required Actions that are appropriate for a possible condition that could be encountered. This change is acceptable because the proposed Required Actions are reasonable and necessary to ensure the reactor is maintained in a safe condition. This change is more restrictive because it provides for additional actions that the CTS does not require. M.4 Not used. M.5 Not used. RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES LA.1 (Type I - Removing Details of System Design and System Description, Including Design Limits) CTS 3.9.2 states that two source range neutron flux monitors shall be operating, "each with continuous visual indication in the control room." ITS 3.9.2 LCO states that two source range neutron flux monitors shall be OPERABLE. This changes the CTS by moving the requirement that each CNP Units 1 and 2 Page 2 of 6 Attachment 1, Volume 14, Rev. 1, Page 29 of 188
Attachment 1, Volume 14, Rev. 1, Page 30 of 188 DISCUSSION OF CHANGES ITS 3.9.2, NUCLEAR INSTRUMENTATION channel has a continuous visual indication in the control room from the CTS to the Bases. The removal of this detail, which is related to system design, from the Technical Specifications, is acceptable because this type of information is not necessary to be included in the Technical Specifications to provide adequate protection of public health and safety. The ITS retains the requirement that two channels be OPERABLE and continues to require the associated Surveillance to verify OPERABILITY. This change is acceptable because the removed information will be adequately controlled in the ITS Bases. Changes to the Bases are controlled by the Technical Specification Bases Control Program in Chapter 5. This program provides for the evaluation of changes to ensure the Bases are properly controlled. This change is designated as a less restrictive removal of detail change because information relating to system design is being removed from the Technical Specifications. LESS RESTRICTIVE CHANGES L.1 Not used. L.2 (Category 4 - Relaxation of Required Action) CTS 3.9.2 Action a states that with fewer than two source range neutron flux monitors operating, immediately suspend all operations involving positive reactivity changes except addition of water from the RWST, provided the boron concentration in the RWST is greater than the minimum required by Specification 3.1.2.7.b.2 (i.e., 2400 ppm). ITS 3.9.2 Required Action A.2 states "Suspend operations that would cause introduction of coolant into the RCS with boron concentration less than required to meet the boron concentration of LCO 3.9.1, Boron Concentration." This allows positive reactivity changes provided they do not reduce the boron concentration below the refueling limit. This changes the CTS requirements by allowing limited positive reactivity additions from sources in addition to the RWST. This change is acceptable because the Required Actions are used to establish remedial measures that must be taken in response to the degraded conditions in order to minimize risk associated with continued operation while providing time to repair inoperable features. The Required Actions are consistent with safe operation under the specified Condition, considering the OPERABLE status of the redundant systems or features. This includes the capacity and capability of remaining systems or features, a reasonable time for repairs or replacement, and the low probability of a DBA occurring during the repair period. The requirement to maintain refueling boron concentration within limits will continue to ensure the unit will be operated within the assumptions of the safety analyses. This change is designated as less restrictive because less stringent Required Actions are being applied in the ITS than were applied in the CTS. L.3 (Category 4 - Relaxation of RequiredAction) CTS 3.9.2 Action a requires the immediate suspension of CORE ALTERATIONS or positive reactivity changes except for the addition of water from the RWST, provided the boron concentration in the RWST is greater than the minimum required by Specification 3.1.2.7.b.2, in the event one source range neutron flux monitor with CNP Units 1 and 2 Page 3 of 6 Attachment 1, Volume 14, Rev. 1, Page 30 of 188
Attachment 1, Volume 14, Rev. 1, Page 31 of 188 DISCUSSION OF CHANGES ITS 3.9.2, NUCLEAR INSTRUMENTATION audible indication in the containment is not operating. ITS 3.9.2 ACTION C requires initiation of action to isolate unborated water sources in the event the required source range audible count rate circuit is inoperable. This changes the CTS by replacing the Action to immediately suspend CORE ALTERATIONS or positive reactivity changes except for the addition of water from the RWST, provided the boron concentration in the RWST is greater than the minimum required by CTS 3.1.2.7.b.2, in the event one source range monitor with audible indication in the containment is not operating, with the Action to initiate action to isolate unborated water sources. The purpose of CTS 3.9.2 Action a is to provide assurance that activities that could result in an inadvertent criticality will not occur when the required source range audible count rate circuit is inoperable in MODE 6. This change is acceptable because the Required Actions are used to establish remedial measures that must be taken in response to the degraded conditions in order to minimize risk associated with continued operation while providing time to repair inoperable features. The Required Actions are consistent with safe operation under the specified Condition, considering the OPERABLE status of the redundant systems or features. This includes the capacity and capability of remaining systems or features, a reasonable time for repairs or replacement, and the low probability of a DBA occurring during the repair period. ITS 3.9.2 ACTION C requires actions to be taken to isolate sources of unborated water. This provides assurance that rapid dilution of boron concentration, which could result in rapid reduction in shutdown margin, will not occur. This change preserves the assumptions and conclusions of the boron dilution analysis. This change is designated as less restrictive because less stringent Required Actions are being applied in the ITS than were applied in the CTS. L.4 (Category 11 - 18 to 24 Month Surveillance Frequency Change, Channel Calibration Type) CTS 4.9.2.b requires a CHANNEL CALIBRATION of each source range neutron flux monitor every 18 months. ITS SR 3.9.2.2 requires the performance of a CHANNEL CALIBRATION every 24 months. This changes the CTS by extending the Frequency of the Surveillance from 18 months (i.e., a maximum of 22.5 months accounting for the allowable grace period specified in CTS 4.0.2 and ITS SR 3.0.2) to 24 months (i.e., a maximum of 30 months accounting for the allowable grace period specified in CTS 4.0.2 and ITS SR 3.0.2). The purpose of the CHANNEL CALIBRATION required by CTS 4.9.2.b is to ensure the source range neutron flux monitor will function correctly to ensure the safety analysis can be met. Extending the SR Frequency is acceptable because the source range neutron flux monitoring channels are designed to be highly reliable. Furthermore, a CHANNEL CHECK for the source range neutron flux monitoring channels is performed on a more frequent basis (ITS SR 3.9.2.1). The CHANNEL CHECK provides a qualitative demonstration of the OPERABILITY of the instrument. This change was evaluated in accordance with the guidance provided in NRC Generic Letter No. 91-04, "Changes in Technical Specification Surveillance Intervals to Accommodate a 24-Month Fuel Cycle," dated April 2, 1991. The CNP Units 1 and 2 Page 4 of 6 Attachment 1, Volume 14, Rev. 1, Page 31 of 188
Attachment 1, Volume 14, Rev. 1, Page 32 of 188 DISCUSSION OF CHANGES ITS 3.9.2, NUCLEAR INSTRUMENTATION impacted source range neutron flux monitoring instrumentation was evaluated through a failure analysis and a qualitative drift analysis: Westinghouse Source Range, Neutron Flux This function is performed by SRM Neutron Flux Detectors (Westinghouse Model WL-23706), SRM Neutron Flux Drawers (Westinghouse Model 6051 D50G01), a Weschler HX-252 Indicator, and a Tracor Westronics Recorder (Model 4200 (Unit 1)and Model 4220 (Unit 2)). These system components were not evaluated for drift but were justified for extension based on engineering judgment. SRMs satisfy their design function if calibration is sufficient to ensure neutron level is observable when the reactor is shutdown. This is verified by CHANNEL CHECKS at least every 12 hours when the reactor is shutdown. The SRMs must be operational in MODE 6. SRM response to reactivity changes is distinctive and well known to plant operators, and SRM response is closely monitored during these reactivity changes. Additionally, since there is very little neutron activity during loading, refueling, shutdown, and approach to criticality, a neutron source is placed in the reactor during approach to criticality to provide a minimum observable SRM neutron count rate attributable to core neutrons of at least 2 counts per second. During plant shutdowns and startups, overlap between the IRM channels and the SRM channels is routinely verified to ensure performance of the SRM channels. There is also more frequent testing, including a COT every 184 days in MODES 1 and 2 and every 31 days in MODES 3, 4, and 5, to verify operation of the electronics for the source range trip. Therefore, any substantial degradation of the SRMs will be evident and long term drift has no impact on the accuracy of this circuit. The results of these analyses will support a 24 month Surveillance interval. Thermo Gamma-Metrics Neutron Flux Monitors The function is performed by a Wide Range Detector Assemblies (Gamma Metrics model numbers 200749-103 and 200574-11), Signal Processing Drawers (Gamma Metrics model numbers 900180-101 and 900091-101), a Weschler HX-252 Indicator, and a Tracor Westronics Recorder (Series 4200). These system components were not evaluated for drift but were justified for extension based on engineering judgment. SRMs satisfy their design function if calibration is sufficient to ensure neutron level is observable when the reactor is shutdown. This is verified by CHANNEL CHECKS at least every 12 hours when the reactor is shutdown. The SRMs must be operational in MODE 6. SRM response to reactivity changes is distinctive and well known to plant operators, and SRM response is closely monitored during these reactivity changes. Therefore, any substantial degradation of the SRMs will be evident and long term drift has no impact on the accuracy of this circuit. The results of these analyses will support a 24 month Surveillance interval. Based on the design of the instrumentation and the qualitative drift evaluations, it is concluded that the impact, if any, from this change on system availability is minimal. A review of the Surveillance test history was performed to validate the above conclusion. Those tests that were classified as failures were evaluated and primarily involved components found with out of tolerance calibration data. The other failures were reviewed and those failures did not invalidate the CNP Units 1 and 2 Page 5 of 6 Attachment 1, Volume 14, Rev. 1, Page 32 of 188
Attachment 1, Volume 14, Rev. 1, Page 33 of 188 DISCUSSION OF CHANGES ITS 3.9.2, NUCLEAR INSTRUMENTATION conclusion that the impact, if any, on system availability from this change is minimal. In addition, the proposed 24 month Surveillance Frequency, if performed at the maximum interval allowed by ITS SR 3.0.2 (30 months) does not invalidate any assumptions in the unit licensing basis. This change is designated as less restrictive because Surveillances will be performed less frequently under the ITS than under the CTS. CNP Units 1 and 2 Page 6 of 6 Attachment 1, Volume 14, Rev. 1, Page 33 of 188
Attachment 1, Volume 14, Rev. 1, Page 34 of 188 Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs) Attachment 1, Volume 14, Rev. 1, Page 34 of 188
Attachment 1, Volume 14, Rev. 1, Page 35 of 188 CTS-
- Nudear Instrumentation 3.9.~
3.9 REFUELING OPERATIONS
- 1 3.9?. Nuclear Instrumentation II i LCO 3.9.i Two source range neutron flux monitors shall be OPERABLE. 0
.K7,z i
i
.1 COne source range audible OPERABLE.0 Icount ratezircult shall be 0 APPLICABILITY: MODE 6.
ACTIONS CONDITION REQUIRED ACTION ICOMPLETION TIME A. Onerequirecsource A.1 Suspend CORE Immediately range neutron flux ALTERATIONS. monitor inoperable. AND 4*0a4 A.2 Suspend operations that Immediately would cause Introduction._ 7,~ Into the RCSbEPwith boron concentration less D I than required to meet the boron concentration of
--- 4_CO 391 B. Twodrequiredcsource B.1 Initiate action to restore Immediately range neutron flux one source range neutron monitors Inoperable. flux monitor to OPERABLE I.i-i -jDoe M.t AND status.
3 1 B.2 Performn SR 3.9.1.1. Once per 12 hours
'i i
I WOG STS 3.9.3 -1 Rev. 2, 04/30/01 Attachment 1, Volume 14, Rev. 1, Page 35 of 188
Attachment 1, Volume 14, Rev.. 1, Page 36 of 188 Nuclear Instrumentation 3.9. ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME L . I i' C.1 Initiate action to isolate Immediatelyo
- REVIEW (S NOTE- unborated water sources.
t1 Condition s Included only for plan at assume a born ilution event is a m ated byopera lsponse o an a ble
~source range ication.
CORequired source range audible a ount rateocircui7T-operable. SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY q.?. q SR 3.9.1- Perform CHANNEL CHECK. 12 hours I il SR 3.942 -NOTE- 0 Neutron detectors are excluded from CHANNEL CAUBRATION. Perform CHANNEL CALIBRATION. (moths -1
.1I
'i A, I
- 1 WOG STS- 3.9.3 - 2 Rev. 2, 04130/01 Attachment 1, Volume 14, Rev. 1, Page 36 of 188
Attachment 1, Volume 14, Rev. 1, Page 37 of 188 JUSTIFICATION FOR DEVIATIONS ITS 3.9.2, NUCLEAR INSTRUMENTATION
- 1. CNP has analyzed a boron dilution event in MODE 6. Therefore, ISTS 3.9.2 is not included in the ITS and ISTS 3.9.3 is renumbered as ITS 3.9.2.
- 2. The brackets are removed and the proper plant specific information/value is provided.
- 3. Editorial correction to be consistent with the format of the ITS.
- 4. Changes have been made to be consistent with changes made to another Specification. I CNP Units 1 and 2 Page 1 of 1 Attachment 1, Volume 14, Rev. 1, Page 37 of 188
Attachment 1, Volume 14, Rev. 1, Page 38 of 188 Improved Standard Technical Specifications (ISTS) Bases Markup and Justification for Deviations (JFDs) Attachment 1, Volume 14, Rev. 1, Page 38 of 188
Attachment 1, Volume 14, Rev. 1, Page 39 of 188 Nuclear Instrumentatior B 3.945o .i B 3.9 REFUELING OPERATIONS B 3.9v3 Nuclear Instrumentation BASES BACKGROUND
- REVIEWER' OTE -
Bracketed optis are provided for rce range OPERABIL requireme a Include audible al or count rate functio These 0 options plyto plants that assu e a boron dilution eve that is mitig d by operator respon to an audible Indictio For plants that Iste all boron dilution pa (per LCO 3.9.2) the urce range i ERABILITY Includes a visual monitoring f ction. The source range neutron flux monitorstrused during refueling jFrjjrrfr m oortos to monitor thA moeQ reactivity condition. lie~ibsouroe 2-( esXz~eragenutron lgrrmonhors ar cart of th-eNu-clear Instrumentation l =.-sJV&T Syse (NIS).!sheseedetectors are located external to the reactor vessel ad detect neutrons le-aking from the core. ~ The source range neutron flux monitors are BF3 detectors ( operating In the proportional region of the gas filled detector characteristic curve. The detectors monitor the neutron flux in counts per second. The Instrument range covers six decades of neutron flux (1E+6 cps) I a in u C The detectors also [rovde (2J continuous visual indication In the control oromind an audibla
. -,ioou t ratebo alert operators to a possible dilution acciden The IS is .acordance with the criteria presented In Refernce 1.
APPLICABLE Two OPERABLE source range neutron flux monitors are required to D4 -0 SAFETY provide a signal to alert the operator to unexpected changes In core ANALYSES reactivity such as with a boron dilution accident (Ref. 2) or an improperly (2) loaded fuel assembly.0rhe audible count rate fromthesourceran e neutron flux monItors provides prompt and definite o f any boron 5 dilution. The count rate Increase is proportional to the subcritical multiplication factor and allows operators to promptly recognize the Initiation of a boron dilution event Prompt recognition of the initiation of a boron dilution event is consistent with the assumptions of the safety analysis and is necessary to assure sufficient time Isavailable for Isolation of the primary water makeup source before SHUTDOWN
- MARGIN Islost (Ref. 2)0)
WOG STS B 3.9.3-1 Rev. 2, 04130/01 Attachment 1,' Volume 14, Rev. 1, Page 39.of 188 -
Attachment 1, Volume 14, Rev. 1, Page 40 of 188 B 3.9.2 0 INSERT I (i.e., the Westinghouse source range neutron flux monitors and the Thermo Gamma-Metrics neutron flux monitors) 0 INSERT 2 The Thermo Gamma-Metrics neutron flux monitors are part of the Thermo Gamma-Metrics Neutron Flux Monitoring System. Both of Q3 INSERT 3 (selectable between proportional source range neutron flux monitors) Q INSERT 4 There are two Thermo Gamma-Metrics neutron flux monitors. Each monitor includes two fission chamber detectors capable of monitoring a wide range from source level (shutdown) to full power reactor operation. In the source range, the detectors monitor the neutron flux in counts per second and are capable of detecting six decades of neutron flux. The detectors also provide continuous visual indication in the control room of source count rate and a source rate of change. Insert Page B 3.9.3-1 Attachment 1, Volume 14, Rev. 1, Page 40 of 188
Attachment 1, Volume 14, Rev. 1, Page 41 of 188 Nuclear instrumentation.&f) B3.9. BASES APPLICABLE SAFETY ANALYSES (continued) e need f analysis for an n ron dilution e t The source range neutron flux monitors satisfy Criterion 3 of 10 CFR 50.36(c)(2)(11). LCO This LCO requires that two source range neutron flux monntoto ~ OPERABLE to ensure that redundant monitoring capability Isavailable to z detect changes In core reactivity. To be OPERABLE, each monitor must__ _________C= i The ~~~~~~neetfrufentyanayi provide visualorlndlcationffn an nohld boro the conto ronan addition at least onedp 3iuio.3.gffz 2 (JEm monito ~must providean OPERAB ~audible [a?4 uta9 APPLICABILITY In MODE 6. the source range neutron flux monitors must be OPERABLE to determine changes Incore reactivity. There are no other direct means available to check core reactivity levels. In MODES 2, 3.4, and 5, these same Instalied source range detectors and circuitry are also required to be OPERABLE by SCO dTrip 3.3.1, Sstem RTS ACTIONS A.1 and A,2 uirr 1 V With only oneuource range neutron flux monitor OPERABLE, redundancy has been lost. Since these Instruments are the only direct means of monitoring core reactivity conditions, CORE ALTERATIONS and Introduction of coolant Into the RCS with bonon concentration less than required to meet the minimum boron concentration of LCO 3.9.1 must be suspended Immediately. Suspending positive reactivity additions that could result In failure to meet the minimum boron concentration limit Is required to assure continued safe operation. Introduction of coolant Inventory must be fromi sources that have a boron concentration greater than that what would be required In the RCS for minimum refueling boron concentration. This may result inan overall reduction In RCS boron concentration, but provides acceptable margin to maintaining subcritical operation. Performance of Required Action A.1 shall not preclude completion of movement of a component to a safe position. WOG STS B 3.9.3 -2 Rev. 2, 0430/01 Attachm.ient 1, Volumre 14, Rev. 1, Page 41O of 188
Attachment 1, Volume 14, Rev. 1, Page 42 of 188 B 3.9.2 Q INSERT5 (any combination of Westinghouse source range neutron flux and Thermo Gamma-Metrics neutron flux monitors) Q) INSERT 6 (which must be a Westinghouse source range neutron flux monitor, since the Thermo Gamma-Metrics neutron flux monitors do not have an audible count rate function) Insert Page B 3.9.3-2 Attachment 1, Volume 14, Rev. 1, Page 42 of 188
Attachment 1, Volume 14, Rev. 1, Page 43 of 188 Nuclear Instrumentation ya) B 3.90 0 BASES 1 ACTIONS (continued) With nource range neutron flux monitor OPERABLE, action to restore a monitor to OPERABLE status shall be Initiated immediately. Once initiated, action shall be continued unfil a source range neutron flux monitor is restored to OPERABLE status. (23 With n ource range neutron flux monitor OPERABLE, there are no direct means of detecting changes in core reactivity. However, since CORE ALTERATIONS and positive reactivity additions are not to be 4i made, the core reactivity condition Is stabilized until the source range neutron flux monitors are OPERABLE. This stabilized condition Is determined by performing SR 3.9.1.1 to ensure that the required boron concentration exists. The Completion Time of once per 12 hours is sufficient to obtain and analyze a reactor coolant sample for boron concentration and ensures that unplanned changes In boron concentration would be Identified. The 4 12 hour Frequency Is reasonable, considering the low probability of a change In core reactivity during this time period. With no audible Indication of a boi
'ount ratOPERABLE, prompt and definite luton event, consistent with the assumptions of 33 the safety analysis, Is lost. In this situation, the boron dilution event may not be detected quickly enough to assure sufficient time is available for operators to manually isolate the unborated water source and stop the dilution prior to the loss of SHUTDOWN MARGIN. Therefore, action must be taken to prevent an Inadvertent boron dilution event from occurring. This Is accomplished by Isolating all the unboraed water flow paths to the Reactor Coolant System. Isolating these flow paths ensures that an Inadvertent dilution of the reactor coolant boron concentration is prevented. The Completion Time of 'Immediatel' assures a prompt response by operations and requires an operator to Initiate actions to Isolate an affected flow path Immediately. Once actions are initiated, they must be continued until all the necessary flow paths are Isolated or the 3 circuit Is restored to OPERABLE status; .
.3V WOG STS B 3.9.3 - 3 Rev. 2, 04/30101 Attachment 1, Volume 14, Rev. 1, Page.43 of 188
Attachment 1, Volume 14, Rev. 1, Page 44 of 188 Nuclear Instrumentatio~njO...~ B 3.9J BASES* SURVEILLANCE SRct3o9iy19 REQUREMNTS SR 3.9.0.1Is the performance of a CHANNEL CHECK, which I
.1 I
cornparison of the parameter Indicated on one channel to a Simia UO-49'-araieer ntther channels It Is based on the assumption tha te two Indication channels should be consistent with core conditions. Changes 4 iw Infuel loading and core geometry can result Insignificant differences
*between source range channels, but each channel should be consistent with Its local conditions.
l1 The Frequency of 12 hours Is consistent with the CHANNEL CHECK Frequency specified slrhilarly for the same Instruments In LCO 3.3.1.
-outage. Operating experience has shown the pass the Surveillance when performed at the,
- 2. (TSAR, Sectio<oj>\'"
.1 i WOG STS B 3.9.3 - 4 Rev. 2, 04130101 Attachment.1, Volume'14,Re'v. 1, Page 44 of 188
Attachment 1, Volume 14, Rev. 1, Page 45 of 188 B 3.9.2 Q INSERT7 CHANNEL CALIBRATION is a complete check of the instrument loop, except the detector. Insert Page B 3.9.3-4 Attachment 1, Volume 14, Rev. 1, Page 45 of 188
Attachment 1, Volume 14, Rev. 1, Page 46 of 188 JUSTIFICATION FOR DEVIATIONS ITS 3.9.2 BASES, NUCLEAR INSTRUMENTATION
- 1. Changes are made to reflect those changes made to the ISTS. Subsequent requirements are renumbered or revised, where applicable, to reflect the changes.
- 2. The brackets have been removed and the proper plant specific information/value has been provided.
- 3. Changes are made (additions, deletions, and/or changes) to the ISTS Bases which reflect the plant specific nomenclature, number, reference, system description, analysis, or licensing basis description.
- 4. The specific accuracy of the source range neutron flux monitors is not part of the licensing basis of CNP and has been deleted.
- 5. Typographical/grammatical error corrected.
- 6. CNP Units I and 2 were designed and under construction prior to the promulgation of 10 CFR 50, Appendix A. CNP Units 1 and 2 were designed and constructed to meet the intent of the proposed General Design Criteria, published in 1967.
However, the CNP UFSAR contains discussions of the Plant Specific Design Criteria (PSDCs) used in the design of CNP Units I and 2. Bases references to the 10 CFR 50, Appendix A, criteria have been replaced with references to the appropriate section of the UFSAR.
- 7. Changes are made to be consistent with similar words in other places in the ITS Bases.
CNP Units 1 and 2 Page 1 of 1 Attachment 1, Volume 14, Rev. 1, Page 46 of 188
Attachment 1, Volume 14, Rev. 1, Page 47 of 188 Specific No Significant Hazards Considerations (NSHCs) Attachment 1, Volume 14, Rev. 1, Page 47 of 188
Attachment 1, Volume 14, Rev. 1, Page 48 of 188 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.9.2, NUCLEAR INSTRUMENTATION There are no specific NSHC discussions for this Specification. CNP Units 1 and 2 Page 1 of 1 Attachment 1, Volume 14, Rev. 1, Page 48 of 188
, Volume 14, Rev. 1, Page 49 of 188 ATTACHMENT 3 ITS 3.9.3, Containment Penetrations , Volume 14, Rev. 1, Page 49 of 188 , Volume 14, Rev. 1, Page 50 of 188 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) , Volume 14, Rev. 1, Page 50 of 188
Attachment 1, Volume 14, Rev. 1, Page 51 of 188 ITS 3.9.3 ITS 314 LIMWMIG CONDIUMON FOR OPlRATIOffMD SURVLLANC3;kQUZV Sf4.9 RZFUD.IG OftNATWMONS-LCO 3.9.3 3.9A 71 ;1 'hling pewhadns daa be tn dhe Alflowing a LCO 3.9.3.a R.
- 7b ~nat door chewd md beld kiphwe by a n1ieimof 1=boki6 b..
- Te sirdo& dooin an cmfro~odIn &aiowirg mcw LCO 3.9.3.b L -Ao -o f wIe&wIm e& 0ck ,isclod-00qg r
- 2. Batliudockdoors esybe %w~urde&
g064 orinmachu~ocki0P~tABL I& A-&HrntftI individhml kay~ftabb's an~ttr 1. =~~Ohio& iff"Muil"&! LCO 3.9.3.c
- E.chpanetcatimpmOvidwtgdrecacces Aitmx a0ni0phoe to d octzld mtp~
s~amt dalbe
- 1. Goedby m ihodm valve,bllufnd &nAaial valve. at equlvalast cc I
- 2. Be capablo of befin docd by n CEABIX -- C tpaw aed E~xlut ifhout peceftrzfic f1ow path) jmqvMEq &mecta ufl te ac1sspb~etotfr afEacaftmoxiprs h: ~ L J D AHi nf ~ 0 3o f i uledhhaad el th co ACTION A WhI& do of fth &aboie ia&ZedJUo nod~t nw of kradited fad in to emu : bawteg T of Spedcilkdoi SR 3.9.3.1.
SR 3.9.3.2 COOK NUCLEAR PLANT-UN1rl Pap314 9-4 AMVMMIM4M 259 Page 1 of 8 Attachment 1, Volume 14, Rev. 1, Page 51 of 188
Attachment 1, Volume 14, Rev. 1, Page 52 of 188 ITS 3.9.3 ITS 3/4 ARCIG CONDrONS FOR OPfRA1ONAND smVEILANCE E mQU~ 3M.9 RKTIJWG OPRATXONS SURVEMANCE SR 3.9.3.1 L Veri1 e - rag iavfleqqfsfii.a I SR 3.9.3.2 b. Tcstieng g Cmila~ PWe mld By iao=n Wum pa *c plcable Xdo of Spmcatin 4.63.1X COOK NUCLEAR PLANrT-NUT I Pt3M 9/4
- AMENDNT 259 I Page 2 of 8 Attachment 1, Volume 14, Rev. 1, Page 52 of 188
Attachment 1, Volume 14, Rev. 1, Page 53 of 188 ITS 3.9.3 ITS 0 3/4 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4.6 CONTAINMENT SYSTEMS SURVEILLANCE REOUIREMENTS (Coninud) 44.6.3.1.2 Each containment isolation valve shall be demonstrated OPERABLE at least once per 18 months by: (S 36T3 a. Verifying that on a Phase Acontaiment isolation test signal, each Phase A isolation valve actuates to its isolation position.
- b. Verifying that on a Phase B containment isolation test signal, each Phase B isolation valve k 2rh12te;- fn he- inlatien positienno
-- GAcdd Rroposed SR 3.9.3.2 NoteIL3\
SR 3.9.3.2 C. Verifying that on a Containment Purge and Exhust isolation sig . each Purge and Exhaust valve actuates to its isolation position. c o [Seelrs (S3.6.3 4.6.3.1.3 The isolation tune o ea power operat or automatic containment isolation valve s4 determined to be within its imit when tested pursuant to Specification 4.0.5. COOK NUCLEAR PLANT-UNIT I Page 3/4 6-15 AMENDMENT 47, 444. 444. 468. 4&1, 275 Page 3 of 8 Attachment 1, Volume 14, Rev. 1, Page 53 of 188
Attachment 1, Volume 14, Rev. 1, Page 54 of 188 ITS 3.9.3 ITS REFUELING OPERATIONS COI(TAINI4ENT PURGE AND EXHAUST ZSOLATION SYSE LIM4ING CONDITION FOR OPERATION LCO 3.9.3.c.2 3.9.3 ,Thq Contairment Purpg anod Exhaust isolation systan shall be A. U CAU AlCAI Duringoplta rt ormovement of irradiated fuel within ACTION: LCO 3.9.3.c1 With the Cnntauirmnt Purge and Exhaust Isolation system inoperable. close each of the Purge and Exhaust penetrations providins direct from the cntainmnt
- s*rere to tho eutrdoaarphet.
aess P2BM ~ acfcto
=ffibls. 3.. 423 SR 3.9.3.2 recaie~onfl 51 Tre eecffarwe containaafnt radiation lnstrunentatSeiIT monitor. 3.30
- 0. C. C:CX-.UNITlI- 314 9-10 Anenduent No. 80 Page 4 of 8 Attachment 1, Volume 14, Rev. 1, Page 54 of 188
Attachment 1, Volume 14, Rev. 1, Page 55 of 188 ITS 3.9.3 ITS 314 LMdlING CONDITIONS FOR OPERATION AND SURVELnlANCE P.EQUIREMENrS 314.9 REFUELING OPERATIONS COMTAINGMen PEFATION OLRNG LIAMlNG CONDITION FOR OPERA1171 LCO 3.9.3 3-9.4 The containrent building penetrations shall be in the following stahis LCO 3.9.3.a a. The equipmict door closed and held in place by a rmmnix-m of four bolts,
- b. The airlock doors are confroiled in the fbllowing LCO 3.9.3.b
- 1. A:iniu of one door in each airlock is closed, i
- 2. Boh airlock doors maybe oven trovided:
La. .One door ineadiairlokisOPERABLEJ.
- b. _aRcinrg cavity kw greater than23 e ieel, and
-e LCO 3.9.3.c C.
IC. .4z ignated individual isxvslable at all tifes to c ele airke k if reouL-adj Each penetration providing direct access from the contaimrrent atmosphere to tbe.outside atmosphere 9 shall be eithe
- 1. Closedby an isolation valvc, blind flnge, mnual valve, or equivalent, or I
- 2. Be capable of being closed by an OPERABLE automatic Containment Purge and Exhaust isolation valve.
NOIE1 Penetration tlow path(s) providing direct access from the containment atrimsphere to the outside atmosphere via the auxiliary building vent may be unisolated wnder administrative controls. APPLICABILI'M : During POR AL
~ELI~ X:
ll1Lf Duing R AL TONS or 2 T ~ !of irradiated fuzel within the con in nt ACTI0: ACTION A With the requrets of Ihe above specification not satisfied, im* ediately suspend op0ops. b CO oe iradiated fuel in the containment building.1 of on of Specaia ue SURVEnU.ANCE REQUIRElYIL.2 SR 3.9.3.1. 4.9.4 Each of the above ontahonent iding penetrations shall be determined to be in is required status SR 3.9.3.2 th l00 tbe pc e7__during ,CORE _ ____ leostast o__rf _S - of irradiated fuel in the coraretbidn by. X Forthepurpose Specification, an OP is or tispable of be o secued Cablor bostes transversing thea abaflbe designed to low for manner(e.g., tvanaimely LA. quick disco ). COOK NUCLEAR PLANT-lNlT 2 Page 3/4 9.4 AMENDMENT , M,4, 242 Page 5 of 8 Attachment 1, Volume 14, Rev. 1, Page 55 of 188
Attachment 1, Volume 14, Rev. 1, Page 56 of 188 ITS 3.9.3 ITS 3/4 LfNG CONDONS FOR OPZRAONADDSURVYUANCE RZErQUI 4NTm 3143 RFUElNG OPERATIONS SR 3.9.3.1
- b. Toting oe Popge nd Edtaut isolsta valv per to swamble pau of SR 3.9.3.2 Specitxi4.63.1 2 COOK NUCLEAR PIANT-UNWT 2 Pap 3/4 94a AbMNDMENT 97, 34,242 I
Page 6 of 8 Attachment 1, Volume 14, Rev. 1, Page 56 of 188
Attachment 1, Volume 14, Rev. 1, Page 57 of 188 ITS 3.9.3 ITS 0 3/4 LMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4.6 CONTAINMENT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) 4.6.3.1.2 Each containment isolation valve specified shall be demonstrated OPERABLE at least r See IITSI once per 18 months by: L6 J I a. Verifying that on a Phase A containment Isolation test signal, each Phase A isolation valve actuates to its isolation position.
- b. Verifying that on a Phase B containment isolation test signal, each Phase B isolation valve actuates to Its Isolation position.
3.93d C.poe that on a Cotailnment Purge andoeExhaust isolation signal, each Purge and93. . vR SR 3.9.3.2 c. Verifying and Exhaust valve actuates to its isolation position, culo iuae e The isolation time of each power operated or automaic containment isolation valve shall be See 3 3 4 determined to be within its limit when tested pursuant to Specification 4.0.5. ) L4 COOK NUCLEAR PLANT-UNIT 2 Page 3/4 6-14 AMENDMENT 97A,4,458.465, OU, 257 I Page 7 of 8 Attachment 1, Volume 14, Rev. 1, Page 57 of 188
Attachment 1, Volume 14, Rev. 1, Page 58 of 188 ITS 3.9.3 ITS COWW M1 Wl S?2oW821 LCO 3.9.3.c.2 M.9. SUe COC&UeaUea tw~t ad Xzhabua LUelat:1 Syfl shall to APFlJCAL22LT:Pu1ngr uwswinmt of LmdLaxed fuel vitlllu the Cesautalse:. Vith the Cougainaft ?wgs and Xxtnst £afiatLoa xyn Lnopenble. Blase LCO0 3.9.3.c.1 eatch of the 1uXe and fthout ass~nertimsmwdis dLm. aaess from ths SR 3..3.2 a.. 9'tal~ Cotwsyordto Ch Oansd bau:£eat sl b 333.. SR .9..2 16 .1. = Zdnatbaltif TMC4=&I=t Iuk damtraed im~looto P=Az* f sdi t 19KbL) WUc IMCIUel PUeE - VM~2- 3/4 9-9 AMMMT NO- 07,O), 151 Page 8 of 8 Attachment 1, Volume 14, Rev. 1, Page 58 of 188
Attachment 1, Volume 14, Rev. 1, Page 59 of 188 DISCUSSION OF CHANGES ITS 3.9.3, CONTAINMENT PENETRATIONS ADMINISTRATIVE CHANGES A.1 In the conversion of the CNP Current Technical Specifications (CTS) to the plant specific Improved Technical Specifications (ITS), certain changes (wording preferences, editorial changes, reformatting, revised numbering, etc.) are made to obtain consistency with NUREG-1431, Rev. 2, "Standard Technical Specifications-Westinghouse Plants" (ISTS). These changes are designated as administrative changes and are acceptable because they do not result in technical changes to the CTS. A.2 CTS 3.9.4.b requires a minimum of one door in each airlock to be closed or allows both airlock doors to be open provided one door in each airlock is OPERABLE, refueling cavity level is greater than 23 feet above the fuel, and a designated individual is available at all times to close the airlock if required. A footnote associated with CTS 3.9.4.b clarifies that for the purpose of this Specification, an OPERABLE air lock door is a door that is capable of being closed and secured. ITS 3.9.3 requires that one door in each air lock is capable of being closed. This changes the CTS by replacing the prescriptive requirements for control of the air lock doors with a more general requirement that the air lock doors must be capable of being closed. Other aspects of this change are discussed in DOC A.3 and DOC LA.1. This change is acceptable because the CTS requirements have not changed. A door that is closed is a door that is also capable of being closed. The ITS requirements preserve the intent of the CTS. This change is designated as administrative because it does not result in technical changes to the CTS. A.3 CTS 3.9.4.b.2.b allows both airlock doors to be open provided, in part, that the refueling cavity level is greater than 23 feet above the fuel. ITS 3.9.3 does not contain this restriction. This change is acceptable because the requirement is duplicative of the requirements of ITS LCO 3.9.6, which requires that refueling cavity water level be maintained > 23 feet above the top of the reactor vessel flange during movement of irradiated fuel assemblies within containment. This change is designated as administrative because it does not result in technical changes to the CTS. A.4 The CTS 3.9.4 and CTS 3.9.9 Actions state "The provisions of Specification 3.0.3 are not applicable." ITS 3.9.3 does not include this statement. This changes CTS by deleting the Specification 3.0.3 exception. This change is acceptable because the technical requirements have not changed. ITS LCO 3.0.3 is not applicable in MODE 6. Therefore, the CTS LCO 3.0.3 exception is not needed. This change is designated as administrative because it does not result in a technical change to the CTS. MORE RESTRICTIVE CHANGES None CNP Units I and 2 Page 1 of 4 Attachment 1, Volume 14, Rev. 1, Page 59 of 188
Attachment 1, Volume 14, Rev. 1, Page 60 of 188 DISCUSSION OF CHANGES ITS 3.9.3, CONTAINMENT PENETRATIONS RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES LA.1 (Type 3 - Removing Procedural Details for Meeting TS Requirements or Reporting Requirements) CTS 3.9.4.b.2.c allows both doors of each airlock to be open provided, in part, that a designated individual is available at all times to close an airlock door if required. A footnote associated with CTS 3.9.4.b clarifies that for the purpose of this Specification, an OPERABLE airlock door is a door that is capable of being closed and secured. The footnote also states that cables or hoses transversing the airlock shall be designed to allow for removal in a timely manner (e.g., quick disconnects). ITS 3.9.3.b requires that one door in each air lock is capable of being closed, but does not provide the level of description provided in the CTS. This changes the CTS by moving the requirement for a designated individual and the details on cables or hoses that transverse the air lock from the CTS to the Bases. The removal of these details for compliance with the LCO from the Technical Specifications is acceptable because this type of information is not necessary to be included in the Technical Specifications to provide adequate protection of public health and safety. The ITS still retains the requirement that the one door in each air lock be capable of being closed. Also, this change is acceptable because these types of procedural details will be adequately controlled in the ITS Bases. Changes to the Bases are controlled by the Technical Specification Bases Control Program in Chapter 5. This program provides for the evaluation of changes to ensure the Bases are properly controlled. This change is designated as a less restrictive removal of detail change because procedural details for meeting Technical Specification requirements are being removed from the Technical Specifications. LESS RESTRICTIVE CHANGES L.1 (Category 2 - Relaxation of Applicability) CTS 3.9.4 and CTS 3.9.9 are applicable during CORE ALTERATIONS and movement of irradiated fuel within the containment. ITS 3.9.3 is applicable during movement of irradiated fuel assemblies within containment. References to CORE ALTERATIONS in CTS 3.9.4 are eliminated in the Applicability, Action, and Surveillances. References to CORE ALTERATIONS in CTS 3.9.9 are eliminated in the Applicability and Surveillances. This changes the CTS by eliminating requirements for containment closure and the Containment Purge and Exhaust Isolation System during CORE ALTERATIONS. The purpose of CTS 3.9.4 is to ensure the containment penetrations are in the condition assumed in the Fuel Handling Accident (FHA) inside containment analysis. The purpose of CTS 3.9.9 is to ensure the containment purge supply and exhaust valves are capable of being closed as assumed in the FHA inside containment analysis. This change is acceptable because the requirements CNP Units I and 2 Page 2 of 4 Attachment 1, Volume 14, Rev. 1, Page 60 of 188
Attachment 1, Volume 14, Rev. 1, Page 61 of 188 DISCUSSION OF CHANGES ITS 3.9.3, CONTAINMENT PENETRATIONS continue to ensure that the structures, systems, and components are maintained in the MODES and other specified conditions assumed in the safety analyses and licensing basis. There are no accidents postulated to occur during CORE ALTERATIONS that result in significant radioactive release except a FHA. The analysis for a FHA assumes that the accident is initiated only by movement of irradiated fuel. Therefore, imposing requirements during CORE ALTERATIONS in addition to during movement of irradiated fuel is unnecessary. This change is designated as less restrictive because the ITS LCO requirements are applicable in fewer operating conditions than in the CTS. L.2 (Category 7- Relaxation Of Surveillance Frequency, Non-24 Month Type Change) CTS 4.9.4 states that specified containment penetration Surveillances shall be performed, in part, "within 100 hours prior to the start of" the specified conditions in the Applicability. ITS SR 3.9.3.1 and ITS SR 3.9.3.2 do not include the "within 100 hours prior to the start of' Frequency. ITS SR 3.0.1 states "SRs shall be met during the MODES or other specified conditions in the Applicability for the individual LCOs, unless otherwise stated in the SR." Therefore, the ITS requires that the Surveillances must be met prior to the initiation of movement of irradiated fuel. This changes the CTS by eliminating the stipulation that the Surveillances be met within 100 hours prior to entering the conditions specified in the Applicability. The purpose of CTS 4.9.4 is to verify the equipment required to meet the LCO is OPERABLE. This change is acceptable because the new Surveillance Frequency has been evaluated to ensure that it provides an acceptable level of equipment reliability. For CTS 4.9.4, the periodic Surveillance Frequency for verifying containment penetrations are in the required status is acceptable during the conditions specified in the Applicability, and is also acceptable during the period prior to entering the conditions specified in the Applicability. This change is designated as less restrictive because Surveillances will be performed less frequently under the ITS than under the CTS. L.3 (Category 7 - Relaxation Of Surveillance Frequency, Non-24 Month Type Change) CTS 4.9.4 and CTS 4.9.9 include a Surveillance Frequency of "once per 7 days" during conditions specified in the Applicability for performing Surveillance of the Containment Purge Supply and Exhaust System. The ITS SR 3.9.3.2 Frequency for the same requirement is 24 months. ITS SR 3.9.3.2 is also modified by a Note that states that SR 3.9.3.2 is not required to be met for containment purge supply and exhaust valve(s) in penetrations that are closed to comply with LCO 3.9.3.c.1. This changes the CTS by changing the Surveillance Frequency from 7 days to 24 months and adding the Note that the SR is not required to be met for containment purge supply and exhaust valve(s) in penetrations that are closed to comply with ITS LCO 3.9.3.c.1. The purpose of CTS 4.9.4 and CTS 4.9.9 is to verify the equipment required to meet the LCO is OPERABLE. This change is acceptable because the new Surveillance Frequency has been evaluated to ensure that it provides an acceptable level of equipment reliability. Containment purge supply and exhaust valve testing is still required, but at a Frequency consistent with the testing Frequency for containment isolation valves required in MODES 1, 2, 3, and 4. This Frequency provides an appropriate degree of assurance that the valves are CNP Units 1 and 2 Page 3 of 4 Attachment 1, Volume 14, Rev. 1, Page 61 of 188
Attachment 1, Volume 14, Rev. 1, Page 62 of 188 DISCUSSION OF CHANGES ITS 3.9.3, CONTAINMENT PENETRATIONS OPERABLE. When containment purge supply and exhaust valve(s) in penetrations are closed to comply with ITS LCO 3.9.3.c.1, the penetrations are in the expected condition (isolated) to mitigate the effects of a fuel handling accident inside containment. Therefore, there is no need for the actuation signal to reposition the valves to the closed position. This change is designated as less restrictive because Surveillances will be performed less frequently under the ITS than under the CTS. L.4 (Category 6 - Relaxation Of Surveillance Requirement Acceptance Criteria) CTS 4.6.3.1.2.c requires verification of the automatic actuation of the Containment Purge and Exhaust valves on a Containment Purge and Exhaust isolation signal (i.e., a test signal). ITS SR 3.9.3.2 specifies that the signal may be from either an "actual" or simulated (i.e., test) signal. This changes the CTS by explicitly allowing the use of either an actual or simulated signal for the test. The purpose of CTS 4.6.3.1.2.c is to ensure that the containment purge and exhaust valves operate correctly upon receipt of an actuation signal. This change is acceptable because it has been determined that the relaxed Surveillance Requirement acceptance criteria are not necessary for verification that the equipment used to meet the LCO can perform its required functions. Equipment can not discriminate between an "actual," "simulated," or "test" signal and, therefore, the results of the testing are unaffected by the type of signal used to initiate the test. This change allows taking credit for unplanned actuation if sufficient information is collected to satisfy the Surveillance test requirements. The change also allows a simulated signal to be used, if necessary. This change is designated as less restrictive because less stringent Surveillance Requirements are being applied in the ITS than were applied in the CTS. CNP Units 1 and 2 Page 4 of 4 Attachment 1, Volume 14, Rev. 1, Page 62 of 188
Attachment 1, Volume 14, Rev. 1, Page 63 of 188 Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs) Attachment 1, Volume 14, Rev. 1, Page 63 of 188
Attachment 1, Volume 14, Rev. 1, Page 64 of 188 Containment Penetrations cv 3.9 REFUELING OPERATIONS 3.9.0) Containment Penetrations 0 e1-LCO 3.9.(7 The containment penetrations shall be in the following status:f fi)
- a. The equipmenthiatc closed and held in place byjouiol-ts ) a i;
)
- b. One door in each air lock Is#apable of beinc closedtnd
- c. Each penetration providing direct access from the containment atmosphere to the outside atmosphere is either 3.9.24 t 1. Closed by a manual or automatic isolation valve, blind flange, or equivatentpr n
- 2. Capable of being dosed by an OPERABLE Containment Purge and Exhaust ~T~rSystem.
- NOTE -
Penetration flow path(s) providing direct access from the containment 'I atmosphere to the outside atmospheres may be unisolated under administrative controls. I APPLICABILITY: During movement ofl irradiated fuel assemblies within containment.
~4q ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more A.1 Suspend movement of Immediately containment ([reyy irradiated fuel penetrations not In assiemblies within required status. containment.
Acb WOG STS 3.9A4-1 Rev. 2. 04/30/01 Attachment 1, Volume 14, Rev. 1, Page 64 of 188
Attachment 1, Volume 14, Rev. 1, Page 65 of 188 Containment Penetrations 3.9;p 0D SURVEILLANCE REQUIREMENTS -_'_ -
.I SURVEILLANCE FREQUENCY 1
J4 SR 3 .9 k1' Verify each required containment penetration is In the required status. 7days
.I I
II / .2.. e 4.L SR 3.9$--
- NOTE-Not required to be met for containment piiend exhaust valve(s) in penetrations closed to comply yf~i with LCO 3.9g.c.1. fZV . 0 0)
Verify each req red containment purg nd exhaust months valve actuates to the isolation position on an actual or simulated actuation signal.
'4 i.
. .1 WOG STS 3.9.4 - 2 Rev. 2, 04130101
- Attachment 1, Volume.14, Rev..1, Page 65 of 188
Attachment 1, Volume 14, Rev. 1, Page 66 of 188 JUSTIFICATION FOR DEVIATIONS ITS 3.9.3, CONTAINMENT PENETRATIONS
- 1. CNP has analyzed a boron dilution event in MODE 6. Therefore, ISTS 3.9.2 is not included in the ITS and ISTS 3.9.4 is renumbered as ITS 3.9.3.
- 2. The brackets are removed and the proper plant specific information/value is provided.
- 3. Typographical/grammatical error corrected.
- 4. The Note has been modified consistent with the current licensing basis.
- 5. Changes have been made to be consistent with changes made in another Specification and to be consistent with plant specific nomenclature.
- 6. These punctuation corrections have been made consistent with the Writer's Guide for the Improved Standard Technical Specifications, NEI 01-03, Section 5.1.3.
CNP Units 1 and 2 Page 1 of 1 Attachment 1, Volume 14, Rev. 1, Page 66 of 188
Attachment 1, Volume 14, Rev. 1, Page 67 of 188 Improved Standard Technical Specifications (ISTS) Bases Markup and Justification for Deviations (JFDs) Attachment 1, Volume 14, Rev. 1, Page 67 of 188
Attachmen 1 Vo e 1 R* . 1, Pe 68 of 1
-Attachmen't 1, Volume 14, Rte'v. 1, Page 68 of 188 Containment Penetrations B 3.94Y .
B 3.9 REFUELING OPERATIONS.
.1 B 3.9X) Containment Penetrations ..0
.J: BASES
.1 ..I BACKGROUND During movement ofirradiated fuel assemblies within i
containment, a release of fission produc radioactivity within contairiment will be restricted from escaping to the environment when the LCO requirements are met' In.MODES 1, 2. 3, and 4, this is accomplished by maintaining containment OPERABLE as described in LCO 3.6.1,"ContainmenL" In MODE 6, the potential for containment
-q .I pressurization as a result of an accident is not likely; therefore, 4 requirements to Isolate the containment from the outside atmosphere can 1 be less stringent. The LCO requirements are referred to as "containment closure' rather than "containment OPERABILITY." Containment closure means that all potential escape paths are dosed or capable of being 1, closed. Since there Is no potential for containment pressurization, the i Appendix J leakage criteria and tests are not required.
iI The containment serves to contain fission product radioactivity that may I be released from the reactor core following an accident, such that offsite radiation exposures are maintained well within the requirements of 10 CFR 100. Additionally, the containment provides radiation shielding from the fission products that may be present inthe containment atmosphere following accident conditions. The containment equipment hatch, which is part of the containment pressure boundary, provides a means for moving large equipment and-components into and out of containment. During movement of ) irradiated fuel assemblies within containment, the equipment hatch must be held In place by at least 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 II when containment closure Isnot required, the door Interlock mechanism I.
may be disabled, allowing both doors of an air lock to remain open for extended periods wh fr containment entry Is necessary. Int During movement of Irradiated fuel assemblies within containment, containment c sure is required; therefore, the door WOG STS
- B3.9.4-1 Rev. 2, 04/30101 Attachment 1, Volume 14, Rev. 1, Page 68 of 188
Attachment 1, Volume 14, Rev. 1, Page 69 of 188 Containment Penetrations* B 1
- B 3.9.0 0E z4 BASES I
BACKGROUND (continued) interlock mechanism may remain disabled, bu one air lock door must I I always remainoapable of beinolosed. T i . * ) *,' i The requirements for containment penetration closure ensure that a I release of fission product radioactivity within containment will be 4 restricted to within regulaton limits. II The Containment PuM__ st System in LA _ I E T ~Inc exn cefe -n g ae tth
;eettlnaninc exhaust penetrabotio \
mnipre~sy le cl desne~uq Pnebai an8 inch / Wexhau ne tI uring MODES 1.,2, 3 and 4, the two valves In
-- _~iti; sur~ nd exhaust penetrations are s 7-Ao t.4W4FJ..qe osed na t in eac of Ie to m y ns ned interrnittentl edosed autom yb Cs,Jo* the Engin Sa tyFeatures &GSoon he te aSpecification In MODE 5.
C
- jr oswlA- \ In MODE 6 Ia e air exchangers are necessary to conduct refueling.
C 329, nsa rs u this A s) sed form(ES E> ZA^on al or i cedb min accordant )
- I LCO Seerd
.3., SfetureActuation §yedo (ESFAS)-
The minipurge system remain erational in MODE 6, and our valves are also dosed b SFAS. [or]*/ .D <\ The mi rge system Isnot used In MO . All four 8 Inch valves are
.seedd In the closed pos~ition. ] > * .
The(e 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 automatid
*isolation valve, or by a manual isolation valve, blind flange, or
.4 equivalent. Equivalent Isolation methods must be approved and may include use of a material that can provide a temporary, atmospheric .4 1-i
.1 pressure, ventilation barrier for the other containment penetrations during
- irradiated fuel movements(~~ 00 . .
WOG STS B 3.9A -2 Rev. 2, 04/30101 Attachment 1, Volume 14, Rev. 1, Page 69 of 188
-Attachment 1, Volume 14, Rev. 1; Page 70 of 188 Containment Penetrations BASES APPLICABLE DrnCR LER INSo ovement of Irradi d fuel as bie )
-SAFETY
- wihin Continment the most ere radiological 9seque ANALYSES ifrom -uei handlin adde(involving handlin recent! rr ae fueM The fuel handling acc eni tulated event thatinvo yes
- 1 Re, dropinga single Irradiated fuel assembly and
- r uremets o LOG3.9 "Refueling Cavity Water Level," in Q
- 1 onun o iniumdecay time of 100 hours pnor toffrradiated 4, re o ison product radioactivity, subsequent to a fuel handling accident, results in dos~es th-at-are *e~~
guideline values
. specifiedin10CFR Se .. an, ion R .1 3 wil e oo CFRO luesothe N staff appro enin
, pefi a CF' 0 Imi Containment P'enetrations satisfy Criterion 3 of 1 0 CFR 50.36(c)(2)(ii).
iW . -AE NOT;- The allowaAnce to ha containment personnel ai doors open and penetration flow p with direct access from econtainment atmosphere to e outside atmosphere to b nisolated during fu l Yovem ent CORE ALTERATIONSis sed on (1) confirmatory . . a a o a fuel handling accidens approved by the NRC whicif icate acceptable adiol sequences a
' y thou thecontainm o fission productcontrolfu ion is not can and will be prompaciosed following contain t evacuation and that the open penet on(s) can andwill be pro ti closed. The time to mclose such pne ions or combinaon of pe trations shall be Included Inthe cornfirmay ddose calculations.
This LCO limits the consequences of a fuel handtong acciden acnt in containment by limiting the potential escape paths for fission product radioactivity released within containment The L.O requires any penetration providing direct access
* . from the containment atmosphere to the outside atmosphere to be closed except for the OPERABLE containment purge and exhaust penetrations WOG STS B 3.9.4- 3 Rev. 2, 04130101 Attach.ment 1, VOlume 14, Rev. 1, Page 70 of 188
Attachment 1,-Volume,14, Rev. 1, Page 71 of 188 Containment Penetrations _(g) 'I B 3.9V BASES .1 LCO (continued)
.k:v and the containmejIpersonnei air lock4 For the OPERABLE containment purgand exhaust penetrations, this LCO ensures that (2) ,
I, these penetrations are isolable by the Containment Purge and Exhaust EM at n System. The OPERABILITY requirements for this LCO ensure
- 1hittha slinnm:\f, a p.,rna nnA .. a *mnkm rifsle,.r rnc
,ehilq tmc na,4fla,4 in *hn UFSAR can be achieved and, therefore, meet the assumptions used In the (2 -
safety analysis to ensure that releases through the valves are terminated, such that radiological doses are within the acceptance limit. The LCO is modified bya Note allowing penetration flow paths with direct 1;kv access from the containment atmosphere to the outside atmospher to; - )jWe be unisolated underadministrative controls. Administrative controls ensure that 1) appropriate personnel are aware of the open status of the penetration flow path during t;PE Aiomovement of-Irradiated fuel assemblies within containment, and 2) specified Individuals are designated and readily available to isolate the flow path in the event of a fuel handling accident. The containment personnel air lock doors mazy be open during mosvement of Irradiated fuel In the containment nd pro dprovlded that one door Iscapable of being closed Inthe nvent of a fuel handling accident. Should a fuel handling accident occur c0 ff EArT Inside containment, one personnel air lock door will be dosed following an evacuation of containment APPLICABILITY The containment penetration requirements are'applicable during movement oftrg lrrediated fuel assemblies within containment because this Iswhen there Isa 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 condue tpoheotential for a fuel handling accident does not exist. [Additionally, due to radio decay, a fuel handlin ident involving handling recently Irr eiaid fuel (i.e., fuel that ccupied part of a l- .4 critical reaco within the previous ) will result In doJs'That-are wel inthe guideline values s fled In 10 CFR I100 amen without .4 nmen closur abili. Ifherefore, under these conditions no requirements are placed on containment penetation status.
- 1 WOG STS B 3.9.4-4 Rev. 2, 04/30101 k
Attachment 1, Volume 14, Re'v. 1,Page 71 of 188
Attachment 1, Volume 14, Rev. 1, Page 72 of 188 B 3.9.3 0 INSERT I A designated individual shall be available at all times during movement of irradiated fuel to close an air lock door if required. Cables or hoses transversing the air lock shall be designed to allow for removal in a timely manner (e.g., quick disconnects). Insert Page B 3.9.4-4 Attachment 1, Volume 14, Rev. 1, Page 72 of 188
Attachment 1, Volume 14, Rev. 1, Page 73.of 188
.( )
Containment Penetrations 53 B 3.9f&k- . .BASES
- APPLICABILITY (continued)
- REVIEWER'S NOTE -
The addition of the term "recently" associated with handlin adiated fuel In all of the containment function Technical Specificatio quirements Is only applicable to those licensees who have demonted by analysis that after sufficient radioactive decay has occurr off-site doses resulting from a fuel handling accident rema elow the Standard Review Plan limits (well within I0CFRO . Additionally, licensees adding the m "recently" must make the following commitment which Isconsiste *th draft NUMARC 93-01, Revision 3,
- Section 112.6 "Safety Assd ment for Removal of Equipment from Service During Shutdo Conditions," subheading 'Containment -
Primary (PWR)/Se dary (BWR)"/ iThe following uldelines are included In the assessment of syses removed service during movement irradiated fuel: uring fuel handlingfcore alterations, ventilati system and'. radiation monitor availability (as defined in RC 91-06) should be assessed, with respec to filtration anonitoring of releases j from the fuel. Following shutdown. I ioactivity in the fuel decays away fairly rapidly. The basis of Technical Specification OPERABILITY amendment iss ofe reduction In doses due to such decay. The goal of mainta ng ventilation system and radiation I
- monitor availability is to duce doses even further below that
- provided by the nt decay.
. - A single no or continency method to promptly close imary or secondary tainment penetrations should be develo d. Such prompt ods need notcompletely block the pe ation or be capabof resisting pressure. . The p ose of the "prompt methods" mentio d above are to enable I ve ation systems to draw the release a postulated fuel handling
- ident In the proper direction such it can be treated and . *
.: \~ monitored."/*
WOG STS *1B 3.94- 5 Rev. 2*0413001 Attachment 1, Volume 14, Rev. 1, Page 73 of 188.
Attachment 1, Volume 14, Rev. 1, Page 74 of 188 Containment Penetrations ( BASES ACTIONSA1 n if thedecontainment Thi
. isacmlse equipment hatch, yimdael air locks, orupnigmvmn any containment penetration that provides direct access from the containment atmosphere Iatus, including the rfte ~ndExhustj~!~)Sytemnot irrtadiatedt felqassm blestwatchinokson contim capable nt.aefomnenof oft; autmaicacuatonwhn hepure ehastvalves are open, the~
uolation munction is not() penerations shalnt providesdredue acopetn ofr m ohe mentofatomponent SVLC musafe position. ure on te open pue nea vav wi emon a ~ Surveillance Is performed bery 7 days during movement of irradiated fuel assemblies within containment. The Surveillance Interva i-s selected to be commensurate with the normal duration of time tcompletefelhandlin o oerations. su nce bere es o eTueinfoperatio ns roe o three tcaivent rvelf f ensures _ostuIated fuel handling accidenti invol in
-thecontainmentwl not result Ina release of significant fissio roduct radiowaneity oo.teenvrnetIexss clnoe3recom
- 1 . Surveillance rve~lao openThis that ea on thedemonstrats g n exhaust valve actuates to Its isolation position on mana lt i an actual or simulated high radiation signal. The n maintains consistencybwith omteer similarsF thenormalu tinofU 2tho f hOe evl n toensurethe bfcha e tOPERA ILIT WOG STS *B 3.9.4- 6 *Rev. 2, 04130101 Attachment 1,Volume 14, Rev.l . Page 74 of 188
Attachment 1, Volume 14, Rev. 1, Page 75 of 188 B 3.9.3 0 INSERT 2 The LCO 3.9.3.c.2 status requirement, which requires penetrations to be capable of being closed by an OPERABLE Containment Purge Supply and Exhaust System, can be verified by ensuring each required Q9 INSERT 3 a small fraction of the guideline values specified in 10 CFR 100 O3 INSERT 4 LCO 3.3.6, "Containment Purge Supply and Exhaust System Isolation Instrumentation," provides additional Surveillance Requirements for the containment purge supply and exhaust valve actuation circuitry. Insert Page B 3.9.4-6 Attachment 1, Volume 14, Rev. 1, Page 75 of 188
Attachment 1, Volume 14, Rev. 1, Page 76 of 188 Containment Penetrations I3 B33.9,0 ( I
- 4. BASES ft SURVEILLANCE REQUIREMENTS (continued)
I
- I during refueling jptions. Every 18 m -a CHANNEL CALIBRATIO perforned. The sy m actuation respon me is\
It demon d every -18 months rng refueling, on a GGERED i TEST $15. SR 3.6.3.5sle5fionstrates that the ion time of each va is In accordac_ 1 hese Survellances performed during MODE 6 will nsure that the v are capable of closing after a postulated fuel
- 14 handling accdnen t to limit a release of fission prdu adioactivity 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 capability. . . -Wj eaN
. . k),
O.O. I0 -. i .4 V1~ IOG STS B 3.9A - 7 Rev. 2,04130101 Attachment 1, Volume 14, Rev. 1, Page 76 of 188
Attachment 1, Volume 14, Rev. 1, Page 77 of 188 JUSTIFICATION FOR DEVIATIONS ITS 3.9.3 BASES, CONTAINMENT PENETRATIONS
- 1. Changes are made to reflect consistency with or those changes made to the Specification. Subsequent requirements are renumbered or revised, where applicable, to reflect the changes.
- 2. The brackets have been removed and the proper plant specific information/value has been provided.
- 3. Changes are made (additions, deletions, and/or changes) to the ISTS Bases which reflect the plant specific nomenclature, number, reference, system description, analysis, or licensing basis description.
- 4. The reference to a Fuel Handling Accident being initiated by CORE ALTERATIONS or the dropping of a heavy object onto irradiated fuel assemblies is deleted from the Applicable Safety Analyses section of the Bases. CORE ALTERATIONS other than irradiated fuel movement inside containment and dropping of a heavy object onto irradiated fuel assemblies are not assumed to initiate a Fuel Handling Accident. Only the dropping of an irradiated fuel assembly is assumed to initiate a Fuel Handling Accident.
- 5. Changes have been made to be consistent with the ISTS.
- 6. Typographical/grammatical error corrected.
- 7. Editorial change for clarity.
CNP Units 1 and 2 Page 1 of 1 Attachment 1, Volume 14, Rev. 1, Page 77 of 188
Attachment 1, Volume 14, Rev. 1, Page 78 of 188 Specific No Significant Hazards Considerations (NSHCs) Attachment 1, Volume 14, Rev. 1, Page 78 of 188
Attachment 1, Volume 14, Rev. 1, Page 79 of 188 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.9.3, CONTAINMENT PENETRATIONS There are no specific NSHC discussions for this Specification. CNP Units I and 2 Page 1 of 1 Attachment 1, Volume 14, Rev. 1, Page 79 of 188
Attachment 1, Volume 14, Rev. 1, Page 80 of 188 ATTACHMENT 4 ITS 3.9.4, Residual Heat Removal (RHR) and Coolant Circulation
- High Water Level Attachment 1, Volume 14, Rev. 1, Page 80 of 188 , Volume 14, Rev. 1, Page 81 of 188 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) , Volume 14, Rev. 1, Page 81 of 188
Attachment 1, Volume 14, Rev. 1, Page 82 of 188 ITS 3.9.4 ITS 0
, VC V!?, ~'. !6 ve 6 ? LATe M.
LCO 3.9.4 .9.1.1 At least one r"Sdual beat removal loop shll be n operation. A ) Zthewater ith level 2 23 R above the top ofIhrece UPY 1TW: IIOD1 $. [yesselflange 1g ACTION A LCO 3.9.4 Note SR 3.9.4.1 4.9.8.1 A residual beat remoal lto shall be detemined to be in operation ad eirculatin# roter eto a t a flow rate of grater then ow equal to 2000 Du at least one par b I 12 lor wurpoese of this apeclfiSction. addition of water from the IWSS does ' sot Genatituts a dilution asstity ptrovded the bOroe Concentration In the IVT Li Sreater than or equal to the idnnm required by specification Ij1 . % , . C. COOK
- MT I 3/4 9-9 AKODKOT 5-. 120 Page 1 of 2 Attachment 1, Volume 14, Rev. 1, Page 82 of 188
Attachment 1, Volume 14, Rev. 1, Page 83 of 188 ITS 3.9.4 ITS 1TT!O OFIATIOWS g .3P!ID26Lt M~ET ItW~tOaL AIM COOTAIIT CThCUTATION L1,11TnG TTIOPERABLE enTdO 3.9.1.1 At least one residjual beat resuival l2. shall b in operation. LCO 3.9.4 AflLT(ITAII2: MtODE
- vsefange a.
ACTION A LCO 3.9.4 Note 1c. 7_U*-,f-roYL21ooz'of 5jwWff1c&tL*n 3.0.3_arg-'not applicable.1 Clmi METU"CE ItTOMWENTN" . 4.9.1.1 A residual eat remvl, loop shall be determined io be in operation ani circulating reactor coolant at a flow rate of greater than or equal to SR 3.9.4.1 2000 gpM at lsait once per Ihours 3P.M4) 12
'e. For purposes of this specifcatieon. addition of water fr'om the RUST does not constitute a dilution activity provided the baron conieneration in the RWST is greater ehan or equal to theinium required by specificeation L I .1.2.7.b.2. A I -G D. C. COOK - UNIT 2 3/4 9.8 AMENDtIMT 1(O. BZ,107 Page 2 of 2 Attachment 1, Volume 14, Rev. 1, Page 83 of 188
Attachment 1, Volume 14, Rev. 1, Page 84 of 188 DISCUSSION OF CHANGES ITS 3.9.4, RESIDUAL HEAT REMOVAL (RHR) AND COOLANT CIRCULATION - HIGH WATER LEVEL ADMINISTRATIVE CHANGES A.1 In the conversion of the CNP Current Technical Specifications (CTS) to the plant specific Improved Technical Specifications (ITS), certain changes (wording preferences, editorial changes, reformatting, revised numbering, etc.) are made to obtain consistency with NUREG-1431, Rev. 2, "Standard Technical Specifications-Westinghouse Plants" (ISTS). These changes are designated as administrative changes and are acceptable because they do not result in technical changes to the CTS. A.2 CTS 3.9.8.1 requires at least one residual heat removal loop to be in operation in MODE 6. ITS 3.9.4 requires one RHR loop to be OPERABLE and in operation in MODE 6 with the water level greater than or equal to 23 feet above the top of the reactor vessel flange. However, ITS 3.9.5 covers the Applicability of MODE 6 with water level less than 23 feet above the top of the reactor vessel flange. This changes the CTS by splitting the requirements associated with CTS 3.9.8.1 into two Applicabilities, one for MODE 6 with water level < 23 feet above the top of the reactor vessel flange, and one for MODE 6 with water level greater than of equal to 23 feet above the reactor vessel flange. The purpose of CTS 3.9.8.1 is to ensure that adequate decay heat removal capability is in operation and that the coolant is circulated in MODE 6. This change is acceptable because the requirements continue to ensure that the process variables are maintained in the MODES and other specified conditions assumed in the safety analyses and licensing basis. MODE 6 RHR and coolant circulation requirements are governed by ITS 3.9.4, "Residual Heat Removal (RHR) and Coolant Circulation - High Water Level," and ITS 3.9.5, "Residual Heat Removal (RHR) and Coolant Circulation - Low Water Level." The combination of ITS 3.9.4 and ITS 3.9.5 ensures that the appropriate RHR loops are available in MODE 6 regardless of the water level. This change is designated as administrative because it does not result in technical changes to the CTS. A.3 CTS 3.9.8.1 Action a states, in part, that with less than one RHR loop in operation, suspend all operations involving an increase in the reactor decay heat load of the Reactor Coolant System. ITS 3.9.4 Required Action A.2 states, in part, that with the RHR loop requirements not met, suspend loading irradiated fuel assemblies in the core. This changes the CTS by requiring that the loading of irradiated fuel assemblies be suspended instead of requiring that all operations involving an increase in the reactor decay heat load be suspended. This change is acceptable because the requirements have not changed. The reactor decay heat load is generated only by irradiated fuel. The only method of increasing the decay heat load of a reactor in MODE 6 is to load additional irradiated fuel assemblies into the core. Therefore, the CTS and ITS requirements are equivalent. This change is designated as administrative because it does not result in technical changes to the CTS. CNP Units I and 2 Page 1 of 5 Attachment 1, Volume 14, Rev. 1, Page 84 of 188
Attachment 1, Volume 14, Rev. 1, Page 85 of 188 DISCUSSION OF CHANGES ITS 3.9.4, RESIDUAL HEAT REMOVAL (RHR) AND COOLANT CIRCULATION - HIGH WATER LEVEL A.4 CTS 3.9.8.1 Action c states "The provisions of Specification 3.0.3 are not applicable." ITS 3.9.4 does not include this statement. This changes CTS by deleting the Specification 3.0.3 exception. This change is acceptable because the technical requirements have not changed. ITS LCO 3.0.3 is not applicable in MODE 6. Therefore, the CTS LCO 3.0.3 exception is not needed. This change is designated as administrative because it does not result in a technical change to the CTS. MORE RESTRICTIVE CHANGES M.1 CTS 3.9.8.1 requires that at least one residual heat removal loop be in operation. ITS 3.9.4 requires that one RHR loop shall be OPERABLE and in operation. This changes the CTS by requiring the RHR loop to be OPERABLE, instead of just in operation. The purpose of CTS 3.9.8.1 is to ensure that adequate decay heat removal and coolant circulation are available in MODE 6. However, the CTS LCO could be interpreted as allowing an RHR loop to be placed in operation that was not OPERABLE. The ITS eliminates this possible misinterpretation. This change is acceptable because the RHR loop must be OPERABLE (i.e., capable of performing its safety function) instead of just being in operation. This change is designated as more restrictive because the ITS contains more specific requirements on a component. M.2 The CTS 3.9.8.1 Actions do not include an action to immediately initiate action to satisfy the RHR loop requirements in the event the RHR loop requirements are not met. ITS 3.9.4 Required Action A.3 requires that action be immediately initiated to satisfy the RHR loop requirements. This changes the CTS by requiring that action be taken immediately to satisfy the RHR loop requirements. The purpose of CTS 3.9.8.1 is to ensure that adequate decay heat removal and coolant circulation are available in MODE 6. Although decay heat is removed from the Reactor Coolant System via natural circulation to the bulk of water contained in the refueling canal, this method of heat transfer can continue for only a discrete amount of time before boiling would occur. This change is acceptable because it requires that action be initiated to restore the RHR loop requirements in order to restore forced coolant flow and heat removal. This change is designated as more restrictive because additional actions will be required in the ITS than are required in the CTS. M.3 CTS 3.9.8.1 Action b states that the RHR loop may be removed from operation for up to 1 hour per 8 hour period during the performance of CORE ALTERATIONS in the vicinity of the reactor pressure vessel hot legs. The ITS LCO 3.9.4 Note states that the required RHR loop may be removed from operation for < 1 hour per 8 hour period, provided no operations are permitted that would cause introduction into the Reactor Coolant System, coolant with boron concentration less than that required to meet the minimum required boron concentration of LCO 3.9.1, "Boron Concentration." This results in two changes CNP Units 1 and 2 Page 2 of 5 Attachment 1, Volume 14, Rev. 1, Page 85 of 188
Attachment 1, Volume 14, Rev. 1, Page 86 of 188 DISCUSSION OF CHANGES ITS 3.9.4, RESIDUAL HEAT REMOVAL (RHR) AND COOLANT CIRCULATION - HIGH WATER LEVEL to the CTS. First, the allowance to remove RHR from operation is no longer restricted to CORE ALTERATIONS in the vicinity of the reactor pressure vessel hot legs. Second, the use of the allowance in the ITS is predicated on prohibiting operations that would cause introduction into the RCS, coolant with a boron concentration less than that required to meet the boron concentration of LCO 3.9.1. This change is acceptable because it applies appropriate controls during periods when RHR is not in operation. The ITS requirement prohibiting operations which would cause a reduction in the RCS boron concentration below that required to maintain the required shutdown margin is necessary to avoid unexpected reactivity changes. Under the ITS definition of CORE ALTERATIONS, many activities that would be considered CORE ALTERATIONS in the CTS, such as core mapping, are not considered CORE ALTERATIONS in the ITS. Therefore, the application of the allowance is expanded in the ITS to cover other activities beyond CORE ALTERATIONS. This change is nominally less restrictive, but represents no practical operational change, and the overall change is considered more restrictive. This change is designated as more restrictive because it imposes a new condition to be met when an RHR loop is not in operation. M.4 CTS 4.9.8.1 requires that a residual heat removal loop shall be determined to be in operation and circulating reactor coolant at a flow rate of greater than or equal to 2000 gpm at least once per 24 hours. ITS SR 3.9.4.1 requires the same verification every 12 hours. This changes the CTS by requiring that RHR loop operation and reactor coolant flow rate be verified every 12 hours instead of every 24 hours. The purpose of CTS 4.9.8.1 is to ensure that adequate decay heat removal and coolant circulation are available in MODE 6. This change is acceptable since it results in an increased Frequency of performance. The 12 hour Frequency is consistent with similar CTS Surveillances in MODES 4 and 5, and with similar SRs in the ITS. This change is designated as more restrictive because the Surveillance will be performed at an increased Frequency in the ITS. RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES None LESS RESTRICTIVE CHANGES L.1 (Category 4- Relaxation of Required Action) CTS 3.9.8.1 Action a states, in part, that with less than one RHR loop in operation, suspend all operations involving a reduction in boron concentration of the Reactor Coolant System. This CNP Units 1 and 2 Page 3 of 5 Attachment 1, Volume 14, Rev. 1, Page 86 of 188
Attachment 1, Volume 14, Rev. 1, Page 87 of 188 DISCUSSION OF CHANGES ITS 3.9.4, RESIDUAL HEAT REMOVAL (RHR) AND COOLANT CIRCULATION - HIGH WATER LEVEL CTS Action is modified by a footnote which states that addition of water from the RWST does not constitute a dilution activity provided the boron concentration in the RWST is greater than the minimum required by Specification 3.1.2.7.b.2 (i.e., 2400 ppm). ITS 3.9.4 Required Action A.1 states that with the RHR loop requirements not met, suspend operations that would cause introduction into the RCS, coolant with boron concentration less than required to meet the boron concentration of LCO 3.9.1, "Boron Concentration." ITS 3.9.1 requires boron concentration to be within limit. This changes the CTS by allowing coolant with boron concentration less than the RCS boron concentration, but greater than the boron concentration limit in ITS LCO 3.9.1, to be added to the RCS from sources other than the RWST when the RHR requirements are not met. The purpose of CTS 3.9.8.1 Action a is to ensure that the required SHUTDOWN MARGIN is maintained during periods when the RHR requirements are not met. This change is acceptable because the Required Actions are used to establish remedial measures that must be taken in response to the degraded conditions in order to minimize risk associated with continued operation while providing time to repair inoperable features. The Required Actions are consistent with safe operation under the specified Condition, considering the OPERABILITY status of the redundant systems of required features, the capacity and capability of remaining features, a reasonable time for repairs or replacement of required features, and the low probability of an accident occurring during the repair period. The Required Actions ensure that the RCS boron concentration is maintained within the limits of ITS LCO 3.9.1, which is sufficient to ensure that adequate SHUTDOWN MARGIN is maintained. This change is designated as less restrictive because less stringent Required Actions are being applied in the ITS than were applied in the CTS. L.2 (Category 4 - Relaxation of Required Action) CTS 3.9.8.1 Action a states, in part, that with less than one RHR loop in operation, close all containment penetrations providing direct access from the containment atmosphere to the outside atmosphere within 4 hours. ITS 3.9.4 Required Actions A.4, A.5, and A.6 state that with the RHR loop requirements not met, within 4 hours close and secure the equipment hatch with at least four bolts, close one door in each air lock, and verify each penetration providing direct access from the containment atmosphere to the outside atmosphere is either closed with a manual or automatic isolation valve, blind flange, or equivalent, or is capable of being closed by an OPERABLE Containment Purge Supply and Exhaust System. This changes the CTS Actions by allowing penetrations capable of being closed by an OPERABLE Containment Purge Supply and Exhaust System to remain open when the RHR requirements are not met. The purpose of CTS 3.9.8.1 Action a is to ensure that radioactive material does not escape the containment should the RHR requirements continue to not be met and boiling occurs in the core. Therefore, containment penetrations are closed to seal the containment. This change is acceptable because the Required Actions are used to establish remedial measures that must be taken in response to the degraded conditions in order to minimize risk associated with continued operation while providing time to repair inoperable features. The Required Actions are consistent with safe operation under the specified Condition, CNP Units 1 and 2 Page 4 of 5 Attachment 1, Volume 14, Rev. 1, Page 87 of 188
Attachment 1, Volume 14, Rev. 1, Page 88 of 188 DISCUSSION OF CHANGES ITS 3.9.4, RESIDUAL HEAT REMOVAL (RHR) AND COOLANT CIRCULATION - HIGH WATER LEVEL considering the OPERABILITY status of the redundant systems of required features, the capacity and capability of remaining features, a reasonable time for repairs or replacement of required features, and the low probability of an accident occurring during the repair period. The Required Actions are consistent with the actions taken for containment closure in CTS 3.9.4 and ITS 3.9.3. Penetrations which can be closed by an OPERABLE Containment Purge Supply and Exhaust System do not need to be closed if RHR is inoperable, since the presence of radioactivity in the containment will cause the valves to close automatically, thus performing the isolation function. This change is designated as less restrictive because less stringent Required Actions are being applied in the ITS than were applied in the CTS. CNP Units I and 2 Page 5 of 5 Attachment 1, Volume 14, Rev. 1, Page 88 of 188
Attachment 1, Volume 14, Rev. 1, Page 89 of 188 Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs) Attachment 1, Volume 14, Rev. 1, Page 89 of 188
Attachment 1, Volume 14, Rev. 1, Page 90 of 188 crs RHR and Coolant Circulation - High WaterLel 3.9 REFUELING OPERATIONS 3.9.% Residual Heat Removal (RHR) and Coolant Circulation - High Water Level j: 3?.611 LCO 3.9.9 One RHR loop shall be OPERABLE and In operation. 0
) - ---- ----- --- )-__ _ _-__ _ ---- _-__
The required RHR loop may 'e topratio
) s for 1 hour per 8 hour period, provided no operations are permitted that would cause 1iAeA'76- JO ginto the Reactor Coolant Syste concentration less than that required to meet the minimum required 1I
/ boron concentration of LCO 3.9.1i 2
' %bl_ _ __ P______ifl 7 _ 2 APPLICABILITY: MODE 6 with the water level 2 23 ft above the top of reactor vessel flange.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. RHR loop requirements A.1 Suspend operations that Immediatel E MI" 0C. not met. would cause introduction Into the RCSFIaiavnith boron concenation less 0 I than required to meet the boron concentration of LCO 3.9.1. A.2 Suspend loading irradiated Immediately I fuel assemblies Inthe core. II .i 'A A.3 Initiate action to satisfy Immediately RHR loop requirements. AND . WOG STS 3.9.5 -1 Rev. 2, 04130101 Attachment 1, Volume 14, Rev. 1, Page 90 of 188
Attachment 1, Volume 14, Rev..1, Page 91 of 188 CTM RHR and Coolant Circulation - High Water Leev ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME pi A.4 Close equipment hatch 4 hours and secure withpouis bolts..
- , AND A.5 Close one door In each air 4 hours lock.
AND, A.60 (q;)each penetration 4 hours providing direct access from the containment atmosphere to the outside ; c d atmospher t a manual or automatic isolation valve, blind flange, or equivalen 05 (A. V e De~tratios (id) cpa~e O Deng dsdb
. nOPERABLE ; ;ri
- Exhaust sat iSystem.
SURVEILLANCE REQUIREMENTS -_._._: '1 SURVEILLANCE FREQUENCY 41 SR 3.9#1 Verify one RHR loop is In operation and drculating 12 hours C .1 reactor coolant at a flow rate of 2 3gpm. "I.,q98 -
--o WOG STS 3.9.5 -2. Rev. 2, 04130/01 Attachment 1,9 Volume 14,. Rev. 1, Page 91.. of 188
Attachment 1, Volume 14, Rev. 1, Page 92 of 188 JUSTIFICATION FOR DEVIATIONS ITS 3.9.4, RESIDUAL HEAT REMOVAL (RHR) AND COOLANT CIRCULATION - HIGH WATER LEVEL
- 1. CNP has analyzed a boron dilution event in MODE 6. Therefore, ISTS 3.9.2 is not included in the ITS and ISTS 3.9.5 is renumbered as ITS 3.9.4.
- 2. Editorial correction to be consistent with the format of the ITS.
- 3. The brackets are removed and the proper plant specific information/value is provided.
- 4. ISTS 3.9.5 Required Actions A.6.1 and A.6.2 are connected by an "OR" logical connector, such that either one can be performed to meet the requirements of the ACTION. However, the two Required Actions are applicable to all the penetrations; either Required Action A.6.1 or Required Action A.6.2 must be performed for all the penetrations. Thus, this will not allow one penetration to be isolated by use of a manual valve and another penetration to be capable of being closed by an OPERABLE Containment Purge Supply and Exhaust System. This is not the intent of the requirement. The requirement is based on ISTS LCO 3.9.4 (ITS LCO 3.9.3),
which requires each penetration to be either: a) closed by a manual or automatic isolation valve, blind flange, or equivalent; or b) capable of being closed by an OPERABLE Containment Purge Supply and Exhaust System. For consistency with the actual LCO requirement, ISTS 3.9.5 Required Actions A.6.1 and A.6.2 have been combined into a single Required Action in ITS 3.9.4 Required Action A.6.
- 5. Changes have been made to be consistent with changes made in another Specification and to be consistent with plant specific nomenclature.
CNP Units 1 and 2 Page 1 of 1 Attachment 1, Volume 14, Rev. 1, Page 92 of 188
Attachment 1, Volume 14, Rev. 1, Page 93 of 188 Improved Standard Technical Specifications (ISTS) Bases Markup and Justification for Deviations (JFDs) Attachment 1, Volume 14, Rev. 1, Page 93 of 188
Attachment 1, Volume 14, Rev. 1, Page 94 of 188 RHR and Coolant Circulation - High Water Levet B 3.9 B 3.9 REFUELING OPERATIONS B 3.9j Residual Heat Removal (RHR) and Coolant Circulation - High Water Level 0 'S BASES BACK( 3ROUND The purpose of the RHR System In MODE 6 is to remove deca heat and sensibl heat from the Reactor Coolant System (RCS as C p_ to provide mixing of borated coolant and to prevent boron I ' stratification (Ref. .1). Heat Is removed from the RCS by circulating reactor coolant through the RHR heat exchanger(s), where the heat Is transferred to the Component Cooling Water System. The coolant is then returned to the RCS via the RCS cold leg(s). Operation of the RHR System for normal cooldown or decay heat removal Is manually accomplished from the control room. The heat removal rate Is adjusted by controlling the flow of reactor coolant through the RHR heat 0 r AJ _SE angers and the bypas Mixing of the reactor coolant Is maintained by this continuous crculation of reactor coolant through the RHR System.
' APPLICA BLE If the reactor coolant temperature Is not maintained below 2000F, boiling SAFETY of the reactor coolant could result. This could lead to a loss of coolant In ANALYSI ZS the reactor vessel. Additionally, boiling of the reactor coolant could lead to a reduction In boron concentration in the coolant due to boron plating out on components near the areas of the boiling activity. The loss of reactor'coolant and the reduction of boron concentration in the reactor coolant would eventually challenge the integ of the fuel caddin ,) ()
which is a fission product barrier. On t H ter is required to be operational In MODE 6, with the water level 2 23 ft above Ithe to of the reactor vessel flanee. to orevent this challenge. The _C osp p rH lor short durations, under the conditon that the boron concentration Is not diluted. This conditional
- f the RHR pump does not result in a challenge to the fission product barrier.
The RHR System satisfies Criterion 4 of 10 CFR 50.36(c)(2)(ii). LCO Only one RHR loop is required for decay heat removal In MODE 6, with the water level z 23 ft above the top of the reactor vessel flange. Only.
.1I one RHR loop Is required to be OPERABLE, because the volume of 4 water above the reactor vessel flange provides backup decay heat i
to removal capability. At least one RHR loop must be OPERABLE and in operation to provide:* i i.1 a. Removal of decay hea 4 WOG STS - B3.9.5 -1 Rev. 2, 04130/01 Attachment 1, Volume.14, Rev. 1, Page 94 of 188
Attachment 1, Volume 14, Rev. 1, Page 95 of 188 B 3.9.4 (i)? INSERT 1 , as well as adjustments in Component Cooling Water System temperature and flow Insert Page B 3.9.5-1 Attachment 1, Volume 14, Rev. 1, Page 95 of 188
Attachment 1, Volume 14, Rev. 1, Page 96 of 188 RHR and Coolant Circulation - High Water Lev i B 3.9P - -i 1 BASES I LCO (continued)
- b. Mixing of borated coolant to minimize the possibility of criticality I
(c catioof readt coantteper i I An OPERABLE RR loop includes an RHR pump, a heat exchanger, II - valves, piping, Instruments, and controls to ensure an OPERABLE flow pathin eterine the en D D. The flow path starts in 0-11% U7 one nofth RCM hot len- nnd Is rpthimpd Ln the RCS cold leas. The ICO Is modfed by a Note that allows the required operating RHR
/T7' loopip5.loop er6b operation forupto1 hourper8hourperiodprovidedno ermitted that would dilute the RCS boron concentration introdud onof coolant into the RCS with boron concentration less than required to meet the minimum boron concentration of LCO 3.9.10 ron Up
. concentration reduction with coolant at boron concentrations less than required to assure the RCS boron concentration Ismaintained Is 4or"P prohibited because uniform concentration distribution cannot be ensured \
without forced circulation. This permits operations such as core mapping or alterations in the vicinity of the reactor vessel hot leg nozzles and RCS to RHR isolation valve testing. During this 1 hour period, decay heat Is removed by natural convection to the large mass of water In the refueling cavity. APPLICABILIT Y One RHR loop must be OPERABLE and In operation in MODE 6, with the water level i 23 ft above the top of the reactor vessel flange, to provide decay heat removal. The 23 ft water level was selected because It
,ad3 corresponds to the 23 ft requirement established for fuel movement In 9, "Refueling Cavity Water Level." Requirements f ae RHR System In other MODES sovered by LCOs in Secton 3. ea Vor
. (D CooantSvtem(R C nd SORIFY . tr rencre codinaS p requirements In MODE 6 with the water cad InLCO 3.9Q, Residual Heat Removal (RHR)
(0 .00 and Coolant Circulation - Lo Water evel.": ACTIONS RHR loop requirements are met by having one RHR loop OPERABLE and In operation, except as permitted In the Note to the LCO. 3,
.I A.
I If RHR loop requirements are not met, there will be no forced circulation to provide mixing to establish uniform boron concentrations. Suspending positive reactivity additions that could result In failure to meet the WOG STS B3.9.5 -2 Rev. 2, 04130101 Attachment 1, Volume 14, Rev. 1, Page 96 of 188
Attachment 1, Volume 14, Rev. 1, Page 97 of 188 RHR and Coolant Circulation - High Water Leve - B 3.99 BASES ACTIONS (continued) minimum 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 what would be required in the RCS for minimum refueling boron concentration. This may result in an overall reduction in RCS boron concentration, but provides acceptable margin to maintaining subcritical operation.
A.2 If RHR loop requirements are not met, actions shall be taken immediately to suspend loading of Irradiated fuel assemblies In the core. With no forced circulation cooling. decay heat removal from the core occurs by natural convection to the heat sink provided by the water above the core. A minimum refueling water level of 23 ft above the reactor vessel flange provides an adequate available heat sink. Suspending any operation that w would Increase decay heat load, such as loading a fuel assembly, Is a prudent action under this condition. A.3 If RHR loop requirements are not met, actions shall be initiated and continued In order to satisfy RHR loop requirements. With the unit in
- j MODE 6 and the refueling water level 2 23 ft above the top of the reactor vessel flange, corrective actions shall be initiated immediately.
If no RHR Is In operation, the following actions must be taken:
- a. The equipment hatch must be closed and secured withtoufbol (;)
- b. One door In each air lock must be close
- c. Each penetration providing direct access from the containment (2) atmosphere to the outside atmosphere must be either closed by a manual or automatic isolation valve, blind flange, or equivalent, or 4 verified to be capable of being closed by an OPERABLE Containment Purg angxhaustjgSystem.
With RHR loop requirements no met, the potential exists for the coolant to boil and release radioactive gas to the containment atmosphere. 4 Performing the actions described above ensures that all containment WOG STS B 3.9.5 - 3 Rev. 2, 04/30/01 Attachment 1, Volume 14, Rev. 1, Page 97 of 188
Attachment 1, Volume 14, Rev. 1, Page 98 of 188 RHR and Coolant Circulation - High Water LevelD B 3.9 BASES ACTIONS (continued) penetrations are either dosed or can be closed so that the dose limits are not exceeded. The Completion Time of 4 hours allows fixing of most RHR problems and is reasonable, based on the low probability of the coolant boiling In that I time. SURVEILLANCE 9 REQUIREMENTS This Surveillance demonstrates that the RHR loop is In operation and circulating reactor coolant. The flow rate Isdetermined by the flow rate necessary to provide sufficient decay heat removal capability and to prevent thermal and boron stratification Inthe core. The Frequency of
- 1 12 hours Issufficient, considering the flow, temperature, pump control.
and alarm Indications available to the operator in the control room for monitoring the RHR System. REFERENCES 1. ;SAR, Section( -D
'4 4 '4 A1 i
WOG STS B 3.9.5 - 4 Rev. 2,04130101 Attachment 1, Volume 14, Rev. 1, Page 98 of 188
Attachment 1, Volume 14, Rev. 1, Page 99 of 188 JUSTIFICATION FOR DEVIATIONS ITS 3.9.4 BASES, RESIDUAL HEAT REMOVAL (RHR) AND COOLANT CIRCULATION
- HIGH WATER LEVEL
- 1. Changes are made to reflect those changes made to the ISTS. Subsequent requirements are renumbered or revised, where applicable, to reflect the changes.
- 2. CNP Units 1 and 2 were designed and under construction prior to the promulgation of 10 CFR 50, Appendix A. CNP Units 1 and 2 were designed and constructed to meet the intent of the proposed General Design Criteria, published in 1967.
However, the CNP UFSAR contains discussions of the Plant Specific Design Criteria (PSDCs) used in the design of CNP Units 1 and 2. Bases references to the 10 CFR 50, Appendix A, criteria have been replaced with references to the appropriate section of the UFSAR.
- 3. These punctuation corrections have been made consistent with the Writer's Guide for the Improved Standard Technical Specifications, NEI 01-03, Section 5.1.3.
- 4. Changes are made (additions, deletions, and/or changes) to the ISTS Bases which reflect the plant specific nomenclature, number, reference, system description, analysis, or licensing basis description.
- 5. The wording has been modified, as Section 3.5 does not provide requirements for the RHR Shutdown Cooling function.
- 6. The brackets have been removed and the proper plant specific information/value has been provided.
- 7. Changes have been made to be consistent with the ISTS.
- 8. Typographical/grammatical error corrected.
CNP Units 1 and 2 Page 1 of I Attachment 1, Volume 14, Rev. 1, Page 99 of 188
Attachment 1, Volume 14, Rev. 1, Page 100 of 188 Specific No Significant Hazards Considerations (NSHCs) Attachment 1, Volume 14, Rev. 1, Page 100 of 188
Attachment 1, Volume 14, Rev. 1, Page 101 of 188 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.9.4, RESIDUAL HEAT REMOVAL (RHR) AND COOLANT CIRCULATION - HIGH WATER LEVEL There are no specific NSHC discussions for this Specification. CNP Units 1 and 2 Page 1 of 1 Attachment 1, Volume 14, Rev. 1, Page 101 of 188
Attachment 1, Volume 14, Rev. 1, Page 102 of 188 ATTACHMENT 5 ITS 3.9.5, Residual Heat Removal (RHR) and Coolant Circulation
- Low Water Level Attachment 1, Volume 14, Rev. 1, Page 102 of 188 , Volume 14, Rev. 1, Page 103 of 188 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) , Volume 14, Rev. 1, Page 103 of 188
Attachment 1, Volume 14, Rev. 1, Page 104 of 188 ITS 3.9.5 ITS LCO 3.9.5 ACTION B bTheresieual heat removal loop m*y be remv4 from oera:tLo See ITS ) to 1 bVr lt r I bMWr Period *fiq tbe pOrtOrMMs .f CCU 1 3.9.4 g 4ALZAU OSO In the VIeLaUtY Of the reaater psessru onsaul1 IC. Tho-%rwtstJons of SJ ltagson 3.0.3 axwhot swelleoble. SR 3.9.5.1 .1.3.1 A resodual beat remoevI ltoo shall be daeermined to be in operation ad Ciraulata r
- e n *t a fa r .t oefrgrecer than or equal to 000 wea* la aours. tete per I 12
,or. pupoeer of this specifieation. addition of not eenstitute a dilution aetivity provided the water from the IWS? does beren ceeoentrattin in ch RM Is Soa2er than atr equal to the mInium required by specification I qI . O.
1D. C. COOK
- UWI I 3/4 9.9 AKCIWUT 110. 120 Page 1 of 4 Attachment 1, Volume 14, Rev. 1, Page 104 of 188
Attachment 1, Volume 14, Rev. 1, Page 105 of 188 ITS 3.9.5 ITS I T::U' CPPA 'Is I *d**!r L!'V-.
','S':'G 0':!^tF:1 CPEU.71C LCO 3.9.5 3.9.8.2 Two irnependent Residual Hat RIMVl (RMR) loops shall be OPEUA3LE I Add nnrx)sMd LUC Note 2 - -
_-0
- I L.
VOI8fCA L"fl : HOE 8 when the water level above tie top o1 r h Eact pressure vessel flango Is less than 23 feet. AMN:CN ACTION A a. With less than the required RJR loops OPEPAILE, imediately Inititns corrsettnv action to return the required RHR loops to OPEMILE 1-A.6 status as soon U Possible. AddproposedRequlredActionA.2 I b. TWhtrovisions of Specfifcation 3.0.3_.7(not aun1lCADu.. . A. e .~lI....e see~su
>U.
- LW ;'^t.UI RL I.9.8.2,,e4ocuirtd Residual val loops
.. shall to Ined -OFRAOL .-
cr Sooc lu:1n 4.ri0.5. LA 4 I Add proposed SR 3.9.5.2 end Note - M.3
!j I' noraw cr . rgercE rwifi-rce way be inoerse!Abfor each ;R locol-II
- - .. ,,_ ,..MT 3;:1 9a jArnd rnt '40.78 Page 2 of 4 Attachment 1, Volume 14, Rev. 1, Page 105 of 188
Attachment 1, Volume 14, Rev. 1, Page 106 of 188 ITS 3.9.5 ITS 3/. 93 RESIDUtAL tALAMTR40VAt tAD COOLANT CTIATTN LtHT'NQ CO TT rat OP"!ATlOL LCO 3.9.5 3.9.1.1 At least ues reslsual hest removal loop shell be In cperation A vewith thewater l level < 23 ft above the top ACTIN L--- yessel flange With less than ona residual heat I ACTION B tSee3TS 3.9.4 Ic. rovisiousof Sperdfication 3.0.3 eArtnot aplieable CIOIJT tt Y AV w . SR 3.9.5.1 4.9.8.1 A residual beat removal loop shall be determined to be in operation and ctrculattig reactor toolant at a flow rate of greater then or equal to 2000 gpm at least once per hours. For purposes of this specification. addition of vater froa the RWST does noc conscictuo a dilution activity provided the boron conetncration in the RUS.b is grecaer ehan or equal to the mipimua required by specification. 1 1.L..7.b.2. D. C. COOK - UITh 2 3/6 9-8 AMMWKENT NO. $Zs 107 Page 3 of 4 Attachment 1, Volume 14, Rev. 1, Page 106 of 188
Attachment 1, Volume 14, Rev. 1, Page 107 of 188 ITS 3.9.5 ITS 0
'TnFUELING OPERATION4S LOW WATER LE'IEL V!IAMUG CONIXT10OM FOR OPERATION 2
-0
-0 LCO 3.9.5 3.9.8.2 Two Independent ResiduaT Heat ReMoval PR c PERABLE.IM Z --- F Add proposed LCO Note 21 AMPLICA8 T1`: MIODE t when the water level above the top of the reactor pressure vessel flange Is less than 23 feet..
ACTICN: ACTION A
- a. With less than the required RHR loops OPERABLE, trcediataly initiate corrective action to return the required MR loops to OPERAELE StatUS S. 0n A3s Sble. AddproposedRequiredAdion
- b. T' - rovislons of Speciflcation 3.0.3 n ot licable. 4 SMRVILLANfCE REOUIRIDIENS 4.9.8.2 uendrTh Resadul loops shall Ittv etsmned OPERABLE I per 5peetleattsen 4R .0.. a5 L.4G
- l4 Add proposed SR:3952adNt M.3e
!,The norTl-r emergency P"owe!.soic may be .1npero].elor each MR loap.[
*t -4
¢;. C. COOK - UHIT 2 3/4 9-Ba Amendrqnt No. 59 Page 4 of 4 Attachment 1, Volume 14, Rev. 1, Page 107 of 188
Attachment 1, Volume 14, Rev. 1, Page 108 of 188 DISCUSSION OF CHANGES ITS 3.9.5, RESIDUAL HEAT REMOVAL (RHR) AND COOLANT CIRCULATION - LOW WATER LEVEL ADMINISTRATIVE CHANGES A.1 In the conversion of the CNP Current Technical Specifications (CTS) to the plant specific Improved Technical Specifications (ITS), certain changes (wording preferences, editorial changes, reformatting, revised numbering, etc.) are made to obtain consistency with NUREG-1431, Rev. 2, "Standard Technical Specifications-Westinghouse Plants" (ISTS). These changes are designated as administrative changes and are acceptable because they do not result in technical changes to the CTS. A.2 CTS 3.9.8.1 requires at least one residual heat removal loop to be in operation in MODE 6. ITS 3.9.5 requires two RHR loops to be OPERABLE and one RHR loop to be in operation in MODE 6 with the water level less than 23 feet above the top of the reactor vessel flange. However, ITS 3.9.4 covers the Applicability of MODE 6 with water level greater than or equal to 23 feet above the top of the reactor vessel flange. This changes the CTS by splitting the requirements associated with CTS 3.9.8.1 into two Applicabilities, one for MODE 6 with water level < 23 feet above the top of the reactor vessel flange, and one for MODE 6 with water level > 23 feet above the reactor vessel flange. The purpose of CTS 3.9.8.1 is to ensure that adequate decay heat removal capability is in operation and that the coolant is circulated in MODE 6. This change is acceptable because the requirements continue to ensure that the process variables are maintained in the MODES and other specified conditions assumed in the safety analyses and licensing basis. MODE 6 RHR and coolant circulation requirements are governed by ITS 3.9.4, "Residual Heat Removal (RHR) and Coolant Circulation - High Water Level," and ITS 3.9.5, "Residual Heat Removal (RHR) and Coolant Circulation - Low Water Level." The combination of ITS 3.9.4 and ITS 3.9.5 ensures that the appropriate RHR loops are available in MODE 6 regardless of the water level. This change is designated as administrative because it does not result in technical changes to the CTS. A.3 CTS 3.9.8.1 Action a states, in part, that with less than one RHR loop in operation, suspend all operations involving an increase in the reactor decay heat load of the Reactor Coolant System. ITS 3.9.5 does not include this requirement. This changes the CTS by eliminating the requirement to suspend operations involving an increase in reactor decay heat load. This change is acceptable because the requirements have not changed. The reactor decay heat load is generated only by irradiated fuel. The only method of increasing the decay head load of a reactor in MODE 6 is to load additional irradiated fuel assemblies into the core. However, ITS LCO 3.9.6 prohibits loading of fuel assemblies into the reactor when the water level is less than 23 feet over the top of the reactor vessel flange. Therefore, when LCO 3.9.5 is applicable there is no method available to increase the reactor decay heat load, and the requirement can be deleted with no effect on plant operations. This change is designated as administrative because it does not result in technical changes to the CTS. CNP Units 1 and 2 Page 1 of 5 Attachment 1, Volume 14, Rev. 1, Page 108 of 188
Attachment 1, Volume 14, Rev. 1, Page 109 of 188 DISCUSSION OF CHANGES ITS 3.9.5, RESIDUAL HEAT REMOVAL (RHR) AND COOLANT CIRCULATION - LOW WATER LEVEL A.4 CTS 3.9.8.1 Action c and CTS 3.9.8.2 Action c state, "The provisions of Specification 3.0.3 are not applicable." ITS 3.9.5 does not include this statement. This changes CTS by deleting the Specification 3.0.3 exception. This change is acceptable because the technical requirements have not changed. ITS LCO 3.0.3 is not applicable in MODE 6. Therefore, the CTS LCO 3.0.3 exception is not needed. This change is designated as administrative because it does not result in a technical change to the CTS. A.5 CTS LCO 3.9.8.2 is modified by footnote *, which states that the normal or emergency power source may be inoperable for each RHR loop. ITS 3.9.5 does not include this statement. This changes the CTS by deleting an allowance already provided in a different portion of the ITS. This change is acceptable because the ITS definition of OPERABLE contains the necessary requirements for a component to perform its safety function. The ITS definition of OPERABLE states that a component is OPERABLE if either the normal or emergency power source is OPERABLE. This change is designated as administrative because it does not result in technical changes to the CTS. A.6 CTS 3.9.8.2 Action a states that with less than the required RHR loops OPERABLE, immediately initiate corrective action to return the required RHR loops to OPERABLE status as soon as possible. ITS 3.9.5 ACTION A includes the same requirement, but also includes an allowance (Required Action A.2) to immediately initiate action to establish > 23 feet of water above the top of reactor vessel flange. This changes the CTS by providing the option to exit the Applicability of the LCO. This change is acceptable because the requirements have not changed. Exiting the Applicability of LCO is always an option to exit an ACTION. Therefore, stating this option explicitly does not change the requirements of the Specification. This change is designated as administrative because it does not result in technical changes to the CTS. MORE RESTRICTIVE CHANGES M.1 The CTS 3.9.8.1 Actions do not include an action to immediately initiate action to restore one RHR loop to operation in the event the RHR loop requirements are not met. ITS 3.9.5 Required Action B.2 requires that action be immediately initiated to restore one RHR loop to operation. This changes the CTS by requiring that action be taken immediately to restore one RHR loop to operation. The purpose of CTS 3.9.8.1 is to ensure that adequate decay heat removal and coolant circulation are available in MODE 6. Although decay heat is removed from the Reactor Coolant System via natural circulation to the bulk of water contained in the refueling canal, this method of heat transfer can continue for only a discrete amount of time before boiling would occur. This change is acceptable because it requires that action be initiated to restore one RHR loop to operation in order to restore forced coolant flow and heat removal. This change CNP Units 1 and 2 Page 2 of 5 Attachment 1, Volume 14, Rev. 1, Page 109 of 188
Attachment 1, Volume 14, Rev. 1, Page 110 of 188 DISCUSSION OF CHANGES ITS 3.9.5, RESIDUAL HEAT REMOVAL (RHR) AND COOLANT CIRCULATION - LOW WATER LEVEL is designated as more restrictive because additional actions will be required in the ITS than are required in the CTS. M.2 CTS 4.9.8.1 requires that a residual heat removal loop shall be determined to be in operation and circulating reactor coolant at a flow rate of greater than or equal to 2000 gpm at least once per 24 hours. ITS SR 3.9.5.1 requires the same verification every 12 hours. This changes the CTS by requiring that RHR loop operation and reactor coolant flow rate be verified every 12 hours instead of every 24 hours. The purpose of CTS 4.9.8.1 is to ensure that adequate decay heat removal and coolant circulation are available in MODE 6. This change is acceptable since it results in an increased Frequency of performance. The 12 hour Frequency is consistent with similar CTS Surveillances in MODES 4 and 5, and with similar SRs in the ITS. This change is designated as more restrictive because the Surveillance will be performed at an increased Frequency in the ITS. M.3 CTS 3.9.8.2 requires two independent RHR loops to be OPERABLE and CTS 3.9.8.1 requires at least one RHR loop to be in operation. ITS SR 3.9.5.2 requires verification every seven days of correct breaker alignment and that indicated power is available to the required RHR pump not in operation. A Note states that the Surveillance Requirement is not required to be performed until 24 hours after a required RHR pump is not in operation. This changes the CTS by adding a Surveillance Requirement. The purpose of ITS 3.9.5 is to require one RHR loop to be in operation and one RHR loop to be held in readiness should it be needed. This change is acceptable because it verifies that the RHR loop that is in standby will be ready should it be needed. This change is designated as more restrictive because it adds a new Surveillance Requirement to the CTS. RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES None LESS RESTRICTIVE CHANGES L.1 (Category 4 -Relaxation of Required Action) CTS 3.9.8.1 Action a states, in part, that with less than one RHR loop in operation, suspend all operations involving a reduction in boron concentration of the Reactor Coolant System. This CTS Action is modified by a footnote which states that addition of water from the RWST does not constitute a dilution activity provided the boron concentration in the RWST is greater than the minimum required by Specification 3.1 .2.7.b.2 (i.e., CNP Units 1 and 2 Page 3 of 5 Attachment 1, Volume 14, Rev. 1, Page 110 of 188
Attachment 1, Volume 14, Rev. 1, Page 111 of 188 DISCUSSION OF CHANGES ITS 3.9.5, RESIDUAL HEAT REMOVAL (RHR) AND COOLANT CIRCULATION - LOW WATER LEVEL 2400 ppm). ITS 3.9.5 Required Action B.1 states that with no RHR loop in operation, suspend operations that would cause introduction into the RCS, coolant with boron concentration less than required to meet the boron concentration of LCO 3.9.1, "Boron Concentration." ITS 3.9.1 requires boron concentration to be within limit. This changes the CTS by allowing coolant with boron concentration less than the RCS boron concentration, but greater than the boron concentration limit in ITS LCO 3.9.1, to be added to the RCS from sources other than the RWST when the RHR loops are not in operation. The purpose of CTS 3.9.8.1 Action a is to ensure that the required SHUTDOWN MARGIN is maintained during periods when the RHR requirements are not met. This change is acceptable because the Required Actions are used to establish remedial measures that must be taken in response to the degraded conditions in order to minimize risk associated with continued operation while providing time to repair inoperable features. The Required Actions are consistent with safe operation under the specified Condition, considering the OPERABILITY status of the redundant systems of required features, the capacity and capability of remaining features, a reasonable time for repairs or replacement of required features, and the low probability of an accident occurring during the repair period. The Required Actions ensure that the RCS boron concentration is maintained within the limits of ITS LCO 3.9.1, which is sufficient to ensure that adequate SHUTDOWN MARGIN is maintained. This change is designated as less restrictive because less stringent Required Actions are being applied in the ITS than were applied in the CTS. L.2 (Category 4 - Relaxation of RequiredAction) CTS 3.9.8.1 Action a states, in part, that with less than one RHR loop in operation, close all containment penetrations providing direct access from the containment atmosphere to the outside atmosphere within 4 hours. ITS 3.9.5 Required Actions B.3, B.4, and B.5 state that with no RHR loop in operation, within 4 hours close and secure the equipment hatch with at least four bolts, close one door in each air lock, and verify each penetration providing direct access from the containment atmosphere to the outside atmosphere is either closed with a manual or automatic isolation valve, blind flange, or equivalent, or is capable of being closed by an OPERABLE Containment Purge Supply and Exhaust System. This changes the CTS Actions by allowing penetrations capable of being closed by an OPERABLE Containment Purge Supply and Exhaust System to remain open when no RHR loop is in operation. The purpose of CTS 3.9.8.1 Action a is to ensure that radioactive material does not escape the containment should the RHR requirements continue to not be met and boiling occurs in the core. Therefore, containment penetrations are closed to seal the containment. This change is acceptable because the Required Actions are used to establish remedial measures that must be taken in response to the degraded conditions in order to minimize risk associated with continued operation while providing time to repair inoperable features. The Required Actions are consistent with safe operation under the specified Condition, considering the OPERABILITY status of the redundant systems of required features, the capacity and capability of remaining features, a reasonable time for repairs or replacement of required features, and the low probability of a DBA CNP Units 1 and 2 Page 4 of 5 Attachment 1, Volume 14, Rev. 1, Page 111 of 188
Attachment 1, Volume 14, Rev. 1, Page 112 of 188 DISCUSSION OF CHANGES ITS 3.9.5, RESIDUAL HEAT REMOVAL (RHR) AND COOLANT CIRCULATION - LOW WATER LEVEL occurring during the repair period. The Required Actions are consistent with the actions taken for containment closure in CTS 3.9.4 and ITS 3.9.3. Penetrations which can be closed by an OPERABLE Containment Purge Supply and Exhaust System do not need to be closed if RHR is inoperable, since the presence of radioactivity in the containment will cause the valves to close automatically, thus performing the isolation function. This change is designated as less restrictive because less stringent Required Actions are being applied in the ITS than were applied in the CTS. L.3 (Category 1 - Relaxation of LCO Requirements) ITS 3.9.5 is modified by two LCO Notes. Note 1 allows all RHR pumps to be removed from operation for
< 15 minutes when switching from one loop to another, provided several conditions are met. Note 2 allows one required RHR loop to be inoperable for up to 2 hours for Surveillance testing, provided that the other loop is OPERABLE and in operation. Neither CTS 3.9.8.1 nor CTS 3.9.8.2 contain these allowances.
This changes the CTS by allowing the LCO to not be met under certain situations. The purpose of CTS 3.9.8.1 and CTS 3.9.8.2 is to ensure sufficient decay heat removal is available in the specified MODES and conditions. This change is acceptable because the LCO requirements continue to ensure that the structures, systems, and components are maintained consistent with the safety analyses and licensing basis. The ITS Notes allow normal operational evolutions, such as pump swapping and surveillance testing, to be performed while in the Applicability of the Specification. These evolutions are necessary to demonstrate RHR OPERABILITY. This change is designated as less restrictive because less stringent LCO requirements are being applied in the ITS than were applied in the CTS. L.4 (Category 5 - Deletion of Surveillance Requirement) CTS 4.9.8.2 requires verification that each RHR loop is OPERABLE per Specification 4.0.5. ITS 3.9.5 does not contain this Surveillance. This changes the CTS by deleting this specific Surveillance. The purpose of CTS Specification 4.0.5 is to require inservice testing in accordance with 10 CFR 50.55a. The purpose of inservice testing of RHR is to detect gross degradation caused by impeller structural damage or other hydraulic component problems. This change is acceptable because the deleted Surveillance Requirement is not necessary to verify that the equipment used to meet the LCO can perform its required functions. Thus, appropriate equipment continues to be tested in a manner and at a Frequency necessary to give confidence that the equipment can perform its assumed function. This Technical Specification will no longer tie RHR loop OPERABILITY to the Inservice Testing Program. This change is acceptable because it is not necessary to perform inservice testing of an RHR loop to determine if it is OPERABLE, as the system is routinely operated and the RHR loops are instrumented so that degradation can be observed. Significant degradation of the RHR System would be indicated by the RHR System flow and temperature instrumentation in the Control Room. This change is designated as less restrictive because Surveillances which are required in the CTS will not be required in the ITS. CNP Units I and 2 Page 5 of 5 Attachment 1, Volume 14, Rev. 1, Page 112 of 188
Attachment 1, Volume 14, Rev. 1, Page 113 of 188 Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs) Attachment 1, Volume 14, Rev. 1, Page 113 of 188
Attachment 1, Volume 14, Rev. 1, Page 114 of 188 CrS RHR and Coolant Circulation - Low Water Level 3.9. 3.9 REFUELING OPERATIONS 3.9. Residual Heat Removal (RHR) and Coolant Circulation - Low Water Level LCO3-9-6p Two RHR loops shall be OPERABLE, and one RHR loop shall be in _~- _- operation. (D-- ')
&'<Tr , q3 8-Wifan
) x. - NOTES -
All RHR pumps may be e~efor s 15 minutes when V switching from one(qttoanother provided: I a. The core outset temperature is maintained 0Wai 1 (i I below saturation temperature 4no Ic. b. No operations are permitted that would cause a reduction of the I1. k~-3 Reactor Coolant Systenboron conntratiorqd <C) go
- c. No draining operations to fu er reduce RCS water volume are permitted.
- 2. One required RHR loop may be Inoperable for up to 2 hours for surveillance testing, provided that the other RHR loop is OPERABLE and In operation.
i
.I APPLICABILITY: MODE 6 with the water level < 23 fl above the top of reactor vessel I flange.
4 1 i I ACTIONS 4-i CONDITION REQUIRED ACTION COMPLETION TIME A. Less than the required A.1 Initiate action to restore Immediately number of RHR loops required RHR loops to OPERABLE. OPERABLE status. OR
.,.4 A.2 Initiate action to establish Immediately 2 23 ft of water above the top of reactor vessel flange.
WOG STS 3.9.6 -1 Rev. 2, 04/30/01 Attachment 1, Volume 14, Rev. 1, Page 114 of 188
- Attachment 1, Volume 14,.Rev. 1, Page 1.15 of 188 RHR and Coolant Circulation -Low Water Level 3.91g?<o ACTIONS (continued)
.1. i CONDITION REQUIRED ACTION COMPLETION TIME
- 1 B. No RHR loop In B.1 Suspend operations that Immediatel operation. would cause Introductia 8 A A{ Into the RCS Dwith I boron concentration less than required to meet the boron concentration of LCID 3.9.,.
. iI A, ('VCcn Conetwt,# j<n B.2 Initiate action to restore Immediately one RHR loop to operation.
AND B.3 Close equipment hatch 4 hours and secure with.&oup bolts. AND B.4 Close one door In each air 4hours lock. AMD B.50 each penetration 4 hours p ding direct access from the containmert atmosphere to the outside atmospherePth a manual or automatic Isolation valve, blind flange, or equivalen
'4 30
- 1
'I WOG STS 3.9.6 - 2 Rev. 2. 04/30/01 Attachment 1, Volume 14, Rev. 1, Page 115 of 188
Attachment 1, Volume 14, Rev. 1, Page 116 of 188 RHR and Coolant Circulation - Low Water Level 3.94WO) 0 e3.1,8,J 4uii a. 4c1 ',IFn 1 SURVEILLANCE REQUIREMENTS A SURVEILLANCE FREQUENCY SR 3.911 Veri one RHR loop is In operation and circulating 12 hours era coolant at a flowgnen aPo rS.. II SR 3.9.f!2 { Verify correct breaker alignment and indicated power 7 days DX,0f-3 iI I available to the required RHR pump that is not in operation. 4
.I I
WOG STS 3.9.6- 3 Rev. 2. 04130101 Attachment 1, Volume 14, Rev. 1, Page 116 of 188
Attachment 1, Volume 14, Rev. 1, Page 117 of 188 3.9.5 0 INSERT 1
-NOTE-Not required to be performed until 24 hours after a required RHR pump is not in operation.
Insert Page 3.9.6-3 Attachment 1, Volume 14, Rev. 1, Page 117 of 188
Attachment 1, Volume 14, Rev. 1, Page 118 of 188 JUSTIFICATION FOR DEVIATIONS ITS 3.9.5, RESIDUAL HEAT REMOVAL (RHR) AND COOLANT CIRCULATION - LOW WATER LEVEL
- 1. CNP has analyzed a boron dilution event in MODE 6. Therefore, ISTS 3.9.2 is not included in the ITS and ISTS 3.9.6 is renumbered as ITS 3.9.5.
- 2. These punctuation corrections have been made consistent with the Writer's Guide for the Improved Standard Technical Specifications, NEI 01-03, Section 5.1.3.
- 3. The brackets are removed and the proper plant specific information/value is provided.
- 4. TSTF-265 was previously approved and incorporated in NUREG-1431, Rev. 2, in similar SRs (e.g., ISTS SRs 3.4.5.3, 3.4.6.3, 3.4.7.3, and 3.4.8.2). Consistent with TSTF-265, a Note is added to ISTS SR 3.9.6.2 that permits the performance of the SR to verify correct breaker alignment and power availability to be delayed until 24 hours after a required pump is not in operation. This provision is required because when pumps are swapped under the current requirements, the Surveillance is immediately not met on the pump taken out of operation. This change avoids entering an Action for a routine operational occurrence. The change is acceptable because adequate assurance exists that the pump is aligned to the correct breaker with power available because, prior to being removed from operation, the applicable pump had been in operation. Allowing 24 hours to perform the breaker alignment verification is acceptable because the pump was in operation, which demonstrated OPERABILITY, and because 24 hours is currently allowed by invoking SR 3.0.3.
This is a new Surveillance Requirement not required in CTS 3.9.8.2.
- 5. Editorial change made to be consistent with the LCO statement.
- 6. Editorial change made to be consistent with the format of the ITS.
- 7. ISTS 3.9.6 Required Actions B.5.1 and B.5.2 are connected by an "OR" logical connector, such that either one can be performed to meet the requirements of the ACTION. However, the two Required Actions are applicable to all the penetrations; either Required Action B.5.1 or Required Action B.5.2 must be performed for all the penetrations. Thus, this will not allow one penetration to be isolated by use of a manual valve and another penetration to be capable of being closed by an OPERABLE Containment Purge Supply and Exhaust System. This is not the intent of the requirement. The requirement is based on ISTS LCO 3.9.4 (ITS LCO 3.9.3),
which requires each penetration to be either: a) closed by a manual or automatic isolation valve, blind flange, or equivalent; or b) capable of being closed by an OPERABLE Containment Purge Supply and Exhaust System. For consistency with the actual LCO requirement, ISTS 3.9.6 Required Actions B.5.1 and B.5.2 have been combined into a single Required Action in ITS 3.9.5 Required Action B.5.
- 8. Changes have been made to be consistent with changes made in another Specification and be consistent with plant specific nomenclature.
- 9. The limit has been changed to be consistent with the same limit provided in Notes to ISTS 3.4.6 and ISTS 3.4.7.
CNP Units I and 2 Page 1 of 1 Attachment 1, Volume 14, Rev. 1, Page 118 of 188
Attachment 1, Volume 14, Rev. 1, Page 119 of 188 Improved Standard Technical Specifications (ISTS) Bases Markup and Justification for Deviations (JFDs) Attachment 1, Volume 14, Rev. 1, Page 119 of 188
Attachment 1, Volume 14, Rev. 1, Page 120 of 188 RHR and Coolant Circulation -Low Water LevelA'.....f '1 B 3.9 REFUELING OPERATIONS B3.9$rRes'idual H~eat Removal (RHR) and Coolant Circulation - Low Water Level BASES, BACKGROUND The purpose of the RHR System in MODE 6 is to remove decay heat and sensible heat from the Reactor Coolant System (RCS ajs rq C(G) to provide mixing of borated coolant, and to preivent ton 0I stratification (Ref. 1). Heat Is removed from the RCS by circulating reactor coolant through the RHR heat exchangers where the heat Is transferred to the Component Cooling Water System. The coolant is then returned to the RCS via the RCS cold leg(s). Operation of the RHR System for normal cooldown decay heat removal is manually accomplished from the control room. The heat removal rate is adjusted by controlling the flow of reactor coolant through the RHR heat
-exchanger(s) and the bypass line Mixing of the reactor coolant is ciulation of reactor coolant through the RHR System.
APPLICABLE If the reactor coolant temperature Is not maintained below 200F, boiling SAFETY of the reactor coolant could result. This could lead to a loss of coolant in ANALYSES the reactor vessel:. Additionally, boiling of the reactor coolant could lead to a reduction In boron concentration In the coolant due to the boron plating out on components near the areas of the boiling activity. The loss of reactor coolant and the reduction of boron concentration Inthe reactor coolant will eventually challenge the Integrity of the fuel cladding, which Is a fission product barrier. Two of the RHR System are required to be OPERABLE, and on peration. in order to prevent this challenge. The RHR System satisfies Criterion 4 of 10 CFR 50.36(c)(2)(ii). LCO In MODE 6, with the water level < 23 ft above the top of the reactor vessel flange, both RHR loops must be OPERABLE. Additionally, one loop of RHR must be in operation in order to provide:
- a. Removal of decay hea - (T -
i 4
- b. Mixing of borated coolant to minimize the possibility of criticaltn Z -
- 1 (g idictionof eacor colat /mperatu~re <
0 '
/
- This LCO Is modified by a Note that permits the RHR pumps to be@)
for s 15 minutes when switching from one to another.
- WOG STS B 3.9.6-1 (@ ) Rev. 2. 04130101 Attachment 1, Volume 14, Rev. 1, Page 120 of 188
Attachment 1, Volume 14, Rev. 1, Page 121 of 188 8 3.9.5 (i) INSERT I , as well as adjustments in Component Cooling Water System temperature and flow Insert Page B 3.9.6-1 Attachment 1, Volume 14, Rev. 1, Page 121 of 188
Attachment 1, Volume 14, Rev. 1, Page 122 of 188 RHR and Coolant Circulation - Low Water Leve t B39
.I BASES LCO (continued)
The circumstances for stopping both RHR pumps are to be limited to f__ situations when the outa e ime is shortlnd the core outlet temperature isrmalntainerJ 10 tirIos Ro o below saturation temperatures The Note rdF diiutoor draining operations when RHR forced flow is 3-.O stopped. * (2. This LCO Ismodified by a Note that allo o HR loop to be inoperable for a period of 2 hours provideh other loop Is OPERABLE and In operation. Prior to dedaring t Inope consideration sreacances tests to be performed on the inoperable loop during a time when these tests are safe and possible.i 'i An OPERABLE RHR loop consists of an RHR pump, a heat exchanger,
.valves, piping. Instruments and controls to ensure an OPERABLE flow pathCnd tonete e The flow path starts In C lone of the RCS hot legs and Is returned t)the RCS cold legs.
APPLICABILITY Two Ri-iR loops are required to be OPERABLE, and one RHR loop must be In operation In MODE 6, with the water level < 23 ft above the top of the reactor vessel flange, to provide decay heat removal. Requirements for the RH ISystem In other MODES are covered by LCOs In Section 3.XReacor Coolant Sy tem (RCS)`an, en oreT g *qyans iECCS]7HR loop requirements in MODEbwith
.4
- the water level i 23 ft are located In LCO 3.96KResidual Heat Removal (RHR) and Coolant Circulation - High Water Leve=L/M ACTIONS A.1 and A.2 If less than the required number of RHR loops are OPERABLE, action shall be irrimediately Initiated and continued until the RHR loop Is restored to OPERABLE status and to operation or until 2 23 ft of water level Is established above the reactor vessel flange. When the water level is 2 23 ft abo the reactor vessel flange, the Applicability changes
. ga o CO 3.9 , and only one RHR loop Is required to be OPERABLE and In operation. An Immediate Completion Time Is a)
Id necessary for an operator to Initiate corrective actions. WOG STS B 3.9.6-2 Rev. 2, 04130/01 Attachment 1, Volume 14, Rev. 1, Page 122 of 188
Attachment 1, Volume 14, Rev. 1, Page 123 of 188 RHR and Coolant Circulation - Low Water Leve e()- B 3.9.. BASES ACTIONS (continued) If no RHR loop Isin operation, there will be no forced circulation to provide mixing to establish uniform boron concentrations. Suspending positive reactivity additions that could result in failure to meet the minimum boron concentration Emit Isrequired 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 refueling boron concentration. This may result In an overall reduction InRCS boron concentration, but provides acceptable margin to maintaining subcrtical operation. i iII If no RHR loop Is In operation, actions shall be initiated immediately, and I continued, to restore one RHR loop to operation. Since the unit IsIn I Conditions A and B concurrently, the restoration of two OPERABLE RHR II loops and one operating RHR loop should be accomplished I expeditiously. 0D I.1I If no RHR IsIn operation, the following actions must be taken: I
- a. The equipment hatch must be closed and secured withour0folto%
- b. One door Ineach air lock must be closeegand ©
- c. Each penetration providing direct access from the containment 0T atmosphere to the outside atmosphere must be either closed by a manual or automatic Isolation valve, blind flange, or equivalent, or verified to be capable of being closed by an OPERABLE Containment Purgedxhaust niMR System. E0 With RHR loop requirements not met, the potential exists for the coolant to boil and release radioactive gas to the containment atmosphere.
Performing the actions stated above ensures that all containment penetrations are either closed or can be dosed so that the dose limits are not exceeded. I The Completion Time of 4 hours allows fixing of most RHR problems and is reasonable, based on the low probability of the coolant boiling in that time. WOG STS B 3.9.6-3 Rev. 2, 04130/01 Attachment 1, Volume 14, Rev. 1, Page 123 of 188
Attachment 1, Volume 14, Rev. 1, Page 124 of 188 RHR and Coolant Circulation - Low Water Level r.45- f B3.4 BASES SURVEILLANCE REQUIREMENTS SRa &3.9) 0D This Surveillance demonstrates that one RHR loop is in operation and circulating reactor coolant. The flow rate Isdetermined by the flow rate necessary to provide sufficient decay heat removal capability and to prevent thermal and boron stratification in the core. In addition, during operation of the RHR loop with the water lovol In the vicinity of the reactor vessel nozzles, the RHR pump suction requirements must be met. The Frequency of 12 hours Is sufficient, considering the flow. temperature, pump control, and alarm Indications available to the operaor for monitoring the RHR System In the control room. 5R 3,'. S2 -Verification that the required pump IsOPERABLE ensures that an additionala RHR pump can be placed in operation, If needed, to maintain decay heat removal and reactor coolant circulation. Verification Is performed by verifying proper breaker alignment and power available to the required pump. The Frequency of 7 days is considered reasonable Inview of other administrative controls available and has been shown to e4acceptable by operating experience. REFERENCES 1. (FSAR, Sectionl<a3) I i WOG STS B 3.9.6-4 Rev. 2, 04130/01 Attachment 1, Volume 14, Rev. 1, Page 124 of 188
Attachment 1, Volume 14, Rev. 1, Page 125 of 188 B 3.9.5 0 INSERT 2 This SR is modified by a Note that states the SR is not required to be performed until 24 hours after a required pump is not in operation. Insert Page B 3.9.6-4 Attachment 1, Volume 14, Rev. 1, Page 125 of 188
Attachment 1, Volume 14, Rev. 1, Page 126 of 188 JUSTIFICATION FOR DEVIATIONS ITS 3.9.5 BASES, RESIDUAL HEAT REMOVAL (RHR) AND COOLANT CIRCULATION
- LOW WATER LEVEL
- 1. Changes are made to reflect those changes made to the ISTS. Subsequent requirements are renumbered or revised, where applicable, to reflect the changes.
- 2. CNP Units 1 and 2 were designed and under construction prior to the promulgation of 10 CFR 50, Appendix A. CNP Units I and 2 were designed and constructed to meet the intent of the proposed General Design Criteria, published in 1967.
However, the CNP UFSAR contains discussions of the Plant Specific Design Criteria (PSDCs) used in the design of CNP Units I and 2. Bases references to the 10 CFR 50, Appendix A, criteria have been replaced with references to the appropriate section of the UFSAR.
- 3. The current wording implies specific restrictions not contained in LCO Note 2.
Therefore, the words have been modified to provide guidance on what should be considered in determining whether or not to use the Note allowance.
- 4. The brackets have been removed and the proper plant specific information/value has been provided.
- 5. Changes are made (additions, deletions, and/or changes) to the ISTS Bases which reflect the plant specific nomenclature, number, reference, system description, analysis, or licensing basis description.
- 6. The wording has been modified, as Section 3.5 does not provide requirements for the RHR Shutdown Cooling function.
- 7. These punctuation corrections have been made consistent with the Writer's Guide for the Improved Standard Technical Specifications, NEI 01-03, Section 5.1.3.
- 8. Typographical/grammatical error corrected.
- 9. Changes are made to be consistent with the ISTS.
CNP Units 1 and 2 Page 1 of 1 Attachment 1, Volume 14, Rev. 1, Page 126 of 188
Attachment 1, Volume 14, Rev. 1, Page 127 of 188 Specific No Significant Hazards Considerations (NSHCs) Attachment 1, Volume 14, Rev. 1, Page 127 of 188
Attachment 1, Volume 14, Rev. 1, Page 128 of 188 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.9.5, RESIDUAL HEAT REMOVAL (RHR) AND COOLANT CIRCULATION - LOW WATER LEVEL There are no specific NSHC discussions for this Specification. CNP Units I and 2 Page 1 of 1 Attachment 1, Volume 14, Rev. 1, Page 128 of 188
, Volume 14, Rev. 1, Page 129 of 188 ATTACHMENT 6 ITS 3.9.6, Refueling Cavity Water Level , Volume 14, Rev. 1, Page 129 of 188 , Volume 14, Rev. 1, Page 130 of 188 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) , Volume 14, Rev. 1, Page 130 of 188
Attachment 1, Volume 14, Rev. 1, Page 131 of 188 ITS 3.9.6 ITS RUMUMK OVII.1ATICIPS 3/M.9.10 UAMI LZVZ - YZACMO VISSL rTwivTwtvt c0Mar106 FMR OIU1ATO LCO 3.9.6 3.9.10 At leant 23 feet of water shall be salnealned over the cop of the reactor pressure vessel flqai. ; ACTION A SymjmZ.Acg 3ZOUZR= -T SR 3.9.6.1 4.9.10 The vatSrl" salb dte ineo be leasnt Its sinLam required 4. tb vlthIum 2 b " m f to the start of n at es oce :L 24 hoursaChaeaeter tri&mant of fuel assomieg ottoatol ro -
- 0. C. COOK - MIT 1 3/4 9.11 Amendment No. 78 Page 1 of 2 Attachment 1, Volume 14, Rev. 1, Page 131 of 188
Attachment 1, Volume 14, Rev. 1, Page 132 of 188 ITS 3.9.6 ITS REFUELIfNG OPERATIONS 3/4.9.10 W'ATER LEVEL - REACTOR VESSEL LIMITING CONOITION FOR OPERATION LCO 3.9.6 3.9.10 At least 23 feet of water shall be maintained over the tap of thc reactor pressure vessel flange. r APPLICABILITY: Ouring movement of fuel assemblies Lor c 1 rodsWWJiR ACTION: ACTION A With the requirements of the abo speciftcation not satstied susend al a ooerati0ns nvolvin movement a fuel assemblies r q w n th pressu vessel.1 Phe provys5ns ofaploctfe E 3.0.3 are no ip=ca SURVEILLANCE REOUIREMENTS SR 3.9.6.1 4.9.10 The water level shall be deteroined to be at least its o*njL 1.3 required depth wit*in 2 hourI.Era o tne start or an at least once pr 24 hoursithereafter cur!ns...meht o fuel asseatielor colet ros a-v.In ,f feassezilo om D..C. CZOK - IJN'T 2 3;.' 9;1C Amendment No. 59 Page 2 of 2 Attachment 1, Volume 14, Rev. 1, Page 132 of 188
Attachment 1, Volume 14, Rev. 1, Page 133 of 188 DISCUSSION OF CHANGES ITS 3.9.6, REFUELING CAVITY WATER LEVEL ADMINISTRATIVE CHANGES A.1 Inthe conversion of the CNP Current Technical Specifications (CTS) to the plant specific Improved Technical Specifications (ITS), certain changes (wording preferences, editorial changes, reformatting, revised numbering, etc.) are made to obtain consistency with NUREG-1431, Rev. 2, "Standard Technical Specifications-Westinghouse Plants" (ISTS). These changes are designated as administrative changes and are acceptable because they do not result in technical changes to the CTS. A.2 CTS 3.9.10 is applicable in MODE 6 during movement of fuel assemblies or control rods within the reactor pressure vessel. ITS 3.9.6 is applicable during movement of irradiated fuel assemblies within containment. This changes the CTS by eliminating the "MODE 6" portion of the Applicability. The change to "irradiated fuel assemblies" from "fuel assemblies" is discussed in DOC L.1. The change from within "the reactor pressure vessel" to within "containment" is discussed in DOC M.1. The change eliminating control rods is discussed in DOC L.2. This change is acceptable because the technical requirements have not changed. Fuel movement in the containment only occurs in MODE 6. Therefore, specifying MODE 6 during movement of fuel is unnecessary. This change is designated as administrative because the technical requirements of the CTS have not changed. A.3 The CTS 3.9.10 Action states 'The provisions of Specification 3.0.3 are not applicable." ITS 3.9.6 does not include this statement. This changes the CTS by deleting the Specification 3.0.3 exception. This change is acceptable because the technical requirements have not changed. ITS LCO 3.0.3 is not applicable in MODE 6. Therefore, the CTS LCO 3.0.3 exception is not needed. This change is designated as administrative because it does not result in a technical change to the CTS. MORE RESTRICTIVE CHANGES M.1 CTS 3.9.10 is applicable during movement of fuel assemblies or control rods within the "reactor pressure vessel" while in MODE 6. The CTS 3.9.10 Action states that with the reactor vessel water level not within limit, suspend movement of fuel assemblies or control rods within the "pressure vessel." The ITS 3.9.6 Applicability is'during movement of irradiated fuel assemblies within "containment." ITS 3.9.6 ACTION A states that with the refueling cavity water level not within limit, suspend movement of irradiated fuel assemblies within "containment." This changes the CTS by expanding the suspension of movement of fuel assemblies from within the "reactor pressure vessel" to within the "containment." The change to "irradiated fuel assemblies" from "fuel assemblies" is discussed in DOC L.1. The change eliminating MODE 6 is discussed in DOC A.2. The change eliminating control rods is discussed in DOC L.2. CNP Units I and 2 Page 1 of 3 Attachment 1, Volume 14, Rev. 1, Page 133 of 188
Attachment 1, Volume 14, Rev. 1, Page 134 of 188 DISCUSSION OF CHANGES ITS 3.9.6, REFUELING CAVITY WATER LEVEL The purpose of CTS 3.9.10 is to ensure the refueling cavity water level is greater than or equal to that assumed in the fuel handling accident analysis. This change is acceptable because the fuel handling accident analysis assumes an irradiated fuel assembly is damaged within the containment, not only within the reactor vessel. In order to protect the initial assumptions of the fuel handling accident analysis, prohibition of irradiated fuel movement within the containment is required. This change is designated as more restrictive because it will prohibit operations that are not prohibited in the CTS. RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES None LESS RESTRICTIVE CHANGES L.I (Category 2 - Relaxation ofApplicability) CTS 3.9.10 states that at least 23 feet of water must be maintained over the reactor pressure vessel flange in MODE 6 during movement of fuel assemblies or control rods within the reactor pressure vessel. The CTS 3.9.10 Action requires suspension of movement of fuel assemblies or control rods within the pressure vessel if the water level requirement is not met. ITS 3.9.6 states the refueling cavity water level shall be maintained > 23 feet above the top of the reactor vessel flange during movement of irradiated fuel assemblies within containment. ITS 3.9.6 Required Action A.1 requires the suspension of movement of irradiated fuel assemblies within containment. This changes the CTS restricting the Applicability and ACTIONS from movement of any "fuel assemblies" within the reactor pressure vessel to movement of "irradiated fuel assemblies" within containment. The change eliminating MODE 6 is discussed in DOC A.2. The change from within "the reactor pressure vessel" to within "containment" is discussed in DOC M.1. The change eliminating control rods is discussed in DOC L.2. The purpose of CTS 3.9.10 is to ensure that the refueling cavity water level is greater than or equal to that assumed in the fuel handling accident analysis. This change is acceptable because the requirements continue to ensure that the process variables are maintained in the MODES and other specified conditions assumed in the safety analyses and licensing basis. The fuel handling accident analysis is based on damaging a single irradiated fuel assembly. An unirradiated fuel assembly does not contain the radioactive materials generated by fission and does not result in significant offsite doses if damaged. This change is designated as less restrictive because the LCO requirements are applicable in fewer operating conditions than in the CTS. L.2 (Category 2 - Relaxation ofApplicability) CTS 3.9.10 requires the refueling cavity water level to be maintained at least 23 feet over the top of the reactor CNP Units 1 and 2 Page 2 of 3 Attachment 1, Volume 14, Rev. 1, Page 134 of 188
Attachment 1, Volume 14, Rev. 1, Page 135 of 188 DISCUSSION OF CHANGES ITS 3.9.6, REFUELING CAVITY WATER LEVEL pressure vessel flange during movement of fuel assemblies or control rods within the reactor pressure vessel while in MODE 6. The CTS 3.9.10 Action requires suspension of all operations involving movement of the fuel assemblies or control rods within the pressure vessel in the event the LCO is not met. CTS 4.9.10 requires a determination of the refueling canal water level during the movement of fuel assemblies or control rods. ITS 3.9.6 requires the refueling cavity water level to be maintained greater than or equal to 23 feet above the top of the reactor vessel flange during movement of irradiated fuel assemblies within containment. This changes the CTS by deleting the requirement that the LCO, ACTIONS, and Surveillance is applicable during control rod movement. The change to "irradiated fuel assemblies" from "fuel assemblies" is discussed in DOC L.1. The change eliminating MODE 6 is discussed in DOC A.2. The change from within "the reactor pressure vessel" to within "containment" is discussed in DOC M.1. The purpose of CTS 3.9.10 is to ensure that the refueling cavity water level is greater than or equal to that assumed in the fuel handling accident analysis. This change is acceptable because the requirements continue to ensure that the process variables are maintained in the MODES and other specified conditions assumed in the safety analyses and licensing basis. The fuel handling accident is based on damaging a single irradiated fuel assembly. Movement of control rods is not assumed to result in a fuel handling accident. This change is designated as less restrictive because the LCO requirements are applicable in fewer operating conditions than in the CTS. L.3 (Category 7 - Relaxation Of Surveillance Frequency, Non-24 Month Type Change) CTS 4.9.10 requires the refueling cavity water level to be determined to be within limit "within 2 hours prior to the start of' and at least once per 24 hours thereafter during movement of fuel assemblies or control rods. ITS SR 3.9.6.1 requires verification that the refueling cavity water level is within limit every 24 hours. This changes the CTS by reducing the Frequency for verifying refueling cavity water level from 2 hours before entering the Applicability of the LCO to 24 hours before entering the Applicability of the LCO. The purpose of CTS 4.9.10 is to ensure that the refueling cavity water level is greater than or equal to that assumed in the fuel handling accident analysis. This change is acceptable because the new Surveillance Frequency has been evaluated to ensure that it provides an acceptable level of equipment reliability. The Frequency of 24 hours is sufficient during the movement of fuel assemblies, therefore it is sufficient before fuel assemblies are moved. ITS SR 3.0.1 requires the SR to be met during the MODES or other specified conditions in the Applicability. Therefore, the water level must be met when fuel assemblies are moved or fuel assembly movement must be suspended immediately (thereby exiting the Applicability of the Specification). Therefore, changing the Frequency from 2 hours before moving fuel assemblies to within 24 hours before moving fuel assemblies has no effect on plant safety. This change is designated as less restrictive because Surveillances will be performed less frequently under the ITS than under the CTS. CNP Units 1 and 2 Page 3 of 3 Attachment 1, Volume 14, Rev. 1, Page 135 of 188
Attachment 1, Volume 14, Rev. 1, Page 136 of 188 Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs) Attachment 1, Volume 14, Rev. 1, Page 136 of 188
Attachment 1, Volume 14, Rev. 1, Page 137 of 188 Cts Refueling Cavity Water Level 3.9.j._ 3.9 REFUELING OPERATIONS 3.9<Refueling Cavity Water Level
- 3.q.}W LCO 3.91 Refueling cavity water level shall be maintained reactor vessel flange.
2 23 ft above the top of 0D
- -I 1/
APPLICABILITY: During movement of irradiated fuel assemblies within containment. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Refueling cavity water A.1 Suspend movement of Immediately level not within limit.Irdae fuel assemblies within containment. SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY
'91/0 SR 3.9$.1
/
Verify refueling cavity water level Is top of reactor vessel flange. 2 23 ft above the 24 hours 0 _ I ... . WOG STS 3.9.7 - 1 Rev. 2, 04130/01 Attachment 1, Volume 14, Rev. 1, Page 137 of 188
Attachment 1, Volume 14, Rev. 1, Page 138 of 188 JUSTIFICATION FOR DEVIATIONS ITS 3.9.6, REFUELING CAVITY WATER LEVEL
- 1. CNP has analyzed a boron dilution event in MODE 6. Therefore, ISTS 3.9.2 is not included in the ITS and ISTS 3.9.7 is renumbered as ITS 3.9.6.
CNP Units I and 2 Page 1 of 1 Attachment 1, Volume 14, Rev. 1, Page 138 of 188
Attachment 1, Volume 14, Rev. 1, Page 139 of 188 Improved Standard Technical Specifications (ISTS) Bases Markup and Justification for Deviations (JFDs) Attachment 1, Volume 14, Rev. 1, Page 139 of 188
Attachment 1, Volume 14, Rev. 1, Page 140 of 188 Refueling Cavity Water LeveIW ) B 3.9r B 3.9 REFUELING OPERATIONS
*1 B 3.9.0 Refueling Cavity Water Level I
BASES jI
'I BACKGROUND The movement of Irradiated fuel assemblies within containment requires Ii a minimum water level of 23 ft above the top of the reactor vessel flange. During refueling, this maintains sufficient water level in the containment, refueling canal, fuel transfer canal, refueling cavity, and spent fuel pool. Sufficient water is necessary to retain iodine fission product activity in the water in the event of a fuel handling accident (Refs. I and 2). Sufficient Iodine activity would be retained to limit offsite doses from the accident to af 10 CFR 100 limitNa sb t4 nudeerl -Lr APPLICABLE During movement of Irradiated fuel assemblies, the water level in the SAFETY refueling canal and the refueling cavity is an Initial condition design ANALYSES parameter in the analysis of a fuel handling accident in containment, as postulated by Regulatory Guide 1.25 (Ref. 1). A minimum water level of 23 ft (Regulatory Position C.1.c of Ref. 1) allows a decontamination factor of 100 (Regulatory Position C.1.g of Ref. 1)to be used in the accident analysis for iodine. This relates to the assumption that 99% of the total Iodine released from the pellet to cladding gap of all the dropped fuel assembly rods Is retained by the refueling cavity water. The fuel pellet to cladding gap Is assumed to contain 10% of the total fuel rod iodine inventory (Ref. 1).
The fuel handling accident analysis inside containment Is described in Referec 2. With a minimum water level of 23 ft and a minimum decay time of ours prior to fuel handling, the analysis and test programs demonstrate that the Iodine release due to a postulated fuel handling accident is adequately captured by the water and offsite doses are maintained within allowable limits (Ref94>a 0 Refueling cavity water level satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii).
; LCO A minimum refueling cavity water level of 23 ft above the reactor vessel flange Isrequired to ensure that the radiological consequences of a -i postulated fuel handling accident inside containment are within acceptable limit asprod t dan eerence WOG STS B 3.9.7 - 1 Rev. 2, 04/30/01 Attachment 1, Volume 14, Rev. 1, Page 140 of 188
Attachment 1, Volume 14, Rev. 1, Page 141 of 188 Refueling Cavity Water Leve i B 3.9Vw 4 BASES A0 m APPLICABILITY LCO 3.91As applicable when moving irradiated fuel assemblies within containment. The LCO minimizes the possibility of a fuel handling accident In containment that Isbeyond the assunplo f the sf analysis. If Irradiated fuel assemblies are not en n co nt aiu there can be no significant radioactivity release as a result of a postulatet 0C fuel handling accident Requirements forJIeI handling accidents in the spent fuel pool are covered by ICO 3.7.t "Fuel Storage Pool Water i Level.'e g ACTIONS A1 With a water level of < 23 ft above the top of the reactor vessel flange, all operations Involvingomovement of Irradiated fuel assemblies within the containment shall be suspended immediately to ensure that a fuel handling accident cannot occur. The suspension of fuel movement shall not preclude completion of .1 movement of a component to a safe position. SURVEILLANCE Swl REQUIREMENTS Verification of a minimum water level of 23 ft above the top of the reactor vessel flange ensures that the design basis for the analysis of the postulated fuel handling accident during refueling operations is met. Water at the required level above the top of the reactor vessel flange limits the consequences of damaged fuel rods that are postulated to result from a fuel handling accident Inside containment (Ref. 2). The Frequency of 24 hours Is based on engineering judgment and Is considered adequate In view of the large volume of water and the normal procedural controls of valve positions, which make significant unplanned level changes unlikely. REFERENCES 1. Regulatory Guide 1.25 March 23 1972. 2.10CSAR. Se100.10.9
- F. -MS 96-0UW c .7.g .4
.4 (. 10 CFR 100.10. II I1 WOG STS 63.9.7-2 Rev. 2. 04/30101 Attachment 1, Volume 14, Rev. 1, Page 141 of 188
Attachment 1, Volume 14, Rev. 1, Page 142 of 188 Refueling Cavity Water Level B 3.90-Q BASES REFERENCES (conUnued) ( t5. M iD WCAP-78 I., Duhn, E., andLogPteJ. jotogical Consequence A x, December 1971. uel Handling ' I Ii I .1 WOG STS B3.9.7-3 Rev. 2. 04/30101 Attachment 1, Volume 14, Rev. 1, Page 142 of 188
Attachment 1, Volume 14, Rev. 1, Page 143 of 188 . JUSTIFICATION FOR DEVIATIONS ITS 3.9.6 BASES, REFUELING CAVITY WATER LEVEL
- 1. Changes are made to reflect those changes made to the ISTS. Subsequent requirements are renumbered or revised, where applicable, to reflect the changes.
- 2. Changes are made (additions, deletions, and/or changes) to the ISTS Bases which reflect the plant specific nomenclature, number, reference, system description, analysis, or licensing basis description.
- 3. The brackets have been removed and the proper plant specific information/value has been provided.
- 4. Typographical/grammatical error corrected.
- 5. Changes are made to be consistent with the ISTS.
CNP Units 1 and 2 Page 1 of 1 Attachment 1, Volume 14, Rev. 1, Page 143 of 188
Attachment 1, Volume 14, Rev. 1, Page 144 of 188 Specific No Significant Hazards Considerations (NSHCs) Attachment 1, Volume 14, Rev. 1, Page 144 of 188
Attachment 1, Volume 14, Rev. 1, Page 145 of 188 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.9.6, REFUELING CAVITY WATER LEVEL There are no specific NSHC discussions for this Specification. CNP Units 1 and 2 Page 1 of 1 Attachment 1, Volume 14, Rev. 1, Page 145 of 188
Attachment 1, Volume 14, Rev. 1, Page 146 of 188 ATTACHMENT 7 Relocated/Deleted Current Technical Specifications (CTS) Attachment 1, Volume 14, Rev. 1, Page 146 of 188
, Volume 14, Rev. 1, Page 147 of 188 CTS 314.9.3, Decay Time , Volume 14, Rev. 1, Page 147 of 188 , Volume 14, Rev. 1, Page 148 of 188 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) , Volume 14, Rev. 1, Page 148 of 188
Attachment 11 Volume 14, Rev. 1, Page 149 of 188 CTS 3/4.9.3 43.3 Ile "~fbed edoa subai tiulWzqubratkasdtendo S. b n~w~~f fe iemtriw COOK fC"-LEA f4tUN1T1 l 93 E4 1 , 20* I*T Page 1 of 2 Attachment 1, Volume 14, Rev. 1, Page 149 of 188
Attachment 1, Volume 14, Rev. 1, Page 150 of 188 CTS 3/4.9.3 3/4 .IfGClDOSFORtOLnX AND SURVANCK 3143 G OPNRATIONS nrCAYlTWl I. 3.93 Tiz sblflbebicalforatlent L bOOm.w.
- b. 1481an
- tSpeif 3.93.a - SftrSbcr 15 ftvugb Juno 15, O of hudinted SpCifct=n 3.93b - Pn 16 fh Septer 14, MOY odisted WMih &z suotical fctcin tIm q all veratiho knovin oflfiratedhjl 9CaaeL liz prwo:iono ofSpe 3.03 are eat applable.
4g31 he; shlli be f* ff etos e ueie d and te of COOK PLANTUJNrr2 Ppg 314 9.3 AMZqDMq3,243 Page 2 of 2 Attachment 1, Volume 14, Rev. 1, Page 150 of 188
Attachment 1, Volume 14, Rev. 1, Page 151 of 188 DISCUSSION OF CHANGES CTS 314.9.3, DECAY TIME ADMINISTRATIVE CHANGES None MORE RESTRICTIVE CHANGES None RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES LA.1 (Category 6 - Relocation of LCO or SR to the TRM) CTS LCO 3.9.3 requires the reactor to be subcritical for a required period of time (100 hours from September 15 through June 15 and 148 hours from June 16 through September 14) prior to movement of irradiated fuel in the reactor pressure vessel. ITS 3.9 does not include the requirements for decay time. This changes the CTS by moving the explicit decay time requirements from the Technical Specifications to the Technical Requirements Manual (TRM). The removal of these details from the Technical Specifications is acceptable because this type of information is not necessary to provide adequate protection of public health and safety. The purpose of CTS LCO 3.9.3 to ensure that sufficient time has elapsed to allow the radioactive decay of the short-lived fission products in the irradiated fuel consistent with the assumptions used in the fuel handling accident analysis. Additionally, two time limits are currently provided to account for decay heat load capacity of the spent fuel storage pool. Although CTS LCO 3.9.3 satisfies Criterion 2 of the Technical Specifications Selection Criteria in 10 CFR 50.36 (c)(2)(ii) (for the radioactive decay assumptions in the fuel handling accident), the requirements for decay time following subcriticality will always be met for a refueling outage because of the operations required prior to moving irradiated fuel in the reactor vessel (e.g., containment entry, removal of vessel head, removal of vessel internals, etc.). Also, this change is acceptable because the removed information will be adequately controlled in the TRM. The TRM is incorporated by reference into the UFSAR and any changes to the TRM are made under 10 CFR 50.59, which ensures changes are properly evaluated. This change is designated as a less restrictive removal of detail change because a requirement is being removed from the Technical Specifications. LESS RESTRICTIVE CHANGES None CNP Units 1 and 2 Page 1 of I Attachment 1, Volume 14, Rev. 1, Page 151 of 188
Attachment 1, Volume 14, Rev. 1, Page 152 of 188 Specific No Significant Hazards Considerations (NSHCs) Attachment 1, Volume 14, Rev. 1, Page 152 of 188
Attachment 1, Volume 14, Rev. 1, Page 153 of 188 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS CTS 314.9.3, DECAY TIME There are no specific NSHC discussions for this Specification. CNP Units 1 and 2 Page 1 of 1 Attachment 1, Volume 14, Rev. 1, Page 153 of 188
, Volume 14, Rev. 1, Page 154 of 188 CTS 314.9.5, Communications , Volume 14, Rev. 1, Page 154 of 188 , Volume 14, Rev. 1, Page 155 of 188 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) , Volume 14, Rev. 1, Page 155 of 188
Attachment 1, Volume 14, Rev. 1, Page 156 of 188 CTS 3/4.9.5 REFUELING OPE TONS COMMUNICATION LIMITING CO ITON FOR OPERATION 3.9.5 Direc coamunicatiOns shall be intained between the cc rol room and personne at the refuelIng static APPLICABILI During CORE ALTERATI S. ACTION: l* When direct comnunications between t e control room and person 1 at the refueling s tion cannot be maintain d1 suspend all CORE AL TIONS. The provis1 ns of Specification 3.0. are not applicable. SURVEILE REQUIREMENTS 4.9.5 Di oct communications betwe the control room and pe onnel at the refue ing station shall be dendstrated within one hour ior to the start of nd at least once per 12 ours during CORE ALTERATI NS. D.C. OOK- UNIT I J -' Page 1 of 2 Attachment 1, Volume 14, Rev. 1, Page 156 of 188
, Volume 14, Rev. 1, Page 157 of 188 CTS 3/4.9.5 REFUELIK OPERATIONS COMIUNI TIONS LINITI GICONDITION FOR OPERATION 3.9.5 rect communications shal be maintained between th control room and pei nnel at the refueling s tion.
APPLICA ILITY: During CORE AlT TIONS. ACTION:l When dl ect communicatlons bete n the control room and pe onnel at the refueli g station cannot be rain alned, suspend all CORE A ERATIONS. The provisi ns of Specification 3.0; are not applicable. SURVEI LANCE REqUIREMENTS 9.5 Direct comonicatlons be een the control room and ersonnel at the refuel ng station shall be dewo strated within one hour p or to the start of and at least once per 12 hou s during CORE -ALTERATIONS D. . COOK - UNIT 2 3/4 9-5 Page 2 of2 , Volume 14, Rev. 1, Page 157 of 188
Attachment 1, Volume 14, Rev. 1, Page 158 of 188 DISCUSSION OF CHANGES CTS 314.9.5, COMMUNICATIONS ADMINISTRATIVE CHANGES None MORE RESTRICTIVE CHANGES None RELOCATED SPECIFICATIONS R.1 CTS 3.9.5 states that direct communications shall be maintained between the control room and personnel at the refueling station during CORE ALTERATIONS. This ensures that refueling station personnel can be promptly informed of significant changes in the facility status or core reactivity conditions during CORE ALTERATIONS. The prompt notification of the control room of a fuel handling accident is not an assumption in the fuel handling accident analysis. While notification is necessary to ensure that the control room is isolated to meet the control room operator dose limits in General Design Criteria 19, the fuel handling accident analysis does not take credit for direct communications between the refueling station and the control room (30 minutes is assumed before control room operator actions are taken). This LCO does not meet the criteria for retention in the ITS; therefore, it will be retained in the Technical Requirements Manual (TRM). 10 CFR 50.36(c)(2)(ii) Criteria Evaluation:
- 1. Communications are not installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary. The Communications Specification does not satisfy criterion 1.
- 2. Communications are not a process variable, design feature, or operating restriction that is an initial condition of a DBA or Transient Analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. The Communications Specification does not satisfy criterion 2.
- 3. Communications are part of the primary success path and are assumed in the mitigation of a DBA which assumes the failure of a fission product barrier. However, communications are not a structure, system or component. The Communications Specification does not satisfy criterion 3.
- 4. Communications are not a structure, system, or component which operating experience or probabilistic risk assessment has shown to be significant to public health and safety. As discussed in Section 4.0, (Appendix A, page A-67) and Table 1 of WCAP-1 1618, communications was found to be a non-significant risk contributor to core damage frequency and offsite releases. I&M has reviewed this evaluation, CNP Units 1 and 2 Page 1 of 2 Attachment 1, Volume 14, Rev. 1, Page 158 of 188
Attachment 1, Volume 14, Rev. 1, Page 159 of 188 DISCUSSION OF CHANGES CTS 3/4.9.5, COMMUNICATIONS considers it applicable to CNP Units 1 and 2, and concurs with this assessment. The Communications Specification does not meet criterion 4. Since the 10 CFR 50.36(c)(2)(ii) criteria have not been met, the communications LCO and associated Surveillances may be relocated out of the Technical Specifications. The communications specification will be relocated to the TRM. Changes to the TRM will be controlled by the provisions of 10 CFR 50.59. This change is designated as a relocation because the LCO did not meet the criteria in 10 CFR 50.36(c)(2)(ii) and has been relocated to the TRM. REMOVED DETAIL CHANGES None LESS RESTRICTIVE CHANGES None CNP Units 1 and 2 Page 2 of 2 Attachment 1, Volume 14, Rev. 1, Page 159 of 188
Attachment 1, Volume 14, Rev. 1, Page 160 of 188 Specific No Significant Hazards Considerations (NSHCs) Attachment 1, Volume 14, Rev. 1, Page 160 of 188
Attachment 1, Volume 14, Rev. 1, Page 161 of 188 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS CTS 3/4.9.5, COMMUNICATIONS There are no specific NSHC discussions for this Specification. CNP Units 1 and 2 Page 1 of 1 Attachment 1, Volume 14, Rev. 1, Page 161 of 188
, Volume 14, Rev. 1, Page 162 of 188 CTS 3/4.9.13, Spent Fuel Cask Movement , Volume 14, Rev. 1, Page 162 of 188 , Volume 14, Rev. 1, Page 163 of 188 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) , Volume 14, Rev. 1, Page 163 of 188
Attachment 1, Volume 14, Rev. 1, Page 164 of 188 CTS 3/4.9.13
}LA./
REFUER IN OPERATIONS SPENT FU I CASK MOVEMENT LIMITING CONDITIONFO PRThl 3.9.13 "Mveentof the Spn auicsk above elevation 620 het shall be done t tho spent fuel cask andling crone operating 1 the Control d Path tode of operation APPLI ILITY: With fuel assembl as in the storage pool. U1th t requirements of th abo e specification not satis ad, place 1e crane load in a safe nditlon. The provisions f Specificu-tion 3 0.3 are not applicable. SURV. LLANCE REOUTREKENTS 493Crane inter'oi hc prevnent rasisng tha bot f the spent fuel cask rere than 6 lnche's aove the top of t'heask D pProtection Syst crclinter and restrict o crane's movewent to t Cotrolled Pat shall be demonstrated OP ILE within 7 days prio o crane opeation in the Controlled P th Mods and at least once r 7 days the after during crane opera Ion in the Controlled Pat Mode. O C. COOK - UNIT 1 3/4 9-17 Amendment No. 23 Page 1 of 2 Attachment 1, Volume 14, Rev. 1, Page 164 of 188
, Volume 14, Rev. 1, Page 165 of 188 CTS 3/4.9.13 REFUE ING OPERATINN SPENTlFUEL CASK NOVVENT It N6 CONDITION t OPE T 3.9. 3 Movennt of the spent t ecsk a above elevation 620 feet shall be ne with the spent fuel csu handling crane operating I the Cant oled Path Modi of operati APP CABILITY: With fuel &ss lies In the storage pool.
ACT ON: Wit the requirements of th Vs specification not s*tis aed, pla a the crane load in a safe ondition. The provisions f Specifics-ti 3.0.3 are not applicable. JVEILLUAhCE REUUIR6ENETS l _
.9.13 Crane Interlek wh ch pr vent raising-the bt t of thescpent uel cask more th n I 1nheb eth toP of the Calsk pro Pro tet1n tem cylinder-nd restric the crane' a ove ent to th Controllod *th shell be demonstratd PERAILE within 7 days prio to crane operation in the Controll Path Mode and at least onc per 7 days therafter during crane a ration In the Controlled Pe h mode.
1D.C. COOK - UNIT 2 l3/4 9-16 Page 2 of 2 , Volume 14, Rev. 1, Page 165 of 188
Attachment 1, Volume 14, Rev. 1, Page 166 of 188 DISCUSSION OF CHANGES CTS 314.9.13, SPENT FUEL CASK MOVEMENT ADMINISTRATIVE CHANGES None MORE RESTRICTIVE CHANGES None RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES LA.1 (Type 6 - Relocation of LCO, SR, or other TS requirement to the TRM, UFSAR, ODCM, QAPP, or lIP) CTS LCO 3.9.13 requires the movement of the spent fuel cask above elevation 620 feet to be done with the spent fuel cask handling crane operating in the Controlled Path Mode of operation. The ITS does not include the requirements for the movement of the spent fuel cask above elevation 620 feet. This changes the CTS by moving the explicit requirements for movement of the spent fuel cask above elevation 620 feet, including the Action and Surveillance Requirement, from the Technical Specifications to the Technical Requirements Manual (TRM). The removal of these details from the Technical Specifications is acceptable because this type of information is not necessary to provide adequate protection of public health and safety. The purpose of CTS LCO 3.9.13 to ensure that, during insertion or removal of spent fuel casks from the spent fuel pool; fuel cask movement will be constrained to the path and lift height assumed in the Cask Drop Protection System safety analysis. Restricting the spent fuel cask movement within these requirements provides protection for the spent fuel pool and stored fuel from the effects of a fuel cask drop accident. These requirements are proposed to be relocated to the TRM since the movement of loads other than fuel assemblies is controlled based on the heavy loads analysis. The bounding design basis fuel handling accident in the auxiliary building assumes a single irradiated fuel assembly is damaged. In addition, as stated in the NRC Safety Evaluation for License Amendments 197 (Unit 1) and 182 (Unit 2), dated July 12, 1995, the controls in place ensure that the potential for other, more severe events that could occur, such as a heavy load drop on irradiated fuel, need not be postulated and analyzed. This change is acceptable because the removed information will be adequately controlled in the TRM. The TRM is incorporated by reference into the UFSAR and any changes to the TRM are made under 10 CFR 50.59, which ensures changes are properly evaluated. This change is designated as a less restrictive removal of detail change because a requirement is being removed from the Technical Specifications. CNP Units 1 and 2 Page 1 of 2 Attachment 1, Volume 14, Rev. 1, Page 166 of 188
Attachment 1, Volume 14, Rev. 1, Page 167 of 188 DISCUSSION OF CHANGES CTS 314.9.13, SPENT FUEL CASK MOVEMENT LESS RESTRICTIVE CHANGES None CNP Units I and 2 Page 2 of 2 Attachment 1, Volume 14, Rev. 1, Page 167 of 188
Attachment 1, Volume 14, Rev. 1, Page 168 of 188 Specific No Significant Hazards Considerations (NSHCs) Attachment 1, Volume 14, Rev. 1, Page 168 of 188
Attachment 1, Volume 14, Rev. 1, Page 169 of 188 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS CTS 3/4.9.13, SPENT FUEL CASK MOVEMENT There are no specific NSHC discussions for this Specification. CNP Units I and 2 Page 1 of I Attachment 1, Volume 14, Rev. 1, Page 169 of 188
Attachment 1, Volume 14, Rev. 1, Page 170 of 188 CTS 3/4.9.14, Spent Fuel Cask Drop Protection System Attachment 1, Volume 14, Rev. 1, Page 170 of 188
, Volume 14, Rev. 1, Page 171 of 188 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) , Volume 14, Rev. 1, Page 171 of 188
Attachment 1, Volume 14, Rev. 1, Page 172 of 188 CTS 3/4.9.14
*LA.
II REFUELi OPERATIONS SPENT FUIL CASK DROP PROTECTION _ _ _ IP Dllt LIMITINJCONDITION FOR OPERATIOJ 3.9.14 xanw e1ght 1Te of a spent fuel cask used WI the Cask Drop Protect on Syste shallI be ila d to 110 tons (nominal APPLI ILITY: At all times. ACTION:l With tae requirements of the a ove specification not s tisfied, place the cr ne load in a safe condi ion. The provisions of Specification 3.0.3 re not applicable. SURVE LLANCE RggqLRtMENTS l 4.9 4 The weight of a spe fuel cask shall be ver fied to be < 110 tons nal prior to its use the3thCask Drop Protei1on Systeo. 2
£{C. COOKS - UNIT 1 l 3/4 9-18 Amendment No. 23 Page 1 of 2 Attachment 1, Volume 14, Rev. 1, Page 172 of 188
Attachment 1, Volume 14, Rev. 1, Page 173 of 188 CTS 3/4.9.14 REFU LINGOPOLTIONSI lSPt FUEL CASKDROP PROUECTISSYSTEM LINIRSI CONDITION FOR OPEM Ox 3.9 14 The maximum weight a £ spent fuel cask used ith the Cask Drop l Pro act10n Systi shell b 1 mitde to 110 tons (noo 1). APP ICAILITY: At al t1 AON: Iih t requiremnts of t above specification not satisfied. place t crane load in a safe ditlon. The provisions F Specification 3 .3 are not applicbic . SrEILLANCE RQIIRO Mll _
.9.14 The weight ofa s shall be fled to be
- 110 tons olnofM) pror10 to 1ts us fuel th cask Cask Drop -Prot io1n Syste.
P. C. COOK UNIT 2 3/4 %17 Page 2 of 2 Attachment 1, Volume 14, Rev. 1, Page 173 of 188
Attachment 1, Volume 14, Rev. 1, Page 174 of 188 DISCUSSION OF CHANGES CTS 3/4.9.14, SPENT FUEL CASK DROP PROTECTION SYSTEM ADMINISTRATIVE CHANGES None MORE RESTRICTIVE CHANGES None RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES LA.I (Type 6 - Relocation of LCO, SR, or other TS requirement to the TRM, UFSAR, ODCM, QAPP, or i1P) CTS LCO 3.9.14 specifies that the maximum weight of a spent fuel cask used with the Cask Drop Protection System be limited to 110 tons (nominal). The ITS does not include this spent fuel cask weight limitation associated with the Cask Drop Protection System. This changes the CTS by moving the explicit spent fuel cask weight limitation associated with the Cask Drop Protection System, including the Action and Surveillance Requirement, from the Technical Specifications to the Technical Requirements Manual (TRM). The removal of these details from the Technical Specifications is acceptable because this type of information is not necessary to provide adequate protection of public health and safety. The purpose of CTS LCO 3.9.14 is to ensure that limitations on the use of spent fuel casks weighing in excess of 110 tons (nominal) are in effect to provide assurance that the spent fuel pool would not be damaged by a dropped fuel cask since this weight is consistent with the assumptions used in the safety analyses for the performance of the Cask Drop Protections System. These requirements are proposed to be relocated to the TRM since the movement of loads other than fuel assemblies is controlled based on the heavy loads analysis. The bounding design basis fuel handling accident in the auxiliary building assumes a single irradiated fuel assembly is damaged. In addition, as stated in the NRC Safety Evaluation for License Amendments 197 (Unit 1) and 182 (Unit 2), dated July 12, 1995, the controls in place ensure that the potential for other, more severe events that could occur, such as a heavy load drop on irradiated fuel, need not be postulated and analyzed. This change is acceptable because the removed information will be adequately controlled in the TRM. The TRM is incorporated by reference into the UFSAR and any changes to the TRM are made under 10 CFR 50.59, which ensures changes are properly evaluated. This change is designated as a less restrictive removal of detail change because a requirement is being removed from the Technical Specifications. CNP Units 1 and 2 Page 1 of 2 Attachment 1, Volume 14, Rev. 1, Page 174 of 188
Attachment 1, Volume 14, Rev. 1, Page 175 of 188 DISCUSSION OF CHANGES CTS 3/4.9.14, SPENT FUEL CASK DROP PROTECTION SYSTEM LESS RESTRICTIVE CHANGES None CNP Units 1 and 2 Page 2 of 2 Attachment 1, Volume 14, Rev. 1, Page 175 of 188
Attachment 1, Volume 14, Rev. 1, Page 176 of 188 Specific No Significant Hazards Considerations (NSHCs) Attachment 1, Volume 14, Rev. 1, Page 176 of 188
Attachment 1, Volume 14, Rev. 1, Page 177 of 188 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS CTS 3/4.9.14, SPENT FUEL CASK DROP PROTECTION SYSTEM There are no specific NSHC discussions for this Specification. CNP Units 1 and 2 Page 1 of I Attachment 1, Volume 14, Rev. 1, Page 177 of 188
Attachment 1, Volume 14, Rev. 1, Page 178 of 188 ATTACHMENT 8 Improved Standard Technical Specifications (ISTS) not adopted in the CNP ITS Attachment 1, Volume 14, Rev. 1, Page 178 of 188
Attachment 1, Volume 14, Rev. 1, Page 179 of 188 ISTS 3.9.2, Unborated Water Source Isolation Valves Attachment 1, Volume 14, Rev. 1, Page 179 of 188
Attachment 1, Volume 14, Rev. 1, Page 180 of 188 ISTS 3.9.2 Markup and Justification for Deviations (JFDs) Attachment 1, Volume 14, Rev. 1, Page 180 of 188
Attachment 1, Volume 14, Rev. 1, Page 181 of 188 I ~sI [Unborated Water Source Isolation ValvSI . 3.9 REFUELING OPERATIONS 3.9.2 l Unborated Water Source Is tion Valves / EVIEWER'S NOTE - This Technical Specification Isnot r quired for units that have analyzed a boron dilutfn event In MODE 6. It is required for those nits that have not analyzed a boron dilution ev t In MODE 6. For units which have nqf analyzed a boron dilution event In MODE 6. the solation of all unborated water sources Is required to preclude lhis event from occurring. LCO 3.92 Each ye used to Isolate unborated water sources sha be secured In the cl d position. APPLICABILITY: M E 6. ACTIONS
-NOTE -
Separate Condition ndy Is allowed for each unborated water sour isolation valve. CONDI/ON REQUIRED ACTION COMPLETION TIME A. A.1 Suspend CORE Immediately OALTERATIONS. Require Action A.3 must becompleted AND whenevr Condition A is entereA.2 Initiate actions to ecure Immediately valve In closed sition. One r more valves not sec ed in closed AND pos ion. A.3 Perform SR .9.1.1. 4 hours ___I i WOG STS 3.9.2 -1 Rev. 2. 04130101 Attachment 1, Volume 14, Rev. 1, Page 181 of 188
Attachment 1, Volume 14, Rev. 1, Page 182 of 188 that isolates unborated water id In the dosed position. WOG STS 3.9.2 - 2 Rev. 2, 04130101 Attachment 1, Volume 14, Rev. 1, Page 182 of 188
Attachment 1, Volume 14, Rev. 1, Page 183 of 188 JUSTIFICATION FOR DEVIATIONS ISTS 3.9.2, UNBORATED WATER SOURCE ISOLATION VALVES
- 1. CNP has analyzed a boron dilution event in MODE 6. Isolation of all unborated water sources in MODE 6 is not required. Therefore, ISTS 3.9.2 is not included in the ITS.
CNP Units 1 and 2 Page 1 of 1 Attachment 1, Volume 14, Rev. 1, Page 183 of 188
Attachment 1, Volume 14, Rev. 1, Page 184 of 188 ISTS 3.9.2 Bases Markup and Justification for Deviations (JFDs) Attachment 1, Volume 14, Rev. 1, Page 184 of 188
Attachment 1, Volume 14, Rev. 1, Page 185 of 188 I B 3.9 REFUELING OPERA IONS B 3.9.2 [ Unborated Wa er Source Isolation VaNves] i BASES BACKGROUND Dun MODE 6 operations a]11 isolation valves for reactor akeup water 41 soures containing unborated water that are connected t the Reactor Coo nt System (RCS) must bie closed to prevent unpla ed boron dilution of the reactor coolant. The Isolation valves mus be secured in the plosed position. Th4 Chemical and Volume Cointrol System is capable f supplying beted and unborated water to the RCS through varinus flow paths. Si a positive reactivity addition made byredudng -le boron co centration is inappropriab during MODE 6, isolaf n of all unborated w er sources prevents an unplanned boron dilution I APPLICABLE T possibility of an Inadvertent boron dilution eve (Ref. 1) occurring SAFETY d ing MODE 6 refueling operations Is precluded b adherence to this ANALYSES L 0, which requires that potential dilution sources e Isolated. Closing th required valves during refueling operations pr ents the flow of ur orated water to the filled portion of the RCS. he valves are used to Is late unborated water sources. These valves ve the potential to in irectly allow dilution of the RCS boron cnce ltion in MODE 6. By i lating unborated water sources, a safety anasis for an uncontrolled be on dilution accident in accordance with the andard Review Plan 4 (F A 2) is not required for MODE 6. Tle RCS boron concentration satisfies Criteri 2 of 41 1 CFR 50.36(c)(2)(ii).
.4 LCO Tt is LCO requires that flow paths to the RC from unborated water .i se rces be Isolated to prevent unplanned b ron dilution during MODE 6 14 al d thus avoid a reduction In SDM.
APPLICABILITY In MODE 6, this LCO Is applicable to prev nt an inadvertent boron di lion event by ensuring Isolation of all urces of unborated water to I ) RCS. F r all other MODES, the boron dilution ccldent was analyzed and was I fo ind to be capable of being mitigated. WOG STS B 3.9.2 - l Rev. 2 04/30101 Attachment 1, Volume 14, Rev. 1, Page 185 of 188
Attachment 1, Volume 14, Rev. 1, Page 186 of 188 [Unborated Water Source islation Valves] { I / B 3.9.2 l BASES ACTIONS The A ONS Table has been modified by a Note that [lows separate Conditi entry for each unborated water source isolation valve. Conti tion of CORE ALTERATIONS Is contingent pon maintaining the unit I compliance with this LCO. With any valve usd to Isolate unbo ted water sources not secured in the closed osition, all opetions involing CORE ALTERATIONS must b suspended immtiately. The Completion Time of "immediatel for performance of Re Ired Action A.1 shall not predude completion f movement of a . cconert co, to a safe position. Co dition A has been modified by a Note to requ that Required A on A.3 be completed whenever Condition A I entered. eventing inadvertent dilution of the reader olant boron concentration dependent on maintaining the unborated w er Isolation valves secured losed. Securing the valves in the closed pos on ensures that the valves nnot be inadvertently opened. The Compi ion Time of "immediately requires an operator to Initiate actions to do an open valve and secure the isolation valvein the dosed position im diately. Once actions are Initiated, they must be continued until the es are secured in the closed position. Due to the potential of having diluted the ron concentration of the reactor coolant, SR 3.9.1.1 (verification boron concentration) must be performed whenever Condition A Isent ed to demonstrate that the required boron concentration exists. T Completion Time of 4 hours Is
.1Z
! {
sufficilent to obtain and analyze a reaco coolant sample for boron concentration.{ SURVEILi~ SR 3.9.2.1l REQUIREMEiS These valves are to be secured CIOse to isolate possible dilution paths. The [ikelihood of a significant reductn in the boron concentration during 1MODE 6 operations is remote due the large mass of borated water in flZthe refueling cavity and the fa h tall unborated water sources are l islate, prcludng adilution. Th boron concentration is checked every l 72 hours during MODE 6 under 3.9.1.1. This Surveillance WOG STSl B 3.9.2 - 21 Rev. 2, 04130101 Attachment 1, Volume 14, Rev. 1, Page 186 of 188
Attachment 1, Volume 14, Rev. 1, Page 187 of 188 [Unborated Water Sour Isolation Valves] I B 3.9.2 BASES SURVEILLANCE REQUIREENTS (continued) l demnstrates that the valves are closed through a s walkdown. nem The 1 day Frequency Is based on engineering Judgrrent and Is con dered reasonable in view of other administratvco ntrols that will ensre that the valve opering Isan unlikely possibilit. I I1 REFERENCES 1. FSAR. Section 1152.4].
- 2. NUREGO0800. Section 15.4.6.
j .4
/ N1-4 -1
.1 Ii W\OG ~STS B 3.9.2 - 3 Rev. 2, 04/30101 Attachment 1, Volume 14, Rev. 1, Page 187 of 188
Attachment 1, Volume 14, Rev. 1, Page 188 of 188 JUSTIFICATION FOR DEVIATIONS ISTS 3.9.2 BASES, UNBORATED WATER SOURCE ISOLATION VALVES
- 1. Changes are made to be consistent with changes made to the Specification.
CNP Units 1 and 2 Page 1 of I Attachment 1, Volume 14, Rev. 1, Page 188 of 188
Attachment 1, Volume 15, Rev. 1, Page i of i
SUMMARY
OF CHANGES ITS CHAPTER 4.0 Change Description Affected Pages A self-identified change to ITS 4.3.1.2.a has been made. This Page 38 of 45. change makes an editorial revision to ITS 4.3.1.2.a. Attachment 1, Volume 15, Rev. 1, Page i of i
Attachment 1, Volume 15, Rev. 1, Page 1 of 45 VOLUME 15 CNP UNITS 1 AND 2 IMPROVED TECHNICAL SPECIFICATIONS CONVERSION ITS CHAPTER 4.0 DESIGN FEATURES Revision I Attachment 1, Volume 15, Rev. 1, Page 1 of 45
Attachment 1, Volume 15, Rev. 1, Page 2 of 45 LIST OF ATTACHMENTS
- 1. ITS Chapter 4.0 Attachment 1, Volume 15, Rev. 1, Page 2 of 45
, Volume 15, Rev. 1, Page 3 of 45 ATTACHMENT I ITS Chapter 4.0, Design Features , Volume 15, Rev. 1, Page 3 of 45 , Volume 15, Rev. 1, Page 4 of 45 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) , Volume 15, Rev. 1, Page 4 of 45
Attachment 1, Volume 15, Rev. 1, Page 5 of 45 ITS 4.0 ITS 4.0 5.0 MIsZ01 TEATORs 4.1 .1. gm tCeTLSTOW AM 4.1.1 s.1.1 The WMIusoas ax" "1 s shown TLC=& 5.1-1. im ztimano tOn all the land w~ithin a circle 4.1.2 5.1. Th 16 unsshal poulsuon b4*~ . 3-2 containment strulures and 1.2l telou~ ? @stslmd L U ua radius of 2 miles fieSawr Ter Cssos imdia ufnwts 4.1.1 S.1.3 The SIT? in Tiguxe 5.1-3. M0h1 form eons o" 124qG4A 4Muans shell lie as show* I1 5.1. ahctor coueaiauti La a steel t, :w oiteed concrcet ding of c7lLud=a shps. with
- dis rc %w uhe folloeint ign foturs:
*. 1 isid dae - 11.5 fest.
b. e. d. le WfS#L ide theshs aclaass o c
- 160 ZEm.*
to ells 324 09 lmz concrets reof IV6
- a. XlLu thickness consrete floor pad 10 feet.
- f. lod.nal thicztess t steel liner, Bldg dm - 3/3 Inches.
- g. Nominal hickenss I steel liner, h - 1/4 inch.
.h.l Itae ra olme - .14 10 cubic fee Jroe eidde (Clv.40t1) toinsid of does.
- COOX NUC..A !&ST.
- C= L 5-1. ?w:rr YO ilr 16a Page 1 of 25 Attachment 1, Volume 15, Rev. 1, Page 5 of 45
Attachment 1, Volume 15, Rev. 1, Page 6 of 45 ITS 4.0 ITS Figure 4.1-1 STATE OF MICHIGAN
)
O2 CL ?:A:-.. - : 5-2 AMDENOI :1v. *55 Page 2 of 25 Attachment 1, Volume 15, Rev. 1, Page 6 of 45
Attachment 1, Volume 15, Rev. 1, Page 7 of 45 ITS 4.0 ITS
- 0. C. COOK-UNIT I 5-3 Acendmat No. 73 Page 3 of 25 Attachment 1, Volume 15, Rev. 1, Page 7 of 45
Attachment 1, Volume 15, Rev. 1, Page 8 of 45 ITS 4.0 ITS 0 5.0 IDESGN FEAUTUR 4.2 51 RFACOR MCOR amatrx of FLEL MWFRI-4.2.1 53.+/-1 Tberatcore shal cotalin193eaael whithchfel _c
*M lfcvJo Zl exclRonegp OM Ibited N0stfton of rzmh T stainless stee5We uel rods with an . od. In accordance Cappoed applcstiof hel rod confituratoos. twy be ted. Futd Initial assem l be lined to thdw deIps tht hne been analyzed wit appicable NRC IA.
Composibton ot . ff-appd des a m and bodn by tests or analysis to coaply ith an felfi eqy desin Inennrched UO2s bases. A lihited nwrber of kid test assemblies t have n leteding m be uhelmaterial l ae in pEacht . fI rod shallvsy a noe nal active fuel I of 144 0In 4... eblet Co .952 w a bavt
, Y.lP of 33p Weiet 43.1.1.sllb
.... _ dnmgr in pbyh dek oII arichrt ct of 4.95iwlvdM pervert U-233. theIntW con loadWS and hAMl bm rft iino~m COMROL ROt] ASSFMRT IPA 4.2.2 5.3.2 Mie reactor core duD contain 53 u knl and to W kno control rod asembles. I* sw kleth conrol rod es Wall conut a na ini l 142 Inc of abori b er ml. lbe alues of absorber mnir D be t0 percent sole, 15 sc ndium and 5 perent cadmim ,AD control rods shalee c d k eubtn ItU The control material shall l A3 be silver Indium cadmium.
as approved by the NRC.
,COOK N'VCLEAR PLANT.U- I Page 54 AM MNDN 1T 3, ;w, u,aM 239 Page 4 of 25 Attachment 1, Volume 15, Rev. 1, Page 8 of 45
Attachment 1, Volume 15, Rev. 1, Page 9 of 45 ITS 4.0 ITS 5.0 DESIGN FEATURES SA.D IED colt systcmI ded al be mainl/ae/ LA.4
/Dhhspecid in Secdo.. 41 ed PSAR. with aWwace for DOM dctadation to the appliabb Suraa mts.
- h. Fopr */mof C. / aw o ,penmM 5.5.1 acy C= eenf in c5llW Wllbbe an e/~txac ith tho lbs oww"o prigna LS gJ*p jc.bf Sarcllance wiR one ciero. Tha is lb CVC baro ac apcanddthmurr. / /
4.3 55FU.STEORFUE 4.3.1 WrWxALMr-91ENTZEJR1 4.3.1.1 541.1 Tespmst bmrag e ,hb ded smd abe maintained wib. 4.3.1.1.b M A kw Cqtdyakm to e1 ta 0.95 wbun floded with unbwWd war. 4.3.1.1.c b. A ,na 97 Incheavomr ditice bet fba as seeshliampOd aced See ITS
~3.7.16J 4.3.1.1.d, C. TW fWmassembl~e wI be clasfldo acceptable for Reglee IJ RaiF 2 orR37.16 JWr~
4.3.1.1.e bwentupon thak sAmumlywa bimpwarW Imdebit aoizdi emicbw Calls weepuble for R4- 1. Ra 2. id R*Oc 3 m-bty rap blahdk Pl 5MS-1 and SA-2. Asasn g w ptebb for sb gp in Ragw . R c d mdm dae design uk 2ta defithe uelna fows: y S3 S 3.7.16 COOKNUCLUARYLAT-UNrr I p.ma 5 AUDZNUK3TIISIAM.3&3,3U4 241 l Page 5 of 25 Attachment 1, Volume 15, Rev. 1, Page 9 of 45
Attachment 1, Volume 15, Rev. 1, Page 10 of 45 ITS 4.0 ITS 0 5.0 DESIGN FEATURES 5.6 FUEL STORAGE (Continued) 4.3.1.1.a, 1. Region I is designed to acconmnodate new fuel with a maximnum nominal enrichment of 4.95 4.3.1.1 .d wt'Y U-235. or spent fuel regardless of the discharge fuel burnup.
- 2. Region 2 is designed to accommodate fuel of 4.95% initial nominal enrichment burned to at 4.3.1.1.e least 50 000 MWD/MtU, or fuel of other enrichments with equivalent reactivity.
- 3. Region 3 is designed to accommodate fuel of 4.95% initial nominal enrichment burned to at least 38.000 MWD/MtU. or fel of other enrichments with equivalent reactivity.
The equivalent reactivity criteria for Region 2 and Region 3 is defined via the following equations: For Reiion 2 Storate -4 See ITS 3.7.16 Minimurn Assembly Average Bumup in MWD/Mll -
-22,670 + 22,220 E - 2,260 E+ 149 E' For Region 3 Storage Mininurm Assembly Average Burnup in MWD/MTU -
.26,745 + 18746 E - 1,631 E' + 98A E' Where E - Initial Peak Enrichment COOK NUCLEAR PLANT-UNIT I .XaV.6 AMEMDMENTS.434,44443,3443 33243 Page 6 of 25 Attachment 1, Volume 15, Rev. 1, Page 10 of 45
Attachment 1, Volume 15, Rev. 1, Page 11 of 45 ITS 4.0 ITS e Figure 4.3-1
. :MMMD MMKX3r-M=--W KZr.W3CX=M.
MC2rr=rrTM1 I wnkrr=rr -
!1093COOCICi VMDK=M=
03C30CX3M MMC3C=KMCX . MinrIlMnOMM 90nonvIlf2m IC222200DESCE 2rlywrWIVIEW-10 Mr,"IrrTywym 113931r::M:M= 5309 N=X=301 0=0EMMOM 70=300 N= I" KM I CMU 13 [Z zc= 3ma A.' D.~! . CCN~ ~ ~ I COCK . .? -
?:.A::+/- 5.7 Page 7 of 25 Attachment 1, Volume 15, Rev. 1, Page 11 of 45
Attachment 1, Volume 15, Rev. 1, Page 12 of 45 ITS 4.0 ITS 0 Figure 4.3-2 Zun"wm au mmn mus C3~ t mu 3 [s m j j b5D I COCKc NUCLUZ ?L*wr mi:: 5 .7& AM"fMW "~. 189 Page 8 of 25 Attachment 1, Volume 15, Rev. 1, Page 12 of 45
Attachment 1, Volume 15, Rev. 1, Page 13 of 45 ITS 4.0 ITS 5.L DESIGN FATURES Rpm 5.6-3 Intenioaally deletuL COOKNUT "PIANT.Urr 1 Pp5Thr AMNDIMries , 243 Page 9 of 25 Attachment 1, Volume 15, Rev. 1, Page 13 of 45
Attachment 1, Volume 15, Rev. 1, Page 14 of 45 ITS 4.0 ITS S.0 DESIGN FEATURE 56 TUEL S=RAGF frIbiUed) CRIT-A} I - 4.3.1.2 5.6.2 Th new e storge rocks ve deslped and shaf be maiisshaed wlth. 4.3.1 .2.a L Westiotboue fe assemblins hfin eter a umxe enrlkhnMe of 4.55 wei& S U.235. or an adchm an bewe 4. and 4.9 weIgh S U-235 wilh greatr an r eful to the tanitxan .nber of itgd Nel boabl sboem pi as shown on Firate 5."Ontepotsloa of the Boo-10 lodb* between l.X and I X and beten J1X and 2.OX is accepable) 4.3.1.2.b b. K, s 0.95 if U flooded wh d water, which fachudes an al0wance kgr unctal ite a described In Section 9.7 of OAeUFSAR. 4.3.1.2.c C. s SO.98 tf modeed b ab u f , which includes an allowaee for mealtks as scribd inSectio 9.7 of fte UFSAR; and 4.3.1 .2.d d. A omina 21 inc cer to ceter dstac betwe Meassenmblies placed Inthe storage racks. DBA~ 4.3.2 5.63 Th spent el strage pool Isdesiged ad shall be malntned to Peve inTen drawn of the pool below elevation 6294. COON NIUCLLR PLANT4RIIT I rp"I AMEM~MUNT U30,SU,2A 239 Page 10 of 25 Attachment 1, Volume 15, Rev. 1, Page 14 of 45
Attachment 1, Volume 15, Rev. 1, Page 15 of 45 ITS 4.0 ITS 0 Figure 4.3-3 5.6 DESIGN FATURES F4 i5.6-4: New I Sore Rack Ukea FI umiabe Aborbr (UFA) Replr 32 -
- 1.0:
- - 1.6:
*... 2.0:
24 - I--- 8-- C - -- o 4.60 PAi 4.60 4.70 4.80 4.90 4.96 5.00 3'u Enrichment (wlo) COOK NtXcEA PFLAr-UNIT I AIMM.ENW 239 I Page 11 of 25 Attachment 1, Volume 15, Rev. 1, Page 15 of 45
Attachment 1, Volume 15, Rev. 1, Page 16 of 45 ITS 4.0 ITS 0 S.0 - DFSIGN FEAUEML 5.6 FUEL STORAGE (Cmtyffued) I 4.3.3 5.6.4 Mhe ud tmppolisdesiedanudshlbenmsd nwidastoaecitylmitdto i ,m 3613 t adIbes. 5.7.1 , andml as CactorqI lbs FSAR s1ll be demoed godyp ftoueacp Ssde Comalnod In the R _~lee./ W low&l foprolnormal i o ii9 A
- u-a F. r . 3 5.J.I eodoiab u1}53-3.
Wb oae nft 5. I COOK NUCLEAR PLAN4JUNrr I Page 5.9 AMENDDEW60, 4ai 486, 201-Page 12 of 25 Attachment 1, Volume 15, Rev. 1, Page 16 of 45
e C
-n c , fi 9.4.
2) 2 S B 3 I a CD 3 9-. 0 I. 0 2 CD
- 02
- .I (D
2-u CD CD a I. 3 CD CA CD en 10 10 -4' 0i 0)
-0 CD I -4' 07 -4 0w ci en 0D
Attachment 1, Volume 15, Rev. 1, Page 18 of 45 ITS 4.0 ITS 4.0 t.0 DU2CN( JAU'LS 4.1 5.1 eS. 4.1.1 5.1.L th ecus io ax" shl ea ho a YLir 5.1-1. I ow ulattoC Zone all the land ithin circle centered on thereadtor _G 4.1.2 3.1.2 The low population oe shalle b1 las s m J 5.i containment structures and a radius of 2 miles Mse am rT-To "4es M td futs 4.1.1 4.1.3 Us1 SM BOMAU forw sassaw sud lUqui effluents shell, b1 as s*awa La YLg~e 5.1-3. I 5.2.1 canetdsa follovin 52 s"cor ee dlg ef C ksanaL"tozas:
=ith ae J
dis hap., a
*teal a dam L .4 reladoced Jf sad hSav the
*. lI 5LasUds a1r - 213 fset.
- b. Komfnl L a h bt
- 160 fast.
C lLtL= thehk a coGsCts walls- i 3' G9
- d. XqaJ thickess Cocrete roof - 2 S
.l ULawn tehcksa £ camcrete floor pd 10 feet.
f 11s1 athickaness msel linar - 3/1 ch. l lte frla Talmo 14 x 10 cutc fee ZStCV a*AND 5.2.2' lhe reactor cousatna bltd 1 desl and shall be Sainta In accordance v the rgn des praoisota coaaized La Sac iou 5.2.2 of the I COOK NUCILLA LAIr - MaT. 2 5.1 AMLCnKUT' NO. SZ, IS: Page 14 of 25 Attachment 1, Volume 15, Rev. 1, Page 18 of 45
Attachment 1, Volume 15, Rev. 1, Page 19 of 45 ITS 4.0 ITS 0 Figure 4.1-1 STATE OF MICHIOAN I *
*I.
- I C20I =, w. ? ,-,_
- 5-2 AMENCMEN- NO. 172 Page 15 of 25 Attachment 1, Volume 15, Rev. 1, Page 19 of 45
Attachment 1, Volume 15, Rev. 1, Page 20 of 45 ITS 4.0 ITS
- a. C.COaK-UNIT 2 5-3 heuent No. 41 Page 16 of 25 Attachment 1, Volume 15, Rev. 1, Page 20 of 45
Attachment 1, Volume 15, Rev. 1, Page 21 of 45 ITS 4.0 ITS 0 5.0 DESIGN FEATURES 4.2 5.3 REACTOR CORE [consisting of a matrix ofl iUL ASSEMBL 1 4.2.1 5.3.1 The reactor core shall contain 193 fuel assemblies with each fuel assembly c a fue cld l obdajircaloejcr ZIRLO except that limited substitutions of zirconium alloy or stainless steel filler rods, LI Initial in accordance wit Mapproved applications of fuel rod configurations. may be used. Fuel assemblies composition of shall be limited to those fuel designs that have been analyzed with applicable NRC staff-approved codes natural or slightly and methods, and shown by tests or analyses to comply with all fuel safety design bases. A limited number enriched U02 as of lead test assemblies that have not completed representative testing may be placed in non-lirmiting core fuel material regions. [Each fuel rrd shall have a nortinal active lueylen 44 inches. Thc initi core ing sha 4.3.1.i.a I have a maximum eptichment of 3.3 weight percent 0235.r Reload fuel shall be similar in physical design 4311ato the initial core loading and may be nominally enriched up to 4.95 weight percent U-235. CONTROL ROD ASSEMBLIES 4.2.2 5.3.2 The reactor core shall contain 53 full len and no part length control rod assemblies. e full length control rod assembes shall contain a nomin 142 inges of absrer materi h nop nal values o g0 absorber materialm s be percent silver, 15 percept indium and 5 percent cadmium. /All control rods shall be clad wi stainless steel tubing. I I ryeesilver control material shall Indium cadmium. l A.3 5 4 REACT1OR COOLANT SYSTM be iapproved by the NRC. DESIGN PRESSURE/AND TEBMPERATUJRE 5.4.1 'Me reactor olant system is designed and shall be Maitaine
- a. Il cordance with the code requireme, spcfied in Section 4.1.6 of the FSAI, Or normal degradation pursuant to the appicabic Surveillance Requirements
- b. For a pressure of 2485 psig. and
- c. For a temperature of 650°F. except for pressurizer which Is 680 0F.
COOK NUCLEAR PLANT-UNIT 2 Page 54 AMENDMENT 58,444, 44, M 494, 220 Page 17 of 25 Attachment 1, Volume 15, Rev. 1, Page 21 of 45
Attachment 1, Volume 15, Rev. 1, Page 22 of 45 ITS 4.0 ITS 5$0 DWGN VATURIS 4.3 5.6 FM ST6RAUIR 4.3.1 4.3.1.1 4.3.1.1.b L. A Kw qulvalctntosMutu 0935 when flooded widh umbometed wiae. 4.3.1.1.c b. A nmcingl 8.97-lndi ce .wrcsted destane between hal esemblics, placed In th stoempr
- 4 3.7.16 4.3.1.1.d, 4.3.1.1.e beseid upon, i6k aembl bum v, ~ Cells ccptabh f Retion Ieemdtmaa1evcet.-
1.Regiost 2. en Resion 3 asmbm* era aesg lceed In FiuvsSi-1 and 5.6-2. Assemnblies digt ccmb for I ege [ARegion 1Rzion3 i monset mm d design aiterla
&Ktdaedmftglonsuas Moie:m See ITS3 4.3.1.1.a. I. Region I Is desiged to maaomsc wfe i~aex~m an1ercueto 3.7.16 4.3.1.1.d 4.95 %llU-233.orspoot fuelregu~asada ulbmw*o.
Weug
- 2. Region,2 kdladedgiaso amoommodaW fuel of CM5l UNAs nomina earkbwmntbqumd to at Ion A0M0 MWDUMT. or faed of otie sockeets %ods equidulet reactiviy. See ITS 4.3.1.1.e
- 3. RegimioIs dadgned to mcs sdrIoel of 4.95% iniie nomins ei ben 3.rned6 COOKNUCIUfZR ?AIT-UMf2 rap ei A U 4U5u 4 4I 3&5l 3 222 Page 18 of 25 Attachment 1, Volume 15, Rev. 1, Page 22 of 45
Attachment 1, Volume 15, Rev. 1, Page 23 of 45 ITS 4.0 ITS 50 D=1;N M&TUR 4.3.1.1.e Toe equikst macdvity u w Resion 2 and Resica 3 is lvia dw obwyng equnoxAaal wu 2Dn~IMDM1 For Remo SBona See ITS J 3.7.16 )
- 22670 + 22.220 B - 2.260 E + 1495'
' Fr Reion 3 Stonu M==AanafAwzapDw InpMWDIU -
.2&745 + 1.746E-.1.63le ++9A4le Whm B - 2dWPeak Ewld COOKtUCLInARGLAntUMs P.V54 Ab9M9r4&416""4, W&U3S24 Page 19 of 25 Attachment 1, Volume 15, Rev. 1, Page 23 of 45
Attachment 1, Volume 15, Rev. 1, Page 24 of 45 ITS 4.0 ITS Figure 4.3-1 TNUM 3.6-1: Scorage So1 hs:trn (Lzxed ThreesZa) I ^ iRl^ - M.P1 ! . le! 1 HIffll rF ^enw e sc+^ nne - + gr 5 sffl ..... i= _,, fE . t
=
i_. [,, ... am Tr :rr
.
- r .
Wn W w Tr TrT_ TrT rtr . [1 :r , _ ... , . : srtw _ ... .. z r: L. r I,,,, _
.,, r_ rs . s _ I . . s .- r TrT ._ . . r I_ . . r . _ - - . . w. r . . lI _
7 3 -_rrr ;sr ... .. ... r _ xe s _ _al Xl r _rsa f rT . . I I I r . I . LTrRlTI rT . T . .r r i . i_, o, , _nTrrT rT rz _ r- wTr L _: te X1 _ _ __s _=- _ _ w Ph i i il x i .,. *S1 . s s . x l: 1 rr _W^i. [ L LE i *i? i _ 11 _ :.wsss [o Li __ s a DITW X iX_ x . _ i r- X oi iEL is u u, - iB . _____fi__ t__r_
- _ _ . _ - _
rT I - - { I l l u l - - z l - l x l - l l - l E - . . { R x X >x . X T ro l l l T rT l [ r l [ . X X l l [ E r l l u l t s =_X_ Ulz. _-ses iis ._ __ _ . x r_< _ rX l T]Iii ,s_ _ r v.w _-
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_. i { x r_[T]
- i i i..- 1 i ] . i wN7 i i i
_ .s .... rw_ _ _ _ _ rx r r _ z rrwsE rTl U C r . rr..7i r
- i . 7i7 . . . - X . i i i j l . [ . . X . X. s . rs TrrrTrrTT * *. . E . X. [rrr77 sx. rrx] Xl . zw irTTrl-r 171W1 z s zw rrtr}
r s7 _ -7 _r _ __r-E _ _ _
* ' ' ffi l nn r r ta I [ T rrrTl ra _1 ffi Xl t__r__
x i I i i i i z l . l i . i i 7 _ i i l z _i s z 5 E s . i i i i i __ s rr sl .... L i, __zI X. s ri ll rT^or {s_]s[r[] *l<. r.rr - s E_ ris .7n . s sn l w ii7Sx w.a*
- I _ X s iiiis7i _ E s sw W x X E l X r r l , _ X.L7r ur x x _ r/r X X x . s
- L - E* . * .. i i i i . . i _ . . . E. . . . i _
U J a i i r >g__ sn_ rJ i i t 1 s i i i i _7 s x x x s x _ _ i-i s ai i -- i {, _7ss^er
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1% 7 3 tML COOK STCLM PLAT T W IT22. 5.7 AU SMIT rO.ra7, 152 Page 20 of 25 Attachment 1, Volume 15, Rev. 1, Page 24 of 45
Attachment 1, Volume 15, Rev. 1, Page 25 of 45 ITS 4.0 ITS Figure 4.3-2 Figure 5.6.2: Ig~cerLu Storsle P1auMn (Checkerboard) i I I i 0=133OWN drME200mr3cls MEMEMMIMCM t I 11 . I. -1un wum 3 SI - I m C3 I wL COOK NUCLEAA PLAiNT UNI 2 5.7& ANDVWINT NO. 152 Page 21 of 25 Attachment 1, Volume 15, Rev. 1, Page 25 of 45
Attachment 1, Volume 15, Rev. 1, Page 26 of 45 ITS 4.0 ITS 0 Sfl hn lmWATL&T . PIgff35.6.3 bzwlioi.Uy deWed.
'VOKINUCaZ"PLAWNrr2 AImwwrr si, 224 Page 22 of 25 Attachment 1, Volume 15, Rev. 1, Page 26 of 45
Attachment 1, Volume 15, Rev. 1, Page 27 of 45 ITS 4.0 ITS 5.0 ADMINISTRATIVE CONTROLS 5.6 FUEL STORAGE (Continued) 4.3.1.2 5.6.2 The new fuel storage racks are designed and shall be maintained with: 4.3.1.2.a a. Westinghouse fuel assemblies having either a maximum enrichment of 4.55 weight 'r U-235, or an enrichment between 4.55 and 4.95 weight % U-235 with the minimum number of integral fuel I burnable absorber pins as shown on Figure 5.6-4 (interpolation of the Boron-10 loading between l.OX and 1.5X and 2.OX is acceptable); 4.3.1 .2.b b. kit < 0.95 if fully flooded with unborated water, which includes an allowance for uncertainties as described in Section 9.7 of the UFSAR: 4.3.1.2.c c. kdt S 0.98 if moderated by aqueous foam, which includes an allowance for uncertainties as described in Section 9.7 of the UFSAR; and 4.3.1 .2.d d. A nominal 21 inch center to center distance between fuel assemblies placed in the storage racks. DRAINAG 4.3.2 5.6.3 The spent fuel storage pool is designed and shall be maintained to prevent inadvertent draining of the pool below elevation 629'4". CAPACITY 4.3.3 5.6.4 The spent fuel storage pool is designed and shall me maintained with a storage capacity limited to no X than 3613 fuel assemblies. COOK NUCLEAR PLANT-UNIT 2 Page 5-9 AMENDMENT I1,443,U7,452, 486, S4n,3, 261 Page 23 of 25 Attachment 1, Volume 15, Rev. 1, Page 27 of 45
Attachment 1, Volume 15, Rev. 1, Page 28 of 45 ITS 4.0 ITS 0 Figure 4.3-3 S.0 DESIGN FEATURES Figure 5.6-4: New Fuel Storage Rack Integral Fuel Burnable Absorber (IFBA) Requirements s 32 24
.b 0E Q .
L 16 02 8 0 4.50 4.60 4.70 4.80 4.90 4.95 5.00 235U Enrichment (w/o) COOK NUCLEAR PLANT-UNIT 2 Page 5-9. AMENDMENT 220 Page 24 of 25 Attachment 1, Volume 15, Rev. 1, Page 28 of 45
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Attachment 1, Volume 15, Rev. 1, Page 30 of 45 DISCUSSION OF CHANGES ITS CHAPTER 4.0, DESIGN FEATURES ADMINISTRATIVE CHANGES A.1 In the conversion of the CNP Current Technical Specifications (CTS) to the plant specific Improved Technical Specifications (ITS), certain changes (wording preferences, editorial changes, reformatting, revised numbering, etc.) are made to obtain consistency with NUREG-1431, Rev. 2, "Standard Technical Specifications-Westinghouse Plants" (ISTS). These changes are designated as administrative changes and are acceptable because they do not result in technical changes to the CTS. A.2 CTS 5.1.2 states "The low population zone shall be as shown in Figure 5.1-2." CTS Figure 5.1-2 provides a map depicting the low population zone. ITS 4.1.2 provides a description of the low population zone; a figure is not provided. This changes the CTS by providing a word description of the low population zone instead of a map. This change is acceptable since it does not change the current requirements. A description is provided consistent with the current map in the figure. This change is designated as administrative because it does not result in a technical change to the Technical Specifications. A.3 CTS Figures 5.6-1 and 5.6-2 provide drawings that depict the various regions of the spent fuel storage pool racks for a normal storage pattern (mixed three zone) and for an interim storage pattern (checkerboard). The key at the bottom of the figures identifies the total number of cells for the various regions. The CTS Figure 5.6-1 key identifies, in part, that there are 1415 Region 2 cells and 1694 Region 3 cells, and the CTS Figure 5.6-2 key identifies, in part, that there are 1415 Region 2 cells and 1379 Region 3 cells. The ITS Figure 4.3-1 key identifies that there are 1439 Region 2 cells and 1670 Region 3 cells, and the ITS Figure 4.3-2 key identifies that there are 1439 Region 2 cells and 1355 Region 3 cells. This changes the keys to clearly identify the actual number of cells depicted in each region. This change is acceptable since it does not change the current requirements. The number of cells listed in the keys for the ITS Figures is consistent with the actual number of cells depicted by the Figures. This change is considered administrative because it does not result in a technical change to the Technical Specifications. MORE RESTRICTIVE CHANGES None RELOCATED SPECIFICATIONS None CNP Units 1 and 2 Page 1 of 4 Attachment 1, Volume 15, Rev. 1, Page 30 of 45
Attachment 1, Volume 15, Rev. 1, Page 31 of 45 DISCUSSION OF CHANGES ITS CHAPTER 4.0, DESIGN FEATURES REMOVED DETAIL CHANGES LA. 1 (Type 1 - Removing Details of System Design and System Description, Including Design Limits) CTS 5.2 describes the various design features of the reactor containment building. The ITS does not contain this information. This changes the CTS by moving the description of the reactor containment building to the UFSAR. The removal of these details, which are related to system design, from the Technical Specifications is acceptable because this type of information is not necessary to be included in the Technical Specifications to provide adequate protection of public health and safety. The ITS still retains requirements on containment OPERABILITY in ITS 3.6.1. Also, this change is acceptable because the removed information will be adequately controlled in the UFSAR. Any changes to the UFSAR are made under 10 CFR 50.59 or 10 CFR 50.71 (e), which ensures changes are properly evaluated. This change is designated as a less restrictive removal of detail change because information relating to system design is being removed from the Technical Specifications. LA.2 (Type 1 - Removing Details of System Design and System Description, Including Design Limits) CTS 5.3.1 contains details of fuel assembly design, such as number of fuel rods per fuel assembly, the fuel rod nominal active fuel length, and the initial core loading maximum enrichment. The ITS does not contain these details, but provides a general statement that, "Each assembly shall consist of a matrix of Zircaloy or ZIRLO fuel rods with an initial composition of natural or slightly enriched uranium dioxide (UO2) as fuel material." This changes the CTS by moving the detailed description of the fuel assemblies to the UFSAR. The removal of these details, which are related to system design, from the Technical Specifications is acceptable because this type of information is not necessary to be included in the Technical Specifications to provide adequate protection of public health and safety. The ITS still retains requirements on fuel assembly enrichment in ITS 4.2.1. In addition, core power distribution requirements, which are dependant upon fuel assembly design, are described in ITS Section 3.2. Also, this change is acceptable because the removed information will be adequately controlled in the UFSAR. Any changes to the UFSAR are made under 10 CFR 50.59 or 10 CFR 50.71(e), which ensures changes are properly evaluated. This change is designated as a less restrictive removal of detail change because information relating to system design is being removed from the Technical Specifications. LA.3 (Type 1- Removing Details of System Design and System Description, Including Design Limits) CTS 5.3.2 contains details of control rod design, such as the nominal length of absorber material, percentage of each absorber material, and control rod cladding material. The ITS does not contain these details, but provides a general statement that, "The control material shall be silver indium cadmium, as approved by the NRC." This changes the CTS by moving the detailed description of the control rod assemblies to the UFSAR. CNP Units 1 and 2 Page 2 of 4 Attachment 1, Volume 15, Rev. 1, Page 31 of 45
Attachment 1, Volume 15, Rev. 1, Page 32 of 45 DISCUSSION OF CHANGES ITS CHAPTER 4.0, DESIGN FEATURES The removal of these details, which are related to system design, from the Technical Specifications is acceptable because this type of information is not necessary to be included in the Technical Specifications to provide adequate protection of public health and safety. The ITS still retains requirements on the control rod material in ITS 4.2.2 and on control rod OPERABILITY in ITS Section 3.1. Also, this change is acceptable because the removed information will be adequately controlled in the UFSAR. Any changes to the UFSAR are made under 10 CFR 50.59 or 10 CFR 50.71(e), which ensures changes are properly evaluated. This change is designated as a less restrictive removal of detail change because information relating to system design is being removed from the Technical Specifications. LA.4 (Type 1 - Removing Details of System Design and System Description, Including Design Limits) CTS 5.4 describes the Reactor Coolant System. The ITS does not contain this information. This changes the CTS by moving the description of the Reactor Coolant System to the UFSAR. The removal of these details, which are related to system design, from the Technical Specifications is acceptable because this type of information is not necessary to be included in the Technical Specifications to provide adequate protection of public health and safety. The ITS still retains requirements on Reactor Coolant System OPERABILITY in ITS Section 3.4. Also, this change is acceptable because the removed information will be adequately controlled in the UFSAR. Any changes to the UFSAR are made under 10 CFR 50.59 or 10 CFR 50.71 (e), which ensures changes are properly evaluated. This change is designated as a less restrictive removal of detail change because information relating to system design is being removed from the Technical Specifications. LA.5 (Type 1- Removing Details of System Design and System Description, Including Design Limits) (Unit 1 only) Unit 1 CTS 5.5 describes the Emergency Core Cooling Systems (ECCS). The ITS does not contain this information. This changes the Unit 1 CTS by moving the description of the ECCS to the UFSAR. The removal of these details, which are related to system design, from the Technical Specifications is acceptable because this type of information is not necessary to be included in the Technical Specifications to provide adequate protection of public health and safety. The ITS still retains requirements on ECCS OPERABILITY in ITS Section 3.5. Also, this change is acceptable because the removed information will be adequately controlled in the UFSAR. Any changes to the UFSAR are made under 10 CFR 50.59 or 10 CFR 50.71(e), which ensures changes are properly evaluated. This change is designated as a less restrictive removal of detail change because information relating to system design is being removed from the Technical Specifications. LA.6 (Type 1 - Removing Details of System Design and System Description, Including Design Limits) (Unit 1 only) Unit 1 CTS 5.7 describes certain general Seismic Classification requirements. The ITS does not contain this information. This changes the Unit 1 CTS by moving the description of these general Seismic Classification requirements to the UFSAR. CNP Units 1 and 2 Page 3 of 4 Attachment 1, Volume 15, Rev. 1, Page 32 of 45
Attachment 1, Volume 15, Rev. 1, Page 33 of 45 DISCUSSION OF CHANGES ITS CHAPTER 4.0, DESIGN FEATURES The removal of these details, which are related to system design, from the Technical Specifications is acceptable because this type of information is not necessary to be included in the Technical Specifications to provide adequate protection of public health and safety. The ITS still retains requirements for various Category I structures, systems, and components. Also, this change is acceptable because the removed information will be adequately controlled in the UFSAR. Any changes to the UFSAR are made under 10 CFR 50.59 or 10 CFR 50.71(e), which ensures changes are properly evaluated. This change is designated as a less restrictive removal of detail change because information relating to system design is being removed from the Technical Specifications. LA.7 (Type 1- Removing Details of System Design and System Description, Including Design Limits) CTS 5.8.1 (Unit 1)and CTS 5.5.1 (Unit 2) describes the location of the meteorological tower. The ITS does not contain this information. This changes the CTS by moving the location of the meteorological tower to the UFSAR. The removal of these details, which are related to system design, from the Technical Specifications is acceptable because this type of information is not necessary to be included in the Technical Specifications to provide adequate protection of public health and safety. 10 CFR 50.36(c)(4) states "Design features to be included are those features of the facility such as materials of construction and geometric arrangements, which, if altered or modified, would have a significant effect on safety and are not covered in categories described in paragraphs (c)(1), (2), and (3)of this section." These paragraphs provide the criteria for safety limits, limiting safety system settings, and limiting control settings; limiting conditions for operation; and surveillance requirements to be included in the Technical Specifications, respectively. The location of the meteorological tower does not meet any of these requirements. Also, this change is acceptable because the removed information will be adequately controlled in the UFSAR. Any changes to the UFSAR are made under 10 CFR 50.59 or 10 CFR 50.71(e), which ensures changes are properly evaluated. This change is designated as a less restrictive removal of detail change because information relating to system design is being removed from the Technical Specifications. LESS RESTRICTIVE CHANGES None CNP Units 1 and 2 Page 4 of 4 Attachment 1, Volume 15, Rev. 1, Page 33 of 45
Attachment 1, Volume 15, Rev. 1, Page 34 of 45 Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs) Attachment 1, Volume 15, Rev. 1, Page 34 of 45
Attachment 1, Volume 15, Rev. 1, Page 35 of 45 CTS Design Features 4.0 4.0 DESIGN FEATURES 4.1 Site Location S,(.I,3 , -!5 (. 3 ( Tefescription of site on.AJ SEP 0D 4.2 Reactor Core 4.2.1 Fuel Assemblies < The reactor-shall contain Orfuel assemblies. Each assembly shall consist of a PD matrix of Zircalloy or ZlRLOqfuel rods with an Initial composition of natural or slightly enriched uranium dioxide (U02 ) as fuel material. Umrited substitutions of 0 zirconium alloy or stainless steel filler rods for fuel rods, in accordance with approved applications of fuel rod configurations, may be used. Fuel assemblies shall be limited to those fuel designs that have been analyzed with applicable NRC staff approved codes and methods and shown by tests or analyses to comply with all fuel safety design bases. A limited number of lead test assemblies that have not completed representative testing may be placed in nonlimiting core regions. 4.2.2 J§ontrol RodgAssemblies Y. 3.2-The reactor core shall contain &6olrodjassemblies.The control material shall beosilver Indium cuadmuedaas approved by the N RC. 0 4.3 Fuel Storage 5.6,1.I 4.3.1 Crdicalitx 4.3.1.1 The spent fuel storage racks are designed and shall be maintained with:
- nr. a. Fuel assemblies having a maximum U-235 enrichment of percer) c2 j
- b. kf &0.95 If fully flooded with unborated water, which includes An )
allowance for uncertainties as described inoection 9QSAF, ( i a c. A norninal9 enter to center distance between fuel s' 6. I.( 1 assemblies placed Inthe Ensue fuel storage rack4&D [d. A nominal j10U.95 inch center to center dNance between fuel) k emblies placed In[low density fuel Stork racks]. 1 09 4S WOG STS 4.0-1 Rev. 2, 04/30/01 Attachment 1, Volume 15, Rev. 1, Page 35 of 45
Attachment 1, Volume "15, Rev. 1, Paged36 of 45 4.0 INSERT 1 4.1.1 Site and Exclusion Area Boundaries The site area and exclusion area boundaries are as shoWn in Figure 4.1-1. 4.1.2 Low Population Zone The low population zone Is all the land within a circle centered on the reactor containment structures and a radius of 2 miles. Insert Page.4.0-1 Attachment 1, Volume 15' Rev. 1, Page 36 of 45
Attachment 1, Volume 15, Rev. 1, Page 37 of 45 CTS Design Features 4.0 4.0 DESIGN FEATURES 4.3 Fuel Storage (continued) S6.-1 c., qew or partially spent fuel assemblies with discharge bum up 7V 3 r a y D allwedunrestricted (3) O storage inei or ra e c ad . (D ' mblies spent fuel assaral ri fzunaceptbiqrrn~e"of Fgure 13.7.1 Jr1 wilE cmZine r.r. C.-2C Nyapproed Spcific documn co0angoe nltcl { withthe ID 4.3.1.2 The new fuel storage racks are de ined and shall be maintained with: 5;' 6. . uel assemblies having maximum 235 enrichment of g percent, CD
- b. k£s 0.95 Iffully flooded with unborated water, which includes a d allowance for uncertainties as described indSection 9.& hePSA*
- c. k," s 0.98 Ifmoderated by aqueous foam, which includes an allowance for uncertainties as described infection 9. of the SA~t and _
5;..t .c d. Anominal ( Ilnch center to center distance between fuel assembliesaced Inthe storage racks. 4.32 Drainaag The spent fuel storage pool is designed and shall be maintained to prevent
*5.6.3L inadvertent draining of the pool below elevation CD 4.3.3 aoarity The spent fuel storage pool Isdesgpnd and shall be maintained with a storage capacity limited to no more than (= fuel assemblies. gjO 0
ATf-j U 0 0b WOG STS 4.0 -2 Rev. 2. 04130/01 Attachment 1, Volume 15, Rev. 1, Page 37 of 45
Attachment 1, Volume 15, Rev. 1, Page 38 of 45 4.0 INSERT 2 Region 1 of Figure 4.3-1 or Figure 4.3-2; Q INSERT 3 Regions 2 and 3 of Figure 4.3-1 or Figure 4.3-2 meeting the initial enrichment and burnup requirements of LCO 3.7.16, "Spent Fuel Pool Storage." Q3 INSERT 4 or a maximum U-235 enrichment within the Acceptable Region of Figure 4.3-3 not to exceed 4.95 weight percent. Linear interpolation of the Boron-1 0 integral fuel burnable absorber (IFBA) loading curves between 1.OX and 1.5X and between 1.5X and 2.OX is acceptable; Insert Page 4.0-2a Attachment 1, Volume 15, Rev. 1, Page 38 of 45
, Volume 15, Rev. 1, Page 39 of 45 4.0
) INSERT 5 Figure 4.1-1 (Page 1 of 1)
Site and Exclusion Area Boundaries Insert Page 4.0-2b , Volume 15, Rev. 1, Page 39 of 45
Attachment 1, Volume 15, Rev. 1, Page 40 of 45 4.0 0 INSERT 6
- 504 REGION I CELLS 0 1439 REGION 2 CELLS 0 1670 REGION 3 CELLS Figure 4.3-1 (Page 1 of 1)
Normal Storage Pattern (Mixed Three Zone) Insert Page 4.0-2c Attachment 1, Volume 15, Rev. 1, Page 40 of 45
Attachment 1, Volume 15, Rev. 1, Page 41 of 45 4.0 0 INSERT 6 (continued) E 158 EMPTY LOCATIONS
- 661 REGION I CELLS a 1439 REGION 2 CELLS E 1355 REGION 3 CELLS Figure 4.3-2 (Page 1 of 1)
Interim Storage Pattern (Checkerboard) Insert Page 4.0-2d Attachment 1, Volume 15, Rev. 1, Page 41 of 45
Attachment 1, Volume 15, Rev. 1, Page 42 of 45 4.0 0 INSERT 7 32 9* .9 9. .9 9. .9 9. 6 -
-['l ii-La Lng 1.0X IFBA Loading (28A, 4.95)
- .5X IFBA Loading 24 IFA oa--g
.0 . . / -
(21.3,4.95) .a. m E Cn I-nL 16 IACCEPTABLE (n 0 (14.2, 4.95) co I_ I_ _ _ __ . _ _/ _ _ L 8
= --/ 7.. -riCCEPTJ3L =
nI 4.50 (0°4.55) 4.60 4.70 235 U tII ____ -1 4.80 _= 4.90 4.95 Enrichment (w/o) Figure 4.3-3 (Page 1of 1) New Fuel Storage Rack Integral Fuel Burnable Absorber (IFBA) Requirements Insert Page 4.0-2e Attachment 1, Volume 15, Rev. 1, Page 42 of 45
Attachment 1, Volume 15, Rev. 1, Page 43 of 45 JUSTIFICATION FOR DEVIATIONS ITS CHAPTER 4.0, DESIGN FEATURES
- 1. The brackets are removed and the proper plant specific information/value has been provided.
- 2. These punctuation corrections have been made consistent with the Writer's Guide for the Improved Standard Technical Specifications, NEI 01-03, Section 5.1.3.
- 3. Changes are made (additions, deletions, and/or changes) to the ISTS which reflect the plant specific nomenclature, number, reference, system description, analysis, or licensing basis description.
- 4. ISTS 4.3.1.1.d, a bracketed requirement, has not been included in the ITS because low density fuel racks are not used in the CNP spent fuel storage pool. Subsequent Specifications have been renumbered, as appropriate, due to this deletion.
CNP Units 1 and 2 Page 1 of I Attachment 1, Volume 15, Rev. 1, Page 43 of 45
Attachment 1, Volume 15, Rev. 1, Page 44 of 45 Specific No Significant Hazards Considerations (NSHCs) Attachment 1, Volume 15, Rev. 1, Page 44 of 45
Attachment 1, Volume 15, Rev. 1, Page 45 of 45 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS CHAPTER 4.0, DESIGN FEATURES There are no specific NSHC discussions for this Chapter. CNP Units 1 and 2 Page 1 of I Attachment 1, Volume 15, Rev. 1, Page 45 of 45
Attachment 1, Volume 16, Rev. 1, Page i of i
SUMMARY
OF CHANGES ITS CHAPTER 5.0 I Change Description Affected Pages A self-identified change for ITS 5.5 and 5.6 has been made. CTS Pages 69, 74, 82, 93, 97, 98, 99, Amendments 281 (Unit 1) and 265 (Unit 2) have been incorporated 101,104,108,116,127,131, into the ITS submittal. This CTS change adopted the allowances of 132,133,135,141,142,176, TSTF-359 and deletes CTS 3.4.10.1 Action d, modifies CTS 3.9.12 190,196, 198, and 204 of 256. Action b, CTS 3.11.1 Action b, CTS 3.11.2.1 Action c, CTS 3.11.2.2 Action b, and adds new CTS 6.8.5. This change does not affect the ITS. The change described in the response to Question 200409200946, Page 149 of 256. Question 200409200950, and Question 200409200954 for ITS 5.5.9 Discussion of Changes (DOC) L.3 has been made. This change revises the Frequency for ITS SR 3.7.10.1 and ITS SR 3.7.12.1 to "46 days on a STAGGERED TEST BASIS," and revises the Frequency for ITS SR 3.7.13.1 to "92 days." A self-identified change for ITS 5.5.9 has been made. This change Page 168 of 256. administratively corrects a typographical error. The change described in the response to Question 200405271640 Page 185 of 256. for ITS 5.5.14 has been made. This change revises ITS 5.5 Justification for Deviations (JFD) 17 to clarify why ITS SR 3.0.2 does not apply to the Frequencies of ITS 5.5.14. A self-identified change for ITS 5.6.5.b has been made. This Pages 194, 202, 207, and 217 change revises ITS 5.6.5.b to include the document describing the of 256. analytical methods for the Overtemperature AT and Overpower AT Allowable Value parameter values as required by WCAP-1 4483-A, "Generic Methodology for Expanded Core Operating Limits Report." A self-identified change for ITS 5.6.6 has been made. This change Pages 219 and 221 of 256. revises the reference to ITS LCO 3.3.3 Condition G to Condition H, because of changes made to the ITS 3.3.3 Conditions in the response to Questions 200406041043 and 200406041123 for ITS 3.3.3. Attachment 1, Volume 16, Rev. 1, Page i of i
Attachment 1, Volume 16, Rev. 1, Page 1 of 256 VOLUME 16 CNP UNITS I AND 2 IMPROVED TECHNICAL SPECIFICATIONS CONVERSION ITS CHAPTER 5.0 ADMINISTRATIVE CONTROLS Revision I Attachment 1, Volume 16, Rev. 1, Page 1 of 256
Attachment 1, Volume 16, Rev. 1, Page 2 of 256 LIST OF ATTACHMENTS
- 1. ITS 5.1
- 2. ITS 5.2
- 3. ITS 5.3
- 4. ITS 5.4
- 5. ITS 5.5
- 6. ITS 5.6
- 7. ITS 5.7
- 8. Relocated/Deleted Current Technical Specifications (CTS)
Attachment 1, Volume 16, Rev. 1, Page 2 of 256
- ----- -- , Volume 16, Rev. 1, Page 3 of 256 ATTACHMENT I ITS 5.1, Responsibility , Volume 16, Rev. 1, Page 3 of 256 , Volume 16, Rev. 1, Page 4 of 256 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) , Volume 16, Rev. 1, Page 4 of 256
Attachment 1, Volume 16, Rev. 1, Page 5 of 256 ITS 5.1 ITS 6.0 ADMINILSTRATIVE CONTROLS 5.1 6.1 RESPONSIBILITY 5.1.1 6.1.1 mhe/ant ftanager shall be responsible for overall succession to this responsibility during his absence. 5.12 6.1.2 The~fiftvanagerl(or during his absence from the coi responsible for t control room command function. l Ito I 0.2 ORGANIZATION I I QNSIEM AND OFFSITE ORGANIZATIONS l L!e Onsite and offsite organizations shall be established for unit operation and corporate management, respectively. The onsite and offsite organizations shall include the positions for activities affecting the safety of the nuclear power plant.
- a. Lines of authority, responsibility, and communication shall be established and defined for the highest management level through intermediate levels to and including all operating organization positions. These relationships shall be documented and updated, as appropriate, in the form of organizational charts. These organizational charts will be documented in the UFSAR and updated in accordance with 10 CFR 50.7 1(e).
See ffs]
- b. The Plant Manager shall be responsible for overall unit safe operation and shall have control over those onsite activities necessary for safe operation and maintenance of the plant.
- c. The Senior Vice President - Nuclear Operations shall have corporate responsibility for overall plant nuclear safety and shall take any measures needed to ensure acceptable performance of the staff in operating, maintaining, and providing technical support to the plant to ensure nuclear safety.
- d. The individuals who train the operating staff and those who carry out health physics and quality assurance functions may report to the appropriate onsite manager; however, they shall have sufficient organizational freedom to ensure their independence from operating pressures.
COOK NUCLEAR PLANT.UNIT I Page 641. AMENDMENTW,13464,U6,CM~ 279 Page 1 of 4 Attachment 1, Volume 16, Rev. 1, Page 5 of 256
Attachment 1, Volume 16, Rev. 1, Page 6 of 256 ITS 5.1 ITS INSERT I 5.1.1 The plant manager or his designee shall approve, prior to implementation, each proposed test, experiment, or modification to systems or equipment that affects nuclear safety. Insert Page 6-1 Page 2 of 4 Attachment 1, Volume 16, Rev. 1, Page 6 of 256
Attachment 1, Volume 16, Rev. 1, Page 7 of 256 ITS 5.1 ITS 6.0 ADMINISTRATIVE CONTROLS 5.1 6.1 RESPONSIBILRTY 5-.11 .1 .1 Tca t Mnge shall be responil froealfacility operation and shall delegate in writing the su. ssion tis responsibility during his absence. 5.1.2 6.12 Thehift& $ana (or during his absence from the control room comnplex, a desi ated individual shalt be responsib cfor the control room command function. IA mnanagenent direcv oti fectsgedb the Site _(M.) Vice President shall Kreissued to all station persri~nel on an annual basis,
§1 6.2 ORGA Z~ON II )
ONSTIE AND OFFSrFE ORGANZA~nONSI 6.2.1 Onsite and offsite organizations shall be established for unit operation and corporate management, respectively. The onsite and offsite organizations shall include the positions for activities affecting the safety of the nuclear power plant
- a. Lines of authority, responsibility, and communication shall be established and defined for the highest management level through intermediate levels to and including all operating organization positions.
These relationships shall be documented and updated, as appropriate, in the form of organizational charts. These organizational charts will be documented in the UFSAR and updated in accordance with 10 CFR 50.71(e). SeeUS]
- b. The Plant Manager shall be responsible for overall unit safe operation and shall have control over those onsite activities necessary for safe operation and maintenance of the plant.
- c. The Senior Vice President - Nuclear Operations shall have corporate responsibility for overall plant I nuclear safety and shall take any measures needed to ensure acceptable performance of the staff in operating, maintaining, and providing technical support to the plant to ensure nuclear safety.
- d. The individuals who train the operating staff and those who carry out health physics and quality assurance functions may report to the appropriate onsite manager; however, they shall have sufficient organizational freedom to ensure their independence from operating pressures.
COOK NUCLEAR PLANT.UNrT 2 Page 6-1. AMENDMET53,417,38,47,497, 261 Page 3 of 4 Attachment 1, Volume 16, Rev. 1, Page 7 of 256
Attachment 1, Volume 16, Rev. 1, Page 8 of 256 ITS 5.1 ITS INSERT I 5.1.1 The plant manager or his designee shall approve, prior to implementation, each proposed test, experiment, or modification to systems or equipment that affects nuclear safety. Insert Page 6-1 Page 4 of 4 Attachment 1, Volume 16, Rev. 1, Page 8 of 256
Attachment 1, Volume 16, Rev. 1, Page 9 of 256 DISCUSSION OF CHANGES ITS 5.1, RESPONSIBILITY ADMINISTRATIVE CHANGES A.1 Inthe conversion of the CNP Current Technical Specifications (CTS) to the plant specific Improved Technical Specifications (ITS), certain changes (wording preferences, editorial changes, reformatting, revised numbering, etc.) are made to obtain consistency with NUREG-1431, Rev. 2, "Standard Technical Specifications-Westinghouse Plants" (ISTS). These changes are designated as administrative changes and are acceptable because they do not result in technical changes to the CTS. A.2 CTS 6.1.2 requires a management directive regarding delegation of the control room command function to be signed by the Site Vice President and issued to all station personnel on an annual basis. ITS 5.1.2 does not include this requirement. This changes the CTS by deleting the requirement to issue this management directive annually. The purpose of CTS 6.1.2 is to specify the plant specific means of implementing the requirement to notify employees of the responsibilities of the Shift Manager. This change is acceptable because CTS 6.1.2 and ITS 5.1.2 state who is responsible for the control room command function. This requirement appears to serve only as a reminder to personnel as to who is in charge. No where else in the CTS or the ITS is a management directive required to remind personnel of a Technical Specification requirement. In addition, this requirement is not considered to be one of the more important requirements since it does not directly impact safety. The Technical Specification control room command function requirement is not being changed. This change is designated as administrative because it does not result in technical changes to the CTS. MORE RESTRICTIVE CHANGES M.1 ITS 5.1.1 requires that the plant manager or his designee approve, prior to implementation, each proposed test, experiment, or modification to systems or equipment that affects nuclear safety. The CTS does not include this requirement. This changes the CTS by adding an approved requirement for the plant manager or his designee. The purpose of the ITS 5.1.1 requirement is to provide additional assurance that the plant manager has direct responsibility for overall unit operation. This change is acceptable because having the plant manager or his designee approve actions affecting nuclear safety is consistent with the CTS 6.2.1 .b (ITS 5.2.1.b) requirement that the plant manager shall be responsible for overall unit safe operation and shall have control over those onsite activities necessary for safe operation and maintenance of the plant. This change is designated more restrictive because it adds a requirement for the plant manager or his designee to the CTS. M.2 CTS 6.1.2 allows a designated individual to assume the responsibility for the control room command function when the Shift Manager is absent from the control room complex. ITS 5.1.2 provides the allowance for the designated CNP Units 1 and 2 Page 1 of 2 Attachment 1, Volume 16, Rev. 1, Page 9 of 256
Attachment 1, Volume 16, Rev. 1, Page 10 of 256 DISCUSSION OF CHANGES ITS 5.1, RESPONSIBILITY individual to assume the responsibility for the control room command function, but provides additional requirements for the designated individual. In MODE 1, 2, 3, or 4, ITS 5.1.2 requires the designated individual hold an active Senior Operator license. In MODE 5 or 6, ITS 5.1.2 requires the designated individual hold an active Senior Operator license or Operator license. This changes the CTS by adding qualification requirements for the designated individual that assumes the control room command function. The purpose of the ITS 5.1.2 requirement is to ensure that the control room command function is maintained. This change is acceptable because the additional requirements ensure that the designated individual assuming the control room command functions meets the appropriate qualification requirements. This change is designated as more restrictive because it adds qualification requirements for the designated individual that assumes the control room command function to the CTS. RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES LA.A (Type 3 - Removing Procedural Details for Meeting TS Requirements or Reporting Requirements) CTS 6.1.1 uses the title "Plant Manager" and CTS 6.1.2 uses the title "Shift Manager." ITS 5.1.1 uses the generic title "plant manager" and ITS 5.1.2 uses the generic title "shift manager." This changes the CTS by moving the specific CNP organizational titles to the UFSAR and replacing them with generic titles. The removal of these details, which are related to meeting Technical Specification requirements, from the Technical Specifications is acceptable because this type of information is not necessary to be included in Technical Specifications to provide adequate protection of public health and safety. The allowance to relocate the specific CNP organizational titles is out of the Technical Specifications is consistent with the NRC letter from C. Grimes to the Owners Groups Technical Specification Committee Chairmen, dated November 10, 1994. The various requirements of the plant manager and shift manager are still retained in the ITS. Also, this change is acceptable because the removed information will be adequately controlled in the UFSAR. Any changes to the UFSAR are made under 10 CFR 50.59 or 10 CFR 50.71(e), which ensures changes are properly evaluated. This change is designated as a less restrictive removal of detail change because information related to meeting Technical Specification requirements are being removed from the Technical Specifications. LESS RESTRICTIVE CHANGES None CNP Units I and 2 Page 2 of 2 Attachment 1, Volume 16, Rev. 1, Page 10 of 256
Attachment 1, Volume 16, Rev. 1, Page 11 of 256 Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs) Attachment 1, Volume 16, Rev. 1, Page 11 of 256
Attachment 1, Volume 16, Rev. 1, Page 12 of 256* Responsibility 5.1 5.0 ADMINISTRATIVE CONTROLS 5.1 Responsibility
- 1. Titles mebers'o'f the u~nit stafhall be specified b~s of an overl I Sta ent'referencing an AN5 1 andard acceptable6 the N RC staff Irorn ich the titles were obtain or an alternative titlmay be designated for\
. his position. Generally, pffirst method Is pre ~le; however, the secondI method is adaptable tqftse unit staffs reqig special titles becaus fj
- . unique organization tructures. /
- 2. The ANSI Sta ard shall be the same SI Standard referenced In S tion 5.3, Unit S Qualifications. If alte tive titles are used, all requir ents of thes echnlcal Specifications ply to the position with the al native title apply with the specified . Unit staff titles shall be sp ified in the
- F Safety Analysis Report Quality Assurance Plan. U stafftitles all be maintained and re ed using those procedures roved for modifying/revising the F I Safety Analysis Report orality Assurance 5.1.1 The plant manager shall be responsible for overall unit operation and shall delegate In writing the succession to this responsibility during his absence.
- The plant manager or his designee shall approve, prior to implementation, each
-D&C- 4.proposed test, experimer fr modification to systems or equipment that affect-o nuclear safety. 5.1.2 The hift for the control room command llhallbe functron. Dunng any absence of tht he control roomhile the unit is In MODE 1, 2, 3, or4, an Individual wian active Seniorkqk Operator
- s license shall be designated to assume the control room command function. During any absence of lth from the control robm'iFii the unit Is in ni Cual with anf actlv license or(E3Operator (I (license shall be designated to assume th-f-Control room command function.
WOG STS Rev. 2, 04130101 Attachment 1, Volume 16, Rev. 1, Page 12 of 256
Attachment 1, Volume 16, Rev. 1, Page 13 of 256 JUSTIFICATION FOR DEVIATIONS ITS 5.1, RESPONSIBILITY
- 1. The brackets are removed and the proper plant specific information/value is provided.
- 2. Grammatical error corrected.
- 3. Typographical error corrected. The terms in 10 CFR 55.4 and 10 CFR 50.54(m) are "Senior Operator" and "Operator."
- 4. The term "control room" in ISTS 5.1.2 has been changed to "control room complex" to be consistent with the current licensing basis.
CNP Units 1 and 2 Page 1 of 1 Attachment 1, Volume 16, Rev. 1, Page 13 of 256
Attachment 1, Volume 16, Rev. 1, Page 14 of 256 Specific No Significant Hazards Considerations (NSHCs) Attachment 1, Volume 16, Rev. 1, Page 14 of 256
Attachment 1, Volume 16, Rev. 1, Page 15 of 256 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 5.1, RESPONSIBILITY There are no specific NSHC discussions for this Specification. CNP Units 1 and 2 Page 1 of I Attachment 1, Volume 16, Rev. 1, Page 15 of 256
--- , Volume 16, Rev. 1, Page 16 of 256 ATTACHMENT 2 ITS 5.2, Organization , Volume 16, Rev. 1, Page 16 of 256 , Volume 16, Rev. 1, Page 17 of 256 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) , Volume 16, Rev. 1, Page 17 of 256
Attachment 1, Volume 16, Rev. 1, Page 18 of 256 ITS 5.2 ITS 6.0 ADMIN[STRATIVE CONTROLS 6.1 RESPONSIBILITY 6.1.1 The Plant Manager shall be responsible for overall facility operation and shall delegate in writing the succession to this responsibility during his absence. See ITS 1 5.1 J 6.1.2 The Shift Manager (or during his absence from the control room complex, a designated individual) shall be responsible for the control room command function. A management directive to this effect signed by the I Site Vice President shall be reissued to all station personnel on an annual basis. 5.2 6.2 ORGANIZATION ONQSIll! AND OFFSl ORGANIZATIQONS 5.2.1 6.2.1 Onsite and offsite organizations shall be established for unit operation and corporate management. respectively. The onsite and offsite organizations shall include the positions for activities affecting the
.safetv nf the nuclear nower nlant.
5.2.1.a a. Lines of authority. responsibility, and communication shall be established and defined for the highest management level through intermediate levels to and including all operating organizationI positions. These relationships shall be documented and upted, as aproprate in the fo organizational charts. Tbseor EnD artgw-ilt documented inlc. Iin Wfordance Wth IO CQK50.71(e 5.2.1 .b b. The')ant lanager shall be responsible for overall unit safe operation and shall have control ove L.( \ those onsite activities necessary for safe operation and maintenance of the last. 5.2.1.c c. Tlhe -Senior Vice Prefident - Nucjefr O~rtoSshallhv oprt epniiiyfroealI ~ tL plant nuclear safety and shall take any measures neededt nueacpal efrac ftet staff in operating, maintaining, and providing technical support to the plant to ensure nuclear safety. 5.2.1 .d d. The individuals who train the operating staff and those who carry out health physics and quality
- assurance functions may report to the appropriate onsite manager, however, they shall have sufficient organizational freedom to ensure their independence from operating pressures.
COOK NUCLEAR PLANT.UNIT I Pae 61 AMENDMENT7 m15,434,436,21, 279 Page 1 of 12 Attachment 1, Volume 16, Rev. 1, Page 18 of 256
Attachment 1, Volume 16, Rev. 1, Page 19 of 256 ITS 5.2 ITS O'INSERT I 5.2.1 .a requirements including the plant-specific titles of those personnel fulfilling the responsibilities of the positions delineated in these Technical Specifications Insert Page 6-1 Page 2 of 12 Attachment 1, Volume 16, Rev. 1, Page 19 of 256
Attachment 1, Volume 16, Rev. 1, Page 20 of 256 ITS 5.2 ITS 6.0 ADMINISTRATIVE CONTROLS 6.2 ORGANIZATION (Continued) FACILITY STAFF 5.2.2 6.2.2 The Facility organization shall be subject to the following: la. Eachjpn-uty shift shall be com f at least the minimum s wcompoosition shown inl3
- b. At leas~~ne licensed Operator shall beXithe control room when fuel i;,Kthe reactor.: In addition, rwfeunit is in Mode I, 2, 3 , at least one licensed Senior rator shall be in the control :
5.2.2.c c. An individual qualified in radiation protection procedures shall be on site when fuel is in the reactor.
- d. All COW ALTERATIONS shall be Wetly supervised by a licensed enior Operator trained or qual ed in refueling and CORE LTERATIONS (SO-CA) wa* has no other concurrent A.3 nsibilities during this operaX /.
5.2.2.d e. The amount Of overtime worked by plant staff members performning safety-related functionssmustt be limited in accordance with NRC Policy Statement on working hours (Generic Letter 82-12).=t
- f. _h§hift Manager and Upit-9upervisor shall hold aSd~iorOperator icoerse.o man ger 5.2.2.e g. Thef oens tor must hold or have held a Senior Operator License at Cook Nuclear Plant I or a similar reactor, or have been certified for equivalent senior operator knowledge. If the 10pera-nsj 3 does not hold a Senior Operator License, then a line rations A.1 middle manage all hold a Senior Operator License for the purposes of directing operational activities.
Ioperaiuons m;a3ag: 5.2.2.c The unexpected absence, for a period of time not to exceed 2 hours, of the on-site individual qualified in radiation protection procedures is permitted provided immediate action is taken to fill the required position. COOK NUCLEAR PLANT-UNIT I Page 6-2 AMENDMENT 7, 42, 454, GUW 279 Page 3 of 12 Attachment 1, Volume 16, Rev. 1, Page 20 of 256
Attachment 1, Volume 16, Rev. 1, Page 21 of 256 ITS 5.2 ITS 0 If 6.0 ADMINISTRATIVE CONTROLS TABIE 6.2-1 MINIMUM SHIFT CREW COMPOSITION 5.2.2.a 5.2.2.f l D not include the licensed oerrtor- CA su ri4us ORE ALTERATIONS. l 5.2.2.b
- Shift crew composition may be less than the minimum requirements for a period of time not to exceed 2 hours in order to accommodate unexpected absence of on duty shift crew members provided immediate action is taken to restore the shift crew composition to within the minimum requirements of Table 6.2-1.
5.2.2.1 ** Shared with Cook Nuclear Plant Unit 2. COOK NUCLEAR PLANT-UNIT I Page 6-3 AMENDMENT63, 80,42, 464, 279 Page 4 of 12 Attachment 1, Volume 16, Rev. 1, Page 21 of 256
Attachment 1, Volume 16, Rev. 1, Page 22 of 256 ITS 5.2 ITS O INSERT2 5.2.2.f An individual shall provide advisory technical support to unit operations shift crew in the areas of thermal hydraulics, reactor engineering, and plant analysis with regard to safe operation of the unit. Insert Page 6-3 Page 5 of 12 Attachment 1, Volume 16, Rev. 1, Page 22 of 256
Attachment 1, Volume 16, Rev. 1, Page 23 of 256 ITS 5.2 ITS 0 6.0 ADMINISTRATIVE CONTROIS 6.3 FACILITY STAFF OUALIFICTIONSI,
. See IrrS 6.3.1 Each member ofthe facility staffshil meetor exceed the minimum qualificationstofANSI N18.1-1971 fod 5 ra inin g Mat cofmthea n ager and hat meet or( exceed th e i re met s an dr emMan requi aaer wh n t ohfn Sncnitr ofI ex J aiTechnical aho Advsor, shall hav 5.2.2.f lsn n~ls;<of rrpone not he lsn fo tnrieh sfl r~dnlrnrf-I3) the OperationsDietr h _
Irnut b ouaifid a sreifid inSecion6.2..X.1 (See ITS 6,.4 TRAINING I . 6.4.1 A retraining and replacement training prograrn for the facility staff shall be maintained under the direction See CTS of the Training Manager and shall meet or exceed the requirements and recommendations of Section 5.5 of 6.0 J ANSI NJ18.1 -1971 and 10 CFR Part S5. 6.5 DELETED COOK NUCLEAR PLANT.UNIT I Page 6.4 AMENDMENT49,63,4,454,4U, 2,22,243, 279 Page 6 of 12 Attachment 1, Volume 16, Rev. 1, Page 23 of 256
Attachment 1, Volume 16, Rev. 1, Page 24 of 256 ITS 5.2 ITS 6.0 ADMINISTRATIVE CONTROIS 6.1 RESPONSIBILITY 6.1.1 The Plant Manager shall be responsible for overall facility operation and shall delegate in writing the succession to this responsibility during his absence. See ITS] 1 5.1 ) 6.1.2 The Shift Manager (or during his absence from the control room complex, a designated individual) shall be responsible for the control room command function. A management directive to this effect signed by the Site Vice President shall be reissued to all station personnel on an annual basis. I 5.2 62 ORGANIZATlON ONSITE AND OFFSITE ORGANIZATIONS 5.2.1 6.2.1 Onsite and offsite organizations shall be established for unit operation and corporate management, respectively. The onsite and offsite organizations shall include the positions for activities affecting the safety of the nuclear r r--.n 5.2.1.a a. lines of authority, responsibility, and communication shall be established and defined for the highest management level through intermediate levels to and including all operading organization positions. [INSER1 These rehdonships shall be documlte and updated, as appropriate, in the form of orgnztoa chrs ieem ju-nkeat ]*l be doc.2te inte)Sad pac-5.2.1.b b. mhfantoanager shall be responsible for overall unit safe operation and shall have control over LA.) those onsite activities necessary for safe operation and maintenance of the plant. Aspecfier corporate officer 5.2.1.c c. he Sior Vice Prdent - Nuetif(perafin WZal hae corporate responsibility for ovcrall plant I nuclear safety and shall take any measures needed to ensure acceptable perfornance of the staff in operating, maintaining, and providing technical support to the plant to ensure nuclear safety. 5.2.1 .d d. The individuals wio train the operating staff and those who carry out health physics and quality assurance functions may report to the appropriate onsite manager; however, they shall have sufficient organizational freedom to ensure their independence from operating pressures. COOK NUCLEAR PLANT-UNIT 2 Page 6-1. AMENDMENT 58,44,U8,2497,261 Page 7 of 12 Attachment 1, Volume 16, Rev. 1, Page 24 of 256
Attachment 1, Volume 16, Rev. 1, Page 25 of 256 ITS 5.2 ITS OINSERT I 5.2.1.a requirements including the plant-specific titles of those personnel fulfilling the responsibilities of the positions delineated in these Technical Specifications Insert Page 6-1 Page 8 of 12 Attachment 1, Volume 16, Rev. 1, Page 25 of 256
Attachment 1, Volume 16, Rev. 1, Page 26 of 256 ITS 5.2 ITS 6.0 ADM1NISTRATIVE CONTROLS 6.2 ORGANIZATION (Continued) FACILUTYSAE 5.22 6.2.2 The Facility organization shall be subject to the following: la. Eactrfut shift shall be conpos te asd temnumsi po d nshwn in Table )
- b. At Icasdenc licensed Operator shall bcXthe control room when fuel isAitthe reactor. In~'additdon.
whel~i unit is in Mode 1, 2, 3, at least one licensed Seniotrator shall be in the cnrlt 5.2.2.c C. An individual qualified in radiation protection procedures shall be on site wihen fuel is in the reactor.
- d. All CO ALTERATIONS shall be 4Kiy supervised by a licensed Benior Operator trained or quar ed in refueling and CORE LTERATIONS (SO-CA) w has no other concurrent A.3 risibilities during this operati/
5.2.2.d e. The amount of overtime worked by plant staff members performing safety-rclated functions must be limited in accordance with NRC Policy Statement on working hours (Generic Letter 82-12).
- l. L hethift Manager and Uqitervisor shall hold a S;ai tOperator Ucense. l 5.2.2.e g. The eons hold or have held a Senior Operator Ucense at Cook Nuclear Plant or a similar reactor, or have been certified for equivalent senior operator knowledge If the O tions nothlaSeirO rao License, then a line [ Imaagr sall hold a Senior Operator License for the purposes ofdirecting operational activities.
5.2.2.c The unexpected absence, for a period of time not to exceed 2 hours, of the on-site individual qualified in radiation protection procedures is permitted provided immediate action is taken to fill the required position. COOK NUCLEAR PLANT-UNIT2 Page 6-2, AMENDMENNTJ,w,US,4,4;, 261 Page 9 of 12 Attachment 1, Volume 16, Rev. 1, Page 26 of 256
Attachment 1, Volume 16, Rev. 1, Page 27 of 256 ITS 5.2 ITS 6.0 ADMINISTRATIVE CONTROLS TABLE 6.2-1 MINIMUM SHIFT CREW COMPOSITION* 5.2.2.a 5.2.2.1 I# Dwxrit
.adOperator include the licensed -CA supervisjpCRE ALTERATIONS. _I 5.2.2.b
- Shift crew composition may be less than the minimum requirements for a period of time not to exceed 2 hours in order to accommodate unexpected absence of on duty shift crew members provided immediate action is taken to restore the shift crew composition to within the minimum requirements ofTable 6.2-I.
5.2.2.f et Shared with Cook Nuclear Plant Unit I COOK NUCLEAR PLANT.UNIT 2 Page 6.3. AMENDMENT 6, A, 43s, 261 Page 10 of 12 Attachment 1, Volume 16, Rev. 1, Page 27 of 256
Attachment 1, Volume 16, Rev. 1, Page 28 of 256 ITS 5.2 ITS O@ INSERT 2 5.2.2.f An individual shall provide advisory technical support to unit operations shift crew in the areas of thermal hydraulics, reactor engineering, and plant analysis with regard to safe operation of the unit. Insert Page 6-3 Page 11 of 12 Attachment 1, Volume 16, Rev. 1, Page 28 of 256
Attachment 1, Volume 16, Rev. 1, Page 29 of 256 ITS 5.2 ITS 6.0 ADMINISTRATIVE CONTROLS 6.3 FACILITY STAFFOUALFICTINS l Se US 6.3.1 Each mcmber of the facility staff shall rneet or exceed the minimum qualifications of ANSI N18.1-1971 f r 5.3 J comparablk positions. except for (1) thc Plant RadiationProtction Manager. who shall rneet or e e aualifications of Regulator= Guide I A. Septeml er 1 5 - 2le Shift Tcchnical Advisor, who shall havalA.) 5.2.2.1f bachelor's degreorcequivakent in ascientific or engineering discipline with ncifictrainingilj t design Iand rcspons a~nd analysis of thcplant for transientsatnd accidents and.n3) the Operations Dietiehr must be qualified as specified in Seto ... lSee ITS 6.4 TRAINING I . 6A.I A retraining and replacement training progratn for the facility staff shall be maintained under the direction (See CTS) of the Training Manager and shall rneet or execeed the requirernents and recomnmendations of Section 5.5 of 6.0 ANSI Nt Rt. 971 anld I n rFR Part -5s 6.5 DELETED COOK NUCLEAR PLANT-UNIT 2 Page64 AMENDMENT 34, 44, 48, 4- ,48, 4WMO,94, 261 Page 12 of 12 Attachment 1, Volume 16, Rev. 1, Page 29 of 256
Attachment 1, Volume 16, Rev. 1, Page 30 of 256 DISCUSSION OF CHANGES ITS 5.2, ORGANIZATION ADMINISTRATIVE CHANGES A.1 In the conversion of the CNP Current Technical Specifications (CTS) to the plant specific Improved Technical Specifications (ITS), certain changes (wording preferences, editorial changes, reformatting, revised numbering, etc.) are made to obtain consistency with NUREG-1431, Rev. 2, "Standard Technical Specifications-Westinghouse Plants" (ISTS). These changes are designated as administrative changes and are acceptable because they do not result in technical changes to the CTS. A.2 CTS 6.2.1.a states, in part, "These organizational charts will be documented in the UFSAR and updated in accordance with 10 CFR 50.71(e)." The ITS does not include the requirement associated with updating the UFSAR in accordance with 10 CFR 50.71 (e). This changes the CTS by deleting these requirements for updating the UFSAR. 10 CFR 50.71(e) provides requirements for periodically updating the UFSAR. This change is acceptable because the requirements deleted from the Technical Specifications are already required by 10 CFR 50.71(e). This change is designated as administrative because it does not result in technical changes to the CTS. A.3 CTS 6.2.2.b states "At least one licensed Operator shall be in the control room when fuel is in the reactor. In addition, while the unit is in Mode 1, 2, 3, or 4, at least one licensed Senior Operator shall be in the control room." CTS 6.2.2.d requires all CORE ALTERATIONS to be directly supervised by a licensed Senior Operator trained or qualified in refueling and CORE ALTERATIONS who has no other concurrent responsibilities during this operation. The ITS does not include these requirements. This changes the CTS by deleting these requirements. 10 CFR 50.54(m)(2)(iii) states 'When a nuclear power unit is in an operational mode other than cold shutdown or refueling, as defined by a unit's technical specifications, each licensee shall have a person holding a senior operator license for the nuclear power unit in the control room at all times. In addition to this senior operator, for each fueled nuclear power unit, a licensed operator or senior operator shall be at the controls at all times." 10 CFR 50.54(m)(2)(iv) states "Each licensee shall have present, during alteration of the core of a nuclear power unit (including fuel loading or transfer), a person holding a senior operator license or a senior operator license limited to fuel handling to directly supervise the activity and, during this time, the licensee shall not assign other duties to this person." This change is acceptable because the requirements deleted from the Technical Specifications are already required by 10 CFR 50.54(m)(2)(iii) and 10 CFR 50.54(m)(2)(iv). This change is designated as administrative because it does not result in technical changes to the CTS. A.4 CTS 6.3.1 provides, in part, qualification requirements for the Shift Technical Advisor (STA), and requires the STA to have a bachelor's degree or equivalent in a scientific or engineering discipline with specific training in plant design, and response and analysis of the plant for transients and accidents. ITS 5.2.2.f requires this individual to meet the qualification requirements of the Commission CNP Units 1 and 2 Page 1 of 4 Attachment 1, Volume 16, Rev. 1, Page 30 of 256
Attachment 1, Volume 16, Rev. 1, Page 31 of 256 DISCUSSION OF CHANGES ITS 5.2, ORGANIZATION Policy Statement on Engineering Expertise on Shift. This changes the CTS by referencing the Commission Policy Statement on Engineering Expertise on Shift for qualification requirements instead of listing the specific qualification requirements. The purpose of the CTS 6.3.1 STA requirements is to specify the minimum qualification requirements for the STA. This change is acceptable because the qualification requirements included in the Commission Policy Statement on Engineering Expertise on Shift encompass the current STA qualification requirements. This change is designated as administrative because it does not result in technical changes to the CTS. MORE RESTRICTIVE CHANGES M.1 CTS 6.2.1.a, regarding documentation and updating of the relationships between operating organization positions, requires the organizational charts to be documented in the UFSAR. ITS 5.2.1.a states "These requirements, including the plant-specific titles of those personnel fulfilling the responsibilities of the positions delineated in these Technical Specifications, shall be documented in the UFSAR." This changes the CTS by requiring that the specific CNP organizational titles be specified in the UFSAR. This change is acceptable because specifying the relationship of the specific CNP organizational titles to the generic titles used in the Technical Specifications and industry standards in the UFSAR continues to ensure that organizational positions and associated responsibilities will be maintained. This change adds this requirement to the Technical Specifications. This change is designated as more restrictive because it requires additional information be maintained in the UFSAR. M.2 CTS Table 6.2-1 requires the minimum shift crew to include one STA (shared between Units 1 and 2) when the unit is in MODE 1, 2, 3, or 4. ITS 5.2.2.f requires, in part, that an individual (shared between Units 1 and 2) provide advisory technical support to the unit operations shift crew in the areas of thermal hydraulics, reactor engineering, and plant analysis with regard to the safe operation of the unit, when the unit is in MODE 1, 2, 3, or 4. This changes the CTS by detailing the specific responsibilities of the STA. The purpose of the CTS Table 6.2-1 STA requirements is to ensure that appropriate engineering expertise is available on shift. This change is acceptable because it clarifies STA requirements consistent with Commission Policy Statement on Engineering Expertise on Shift. This change is designated as more restrictive because it provides specific details of the responsibilities of the STA. RELOCATED SPECIFICATIONS None CNP Units 1 and 2 Page 2 of 4 Attachment 1, Volume 16, Rev. 1, Page 31 of 256
Attachment 1, Volume 16, Rev. 1, Page 32 of 256 DISCUSSION OF CHANGES ITS 5.2, ORGANIZATION REMOVED DETAIL CHANGES LA.1 (Type 3 - Removing Procedural Details for Meeting TS Requirements or Reporting Requirements) CTS 6.2.1.b uses the title "Plant Manager," CTS 6.2.1 .c uses the title "Senior Vice President - Nuclear Operations," and CTS 6.2.2.g uses the title "Operations Director." ITS 5.2.1.b uses the generic title "plant manager," ITS 5.2.1.c uses the generic title "A specified corporate officer," and ITS 5.2.2.e uses the generic title "operations manager." This changes the CTS by moving the specific CNP organizational titles to the UFSAR and replacing them with generic titles. The removal of these details, which are related to meeting Technical Specification requirements, from the Technical Specifications is acceptable because this type of information is not necessary to be included in Technical Specifications to provide adequate protection of public health and safety. The allowance to relocate the specific CNP organizational titles out of the Technical Specifications is consistent with the NRC letter from C. Grimes to the Owners Groups Technical Specification Committee Chairmen, dated November 10, 1994. The various requirements of the plant manager, the specified corporate officer, and the operations manager are still retained in the ITS. Also, this change is acceptable because the removed information will be adequately controlled in the UFSAR. Any changes to the UFSAR are made under 10 CFR 50.59 or 10 CFR 50.71(e), which ensures changes are properly evaluated. This change is designated as a less restrictive removal of detail change because information related to meeting Technical Specification requirements are being removed from the Technical Specifications. LA.2 (Type 3 - Removing Procedural Details for Meeting TS Requirements or Reporting Requirements) CTS 6.2.2 and Table 6.2-1, including footnote #, provide minimum shift crew composition requirements. ITS 5.2.2 only includes the minimum shift crew composition requirements that are not already included in 10 CFR 50.54. This changes the CTS by moving the minimum shift crew composition requirements addressed by 10 CFR 50.54 to the Technical Requirements Manual (TRM). The removal of these details, which are related to meeting Technical Specification requirements, from the Technical Specifications is acceptable because this type of information is not necessary to be included in the Technical Specifications to provide adequate protection of public health and safety. The minimum shift crew composition requirements for licensed operators and senior operators are also contained in 10 CFR 50.54(k), (I), and (m)and do not need to be repeated in the Technical Specifications. The minimum shift crew composition requirements for non-licensed operators are transferred from CTS Table 6.2-1 to ITS 5.2.2.a and the minimum shift crew composition requirements for the STA are transferred from CTS Table 6.2-1 to ITS 5.2.2.f. The relocation of the details of the minimum shift crew composition requirements to the TRM is acceptable considering the controls provided by regulations and the remaining requirements in the Technical Specifications. Also, this change is acceptable because these details will be adequately controlled in the TRM. Any changes to the TRM are made under 10 CFR 50.59, which ensures changes are properly evaluated. This change is designated as a less restrictive removal of detail CNP Units 1 and 2 Page 3 of 4 Attachment 1, Volume 16, Rev. 1, Page 32 of 256
Attachment 1, Volume 16, Rev. 1, Page 33 of 256 DISCUSSION OF CHANGES ITS 5.2, ORGANIZATION change because details for meeting Technical Specification and regulatory requirements are being removed from the Technical Specifications. LA.3 (Type 3 - Removing Procedural Details for Meeting TS Requirements or Reporting Requirements) CTS 6.2.2.f requires the Shift Manager and Unit Supervisor to hold a Senior Operator license. ITS 5.2.2 does not contain this requirement. This changes the CTS by moving the requirement for the Shift Manager and Unit Supervisor to hold a Senior Operator license to the TRM. The removal of these details, which are related to meeting Technical Specification requirements, from the Technical Specifications is acceptable because this type of information is not necessary to be included in the Technical Specifications to provide adequate protection of public health and safety. The requirement for shift supervision to hold Senior Operator licenses is contained in 10 CFR 50.54(m), and does not need to be repeated in the Technical Specifications. The relocation of the details of the shift supervision personnel that are required to hold Senior Operator licenses to the TRM is acceptable considering the controls provided by regulations. Also, this change is acceptable because these details will be adequately controlled in the TRM. Any changes to the TRM are made under 10 CFR 50.59, which ensures changes are properly evaluated. This change is designated as a less restrictive removal of detail change because details for meeting Technical Specification and regulatory requirements are being removed from the Technical Specifications. LESS RESTRICTIVE CHANGES None CNP Units 1 and 2 Page 4 of 4 Attachment 1, Volume 16, Rev. 1, Page 33 of 256
Attachment 1, Volume 16, Rev. 1, Page 34 of 256 Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs) Attachment 1, Volume 16, Rev. 1, Page 34 of 256
Attachment 1, Volume 16, Rev. 1, Page 35 of 256 Organization 5.2 5.0 ADMINISTRATIVE CONTROLS 5.2 Organization
- 10. 2 . 1 5.2.1 Onsite and Offsite Orcanizations Onsite and offsite organizations shall be established for unit operation and corporate management, respectively. The onsite and offsite organizations shall indude the positions for activities affecting safety of the nuclear power plant.
- a. Lines of authority, responsibility, and communication shall be defined and established throughout highest management levels, Intermediate levels, end all operating organization positions. These relationships shall be documented and updated, as appropriate, in organization chartsy functfiorian i um1These requiremen ncluding the plant-specific titles of those personnel fulfilling the respon ilties of the positions delineated In these Technical Specificatior hall be documented in the ,SA an R 6 I b. The plant manager shall be responsible for overall safe operation of the plant and shall have control over those onsite activities necessary for safe operation and maintenance of the plant&_,
- c. A specified corporate officer shall have corporate responsibility for overall C.C plant nuclear safety and shall take any measures needed to ensure acceptable performance of the staff in operating, maintaining, and providing technical support to the plant to ensure nuclear safet y (a)
,, 7.. ( . c d. The individuals who train the operating staff, carry out health physics, or perform quality assurance functions may report to the appropriate onsite manager; however, these Individuals shall have sufficient organizational freedom to ensure their independence from operating pressures.
5.2.2 Unit Staff The unit staff organization shall Include the following: 1b'. te 4 .2 - l a. A non-licensed operator shall be assigned to each reactor containing fuel and an additional non-licensed operator shall be assigned for each control room from which a reactor Is operating in MODEa@1, 2, 3, or C, Twounit - REVI ERS NOTE- t h Two unit si s with bothunits shutdown or defueled requie "a total of three non-/( licensed; erators for the two units. WOG STS 5.2 - 1 Rev. 2, 04/30101 Attachment 1, Volume 16, Rev. 1, Page 35 of 256
Attachment 1, Volume 16, Rev. 1, Page 36 of 256 Organization C-vs 5.2 5.2 Organization 5.2.2 Unit Staff (continued) j
- b. Shift crew composition maybess than the minimum requirement of N.L~4 4.2.1 10 CFR 50.54(m)(2)(i) a 2.2.a and 5.2.2.f for a period of time not to exceed 2 hours In order to accommodate unexpected absence of on-duty shift crew members provided Immediate action Is takel to restore the shift crew composition to within the minimum requirementv
- c. A radiation protection technician shall be on site when fuel Is In the reactor.
The position may be vacant for not more than 2 hours, Inorder to provide for unexpected absence, provided immediate action Is taken to fill the ( 0.2.Z. Ce required positlor,
- d. mriai-ie-vprrdeipu-s 1i o aid laniemeinted to limit deR working ours of personrz who perform safety ret ed functions (e.g., ensed Senior Ractor Operators (SROsY,licensed Reactor Ope tors (ROs), he tenance persoe).
physicists, auxiliary o rators, and key
/-e -D he controls s I Include guidelines on rklng hours that ensure, adequate sh, coverage shall be maint ned without routine hea use of overtime.
Any d ation from the above gu elines shall be authoriz advance by the nt manager or the plant anager's designee, in ordance with ap oved administrative pro dures, and with documn tation of the basis r granting the deviation. outine deviation from t working hour guidelines shall not be a orized. Controls shall be In ded in the procedures to equire a periodic / independent revie be conducted to ensure at excessive hours have not e asiged 9
- e. Th operations ma sstant opeZns manager shall h. d
.al pro avisory echnical support to the unit operations T.Ak .2 -11 shift crew In the areas of thermal hydraulics, reactor engineering, and plant 3 (iD
~. p 4 j analysis with regard to the safe operation of the unit. This Individual shall meet the qualifications specified by the Commission Policy Statement on (O.S. I Engineering Expertise on Shift.
/1 WOG STS 5.2-2 Rev. 2, 04130/01 Attachment 1, Volume 16, Rev. 1, Page 36 of 256
Attachment 1, Volume 16, Rev. 1, Page 37 of 256 5.2 0 INSERT I The amount of overtime worked by unit staff members performing safety related functions must be limited in accordance with the NRC Policy Statement on working hours (Generic Letter 82-12); (Q) INSERT 2 must hold or have held a Senior Operator license at Cook Nuclear Plant ora similar reactor, or have been certified for equivalent Senior Operator knowledge. If the operations manager does not hold an Senior Operator license, then a line operations middle manager shall hold a Senior Operator license for the purposes of directing operational activities. (3 INSERT 3 In MODE 1, 2, 3, or 4, an individual (shared with Unit 2 (Unit 1) and Unit 1 (Unit 2)) Insert Page 5.2-2 Attachment 1, Volume 16, Rev. 1, Page 37 of 256
Attachment 1, Volume 16, Rev. 1, Page 38 of 256 JUSTIFICATION FOR DEVIATIONS ITS 5.2, ORGANIZATION
- 1. ISTS 5.2.1.a is revised to reflect the CNP CTS with respect to documentation and updating of the relationships between operating organization positions. Specifically, the ISTS 5.2.1.a requirement for including these relationships in functional descriptions of departmental responsibilities and relationships, and job descriptions of key personnel positions, or in equivalent forms of documentation is not included in ITS 5.2.1.a. This change is made to achieve consistency with CTS 6.2.1.a, which was approved by the NRC in License Amendments 132 (Unit 1)and 117 (Unit 2),
dated March 9, 1990.
- 2. The brackets are removed and the proper plant specific information/value is provided.
- 3. Grammatical/typographical error corrected.
- 4. The ISTS Reviewer's Note has been deleted since it is not intended to be included in the ITS. The requirements for non-licensed operators for two unit sites addressed in the ISTS Reviewer's Note are not adopted. This change is consistent with the CNP CTS.
- 5. ISTS 5.2.2.d provides requirements for working hour limitations. These requirements are revised in ITS 5.2.2.d to reflect the CNP CTS 6.2.2.e requirements, which were approved by the NRC in License Amendments 77 (Unit 1)and 58 (Unit 2), dated November 23, 1983.
- 6. ISTS 5.2.2.e provides a requirement for the operations manager or the assistant operations manager to hold a Senior Operator license. This requirement is revised in ITS 5.2.2.e to reflect the CNP CTS 6.2.2.g requirements. The CTS 6.2.2.g requirements were approved by the NRC in License Amendments 212 (Unit 1)and 197 (Unit 2), dated November 13, 1996.
- 7. ISTS 5.2.2.f provides requirements for the Shift Technical Advisor (STA). These requirements are revised in ITS 5.2.2.f to reflect the CNP CTS Table 6.2-1 requirements for the STA. The CTS Table 6.2-1 STA requirements were approved by the NRC in License Amendments 49 (Unit 1)and 34 (Unit 2), dated August 25, 1981.
- 8. The referenced requirements are Specifications, not Code of Federal Regulations (CFR) requirements. Therefore, the word "Specifications" has been added to clearly state that 5.5.2.a and 5.5.2f are Specifications.
- 9. The punctuation corrections have been made consistent with the Writer's Guide for the Improved Standard Technical Specifications, NEI 01-03, Section 5.1.3.
CNP Units I and 2 Page 1 of 1 Attachment 1, Volume 16, Rev. 1, Page 38 of 256
Attachment 1, Volume 16, Rev. 1, Page 39 of 256 Specific No Significant Hazards Considerations (NSHCs) Attachment 1, Volume 16, Rev. 1, Page 39 of 256
Attachment 1, Volume 16, Rev. 1, Page 40 of 256 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 5.2, ORGANIZATION There are no specific NSHC discussions for this Specification. CNP Units I and 2 Page 1 of 1 Attachment 1, Volume 16, Rev. 1, Page 40 of 256
, Volume 16, Rev. 1, Page 41 of 256 ATTACHMENT 3 ITS 5.3, Unit Staff Qualifications , Volume 16, Rev. 1, Page 41 of 256 , Volume 16, Rev. 1, Page 42 of 256 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) , Volume 16, Rev. 1, Page 42 of 256
Attachment 1, Volume 16, Rev. 1, Page 43 of 256 ITS 5.3 ITS 6.0 ADMINISTRATIVE CONTROLS 5.3 5.3.1 6.5 DELETED COOK NUCLEAR PLANT-UNrr I Page 64 AMENDMENT 49,63,433,454,415, 492,226,243. 279 Page 1 of 2 Attachment 1, Volume 16, Rev. 1, Page 43 of 256
Attachment 1, Volume 16, Rev. 1, Page 44 of 256 ITS 5.3 ITS 6.0 ADMINISTRATIVE. CONTROLS 5.3 6.3 FACILITY STAFF OUALIFICATIONS p LA.1 5.3.1 6.3.1 Each member of the facility staff shall meet or exceed the minimum qualifications of ANSI N18.1-1971 for comparable positions, except for (1) the R adiation 2otection anager, who shall meet or exceed See RS qualifications of Regulatory Guide 1.8, September 1975.1(2) the Shift Technical Advisor, who shall havea5.2 lbachelor's degreeor equivalent in ascie-nuic orenginenng disciplinecwith specific bainin in plant design _ land response and analysis of the plant for transients and accidents and 3) the1 iperaton nc who must be qualified as specified in Section 6.2.2.g. ( ana 6.4 TRAINGN edpoecipedationi5n3D3.2_ SAdr 6.4.1 A retraining and replacement training program for the facility staff shall be maintained under the direction of the Training Manager and shall meet or exceed the requirements and recommendations of Section 5.5 of See CTS ANSI NI8.1-1971 and 10 CFR Part 55. 6.0 6.5 DELIET6 COOK NUCLEAR PLANT-UNIT 2 Page 64 AMENDMENT 4,44,,498,&4;478, 49q,M,234, 261 Page 2 of 2 Attachment 1, Volume 16, Rev. 1, Page 44 of 256
Attachment 1, Volume 16, Rev. 1, Page 45 of 256 DISCUSSION OF CHANGES ITS 5.3, UNIT STAFF QUALIFICATIONS ADMINISTRATIVE CHANGES A.1 Inthe conversion of the CNP Current Technical Specifications (CTS) to the plant specific Improved Technical Specifications (ITS), certain changes (wording preferences, editorial changes, reformatting, revised numbering, etc.) are made to obtain consistency with NUREG-1431, Rev. 2, "Standard Technical Specifications-Westinghouse Plants" (ISTS). These changes are designated as administrative changes and are acceptable because they do not result in technical changes to the CTS. A.2 ITS 5.3.2 states "For the purpose of 10 CFR 55.4, a licensed Senior Operator and a licensed Operator are those individuals who, in addition to meeting the requirements of Specification 5.3.1, perform the functions described in 10 CFR 50.54(m)." The CTS does not include such a statement. This changes the CTS by clarifying that these individuals must meet all of the qualification requirements referenced in 10 CFR 55.4, ITS 5.3.1, and 10 CFR 50.54(m). This change is acceptable because it clarifies the existing relationship between the Technical Specifications and regulations regarding licensed Senior Operator and Operator qualification requirements. This change is designated as administrative because it does not result in technical changes to the CTS. MORE RESTRICTIVE CHANGES None RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES LA.1 (Type 3 - Removing Procedural Details for Meeting TS Requirements or Reporting Requirements) CTS 6.3.1 uses the titles "Plant Radiation Protection Manager" and "Operations Director." ITS 5.3.1 uses the generic titles "radiation protection manager" and "operations manager." This changes the CTS by moving the specific CNP organizational titles to the UFSAR and replacing them with generic titles. The removal of these details, which are related to meeting Technical Specification requirements, from the Technical Specifications is acceptable because this type of information is not necessary to be included in Technical Specifications to provide adequate protection of public health and safety. The allowance to relocate the specific CNP organizational titles out of the Technical Specifications is consistent with the NRC letter from C. Grimes to the Owners Groups Technical Specification Committee Chairmen, dated November 10, 1994. The various requirements of the radiation protection manager and the operations CNP Units I and 2 Page 1 of 2 Attachment 1, Volume 16, Rev. 1, Page 45 of 256
Attachment 1, Volume 16, Rev. 1, Page 46 of 256 DISCUSSION OF CHANGES ITS 5.3, UNIT STAFF QUALIFICATIONS manager are still retained in the ITS. Also, this change is acceptable because the removed information will be adequately controlled in the UFSAR. Any changes to the UFSAR are made under 10 CFR 50.59 or 10 CFR 50.71(e), which ensures changes are properly evaluated. This change is designated as a less restrictive removal of detail change because information related to meeting Technical Specification requirements are being removed from the Technical Specifications. LESS RESTRICTIVE CHANGES None CNP Units 1 and 2 Page 2 of 2 Attachment 1, Volume 16, Rev. 1, Page 46 of 256
Attachment 1, Volume 16, Rev. 1, Page 47 of 256 Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs) Attachment 1, Volume 16, Rev. 1, Page 47 of 256
Attachment 1, Volume 16, Rev. 1, Page 48 of 256 Qr . H . 53 Unit Staff Qualifications
- 5.0 ADMINISTRATIVECONNTROLS'
.5.3 Unit Staff Qualifications
* . : ' /; .. . / . . : ~'
.- REVIP(tER'S NOTE-. .'/':
Mini m qualifications for memb of the unit staff shall be ecified by use of a: -averali qualification statemreferencing an ANSI St ard acceptable to
- '. . . ~' fheNRC'staff or~by seflpndvdual position 4ualigl ons.; 'Genrly 6.-. h
- ,frtmethod Is prefrbgowv, the second metdI adaptable to thos~nt
., ,,..staffs requirlng spculfato statements §duse unique organj oal .
.. , structures.,
- 3. ' 5.3.1 . Each member of the unit staff shall meet or exceed the mlnimum qualifications of
- : ' i,
- Stpsrd ccptablpfo the NR C sta fte- staff ntcrdb egioS' 2 ':
ide 1.8 shall tor exceed the inimum quali& ons of Regu 1 -'.euaoyG~s, orNISahrds acceptabi NRCFR st 5@ *
- Al 5.3.2 For the purpose of 10 CFR 55.4. a licensed SenIorl5 Operator'nand a licensedIerator( are those Individuals who Inaddition to meeting the requirements of 5.3.1, perform the functions described In .: .
10 CFR 50.54(m). : j'. . . ). WOG STS 5.3 11 Re.2, 04130101; Attachment I, V e16,'.-Rev. 1, Pa :'.'48 of 256
Attachment 1, Volume 16, Rev. 1, Page 49 of 256 5.3 INSERT 1 ANSI N18.1-1971 for comparable positions, except for the radiation protection manager and the operations manager. The radiation protection manager shall meet or exceed the qualifications of Regulatory Guide 1.8, Septemnber 1975. The operations manager shall be qualified as required by Specification 5.2.2.e. Insert Page 5.3-1 Attachment 1, Volume 16, Rev. 1, Pa~ge 49 of 256
Attachment 1, Volume 16, Rev. 1, Page 50 of 256 JUSTIFICATION FOR DEVIATIONS ITS 5.3, UNIT STAFF QUALIFICATIONS
- 1. The ISTS Reviewer's Note has been deleted since it is not intended to be included in the ITS.
- 2. The brackets are removed and the proper plant specific information/value is provided.
- 3. Grammatical/typographical error corrected. The terms in 10 CFR 55.4 and 10 CFR 50.54(m) are "Senior Operator" and "Operator."
- 4. Change made for consistency with the terminology used in other Specifications.
CNP Units 1 and 2 Page 1 of 1 Attachment 1, Volume 16, Rev. 1, Page 50 of 256
Attachment 1, Volume 16, Rev. 1, Page 51 of 256 Specific No Significant Hazards Considerations (NSHCs) Attachment 1, Volume 16, Rev. 1, Page 51 of 256
Attachment 1, Volume 16, Rev. 1, Page 52 of 256 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 5.3, UNIT STAFF QUALIFICATIONS There are no specific NSHC discussions for this Specification. CNP Units 1 and 2 Page 1 of 1 Attachment 1, Volume 16, Rev. 1, Page 52 of 256
, Volume 16, Rev. 1, Page 53 of 256 ATTACHMENT 4 ITS 5.4, Procedures , Volume 16, Rev. 1, Page 53 of 256 , Volume 16, Rev. 1, Page 54 of 256 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) , Volume 16, Rev. 1, Page 54 of 256
Attachment 1, Volume 16, Rev. 1, Page 55 of 256 ITS 5.4 ITS
.0 ADAM'?ITRATWE CONTROLS 5.4 I See rrS I 5.5 J 5.4.1 6+/-1 Wrbm pwtcedm sbUIbe estiblisnd fmpk1med and =mhdawd covenne to ac=vizie iecn~e bekir 5.4.1.a a. The applicable proaed= eccmemed In Appeot A' of Reguktxy Guide 133, Rerr. 2, Febuur, 1971-
- c. DdehA C. Delete&.
Id. C OLPRt}R L 5.4.1.e e. 0 DOSE CM AI NMMA]NUAL G .2 5.4.1.c L *wit Qui~ Pxgnm har eM= an envvzn Iwsist h da= im I RcEulroygdl.2l.RL-.I.Ame4,tndR axoy~;4.1,Racv.I.prl975. 5.4.1 .e g. CYCC orTn rw, yikrovides crob ltoatskddUFSA.R. E> 4.1,. yidie idt m -. tO pr' tF*ar "Cm d Width the z 5.4.1 .d 6.8.2 ged o pe6.1dae din sball be Asmx= Prto 6 UTeP-tL 6.33 Deleted *
-lAdd prrposed Specificaflon 5.4.1.e M.
COOK NUCLEAR PLANT-UNIT 1 Page6 AMENDMNT ,%M4KMM93
.24,I2KZU3,261 Page 1 of 2 Attachment 1, Volume 16, Rev. 1, Page 55 of 256
Attachment 1, Volume 16, Rev. 1, Page 56 of 256 ITS 5.4 ITS 6.8 ADMBNIsrRATMvK CONTROLS 5.4 LB RODRI=PRGA ..See ITS1 f 5 .5 5.4.1 6.8.1 Writtn prfoeure be eunbhdubefiurp vedand uritand wcdg fAativitea refreced bedw. 5.4.1.a L TX aliabk Predu re =ded i AppWakid' A of RePulzy Oiid 133, Rev. 2, Febray 1973. 1-1 JAM proposed Specificefion
- b. Deleted I. 1fOCESSC0 OLP etia()
5.4.1.e ILosrr DOS TIy~oN MANuAL 5.4.1.c 5.4.1.e 5.4.1.d IL Fim rte P oim Pros=a iMp! uilon. 16.8.2. 1E& Md&= =dsftil.j~yl.p6 4 above W'daes "eeJ, iicnIn Prq*m Deacd1idn, A43endlzC, ii~s QulfiaicA Iua 6.83 Deleft& 1 4 4 Add proposed Specificabon 5.4.1.e D e COOK NUCLEAR PLANTAJNIT2 Page 6-6 AMEDMET St, 43% ~I431 M 43343,
",244 Page 2 of 2 Attachment 1, Volume 16, Rev. 1, Page 56 of 256
Attachment 1, Volume 16, Rev. 1, Page 57 of 256 DISCUSSION OF CHANGES ITS 5.4, PROCEDURES ADMINISTRATIVE CHANGES A.1 Inthe conversion of the CNP Current Technical Specifications (CTS) to the plant specific Improved Technical Specifications (ITS), certain changes (wording preferences, editorial changes, reformatting, revised numbering, etc.) are made to obtain consistency with NUREG-1431, Rev. 2, "Standard Technical Specifications-Westinghouse Plants" (ISTS). These changes are designated as administrative changes and are acceptable because they do not result in technical changes to the CTS. A.2 CTS 6.8.1.e requires procedures for implementation of the OFFSITE DOSE CALCULATION MANUAL (ODCM) and CTS 6.8.1.g requires procedures for the implementation of the Component Cyclic or Transient Limits Program. ITS 5.4.1 requires procedures for various activities, but does not specifically list the ODCM and the Component Cyclic or Transient Limits Program. This changes the CTS by removing the explicit requirements for written procedures for implementation of the ODCM and the Component Cyclic or Transient Limits Program. This change is acceptable because implementing procedures for the ODCM and the Component Cyclic or Transient Limits Program are required by ITS 5.4.1.e. ITS 5.4.1.e (added as described in DOC M.2) requires that written procedures be established, implemented, and maintained for all programs and manuals in ITS 5.5 (including the ODCM and Component Cyclic or Transient Limits Program). Therefore, it is not necessary to specifically identify each program in ITS 5.4.1. This change is designated as administrative because it does not result in technical changes to the CTS. MORE RESTRICTIVE CHANGES M.1 ITS 5.4.1.b requires that written procedures shall be established, implemented, and maintained for the emergency operating procedures required to implement the requirements of NUREG-0737 and NUREG-0737, Supplement 1, as stated in Generic Letter 82-33. The CTS does not include this requirement. This changes the CTS by adopting a new requirement for emergency operating procedures. The purpose of ITS 5.4.1.b is to ensure that written procedures are established, implemented, and maintained covering the emergency operating procedures to implement the requirements of NUREG-0737 and NUREG-0737, Supplement 1, as stated in Generic Letter 82-33. This change is acceptable because it is consistent with an existing requirement to comply with NUREG-0737 and NUREG-0737, Supplement 1, as stated in Generic Letter 82-33, for emergency operating procedures. This change is designated more restrictive because it imposes a new requirement for procedures within the Technical Specifications. M.2 ITS 5.4.1.e requires that written procedures shall be established, implemented, and maintained for all programs specified in Specification 5.5. The CTS does not include this requirement for any program except the ODCM and the Component CNP Units 1 and 2 Page 1of 3 Attachment 1, Volume 16, Rev. 1, Page 57 of 256
Attachment 1, Volume 16, Rev. 1, Page 58 of 256 DISCUSSION OF CHANGES ITS 5.4, PROCEDURES Cyclic or Transient Limits Program. This changes the CTS by adopting a new requirement for procedures to address all programs described in ITS 5.5. The purpose of ITS 5.4.1.e is to ensure that written procedures are established, implemented, and maintained covering all programs specified in ITS 5.5. This change is considered acceptable because it requires written procedures, including proper procedure control to address programs required by ITS 5.5. This change is designated more restrictive because it imposes new requirements for procedures within the Technical Specifications. RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES LA.1 (Type 3 - Removing Procedural Details for Meeting TS Requirements or Reporting Requirements) CTS 6.8.1.d requires that written procedures for the PROCESS CONTROL PROGRAM (PCP) be established, implemented, and maintained. The ITS does not include these requirements. This changes the CTS by moving the requirements to the UFSAR. The removal of these details, which are related to meeting Technical Specification requirements, from the Technical Specifications is acceptable because this type of information is not necessary to be included in the Technical Specifications to provide adequate protection of public health and safety. The PCP implements the requirements of 10 CFR 20, 10 CFR 61, and 10 CFR 71. Compliance with these regulations is required by the CNP Units I and 2 Operating Licenses, and written procedures are necessary to ensure compliance with the program. Regulations provide an adequate level of control for the affected requirements, and inclusion of this requirement in the Technical Specifications is not necessary. Also, this change is acceptable because these details will be adequately controlled in the UFSAR. Any changes to the UFSAR are made under 10 CFR 50.59 or 10 CFR 50.71(e), which ensures changes are properly evaluated. This change is designated as a less restrictive removal of detail change because details for meeting Technical Specification and regulatory requirements are being removed from the Technical Specifications. LA.2 (Type 3 - Removing Procedural Details for Meeting TS Requirements or Reporting Requirements) CTS 6.8.1.f requires written procedures be established, implemented and maintained covering the Quality Assurance Program for effluent and environmental monitoring, "using the guidance in Regulatory Guide 1.21, Revision 1,June 1974, and Regulatory Guide 4.1, Revision 1,April 1975." ITS 5.4.1.c does not include the Regulatory Guide references. This changes the CTS by moving the references to the Regulatory Guides to the Quality Assurance Program Description (QAPD). The removal of these details, which are related to meeting Technical Specification requirements, from the Technical Specifications is acceptable CNP Units 1 and 2 Page 2 of 3 Attachment 1, Volume 16, Rev. 1, Page 58 of 256
Attachment 1, Volume 16, Rev. 1, Page 59 of 256 DISCUSSION OF CHANGES ITS 5.4, PROCEDURES because this type of information is not necessary to be included in the Technical Specifications to provide adequate protection of public health and safety. The ITS still retains the requirement for written procedures covering quality assurance for effluent and environmental monitoring. Also, this change is acceptable because these types of procedural details will be adequately controlled in the QAPD. Any changes to the QAPD are made under 10 CFR 50.54(a), which ensures changes are properly evaluated. This change is designated as a less restrictive removal of detail change because references for meeting Technical Specification requirements are being removed from the Technical Specifications. LA.3 (Type 3 - Removing Procedural Details for Meeting TS or Reporting Requirements) CTS 6.8.2 requires that each procedure and administrative policy of Specification 6.8.1, and changes to these documents, including temporary changes, be reviewed prior to implementation in accordance with the QAPD. ITS 5.4 does not include this requirement. This changes the CTS by moving these details of procedure and administrative policy reviews to the QAPD. The removal of these details, which are related to meeting Technical Specification requirements, from the Technical Specifications is acceptable because this type of information is not necessary to be included in the Technical Specifications to provide adequate protection of public health and safety. ITS 5.4.1 still retains the requirement for written procedures required by the Technical Specifications to be established, implemented, and maintained. Regulations provide an adequate level of control for the affected review requirement. The requirements for establishment, maintenance, and implementation of procedures related to activities affecting quality are contained in 10 CFR 50, Appendix B, Criterion II and Criterion V and ANSI N18.7-1976 (ANS 3.2-1976). In accordance with these requirements, the QAPD includes adequate detail with respect to administrative control of procedures related to activities affecting quality and nuclear safety, including the review requirements associated with maintenance of these procedures. Also, this change is acceptable because these types of procedural details will be adequately controlled in the QAPD. Any changes to the QAPD are made under 10 CFR 50.54(a), which ensures changes are properly evaluated. This change is designated as a less restrictive removal of detail change because references for meeting Technical Specification and regulatory requirements are being removed from the Technical Specifications. LESS RESTRICTIVE CHANGES None CNP Units I and 2 Page 3 of 3 Attachment 1, Volume 16, Rev. 1, Page 59 of 256
Attachment 1, Volume 16, Rev. 1, Page 60 of 256 Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs) Attachment 1, Volume 16, Rev. 1, Page 60 of 256
Attachment 1, Volume 16, Rev. 1, Page 61 of 256 Procedures 5.4
- CTS 5.0 ADMINISTRATIVE CONTROLS 6.8 5A Procedures 5A.1 Written procedures shall be established, Implemented, and maintained covering the following activities:
- a. The applicable procedures recommended in Regulatory Guide 1.33, Revision 2, Appendix A, February 197$0 P6c 14.1 b. The emergency operating procedures required to Implement the requirements of NUREG-0737 and(ENUREG-0737, Supplement 1, as stated InrGeneric Letter 82-3:t;)
- c. Quality assurance for effluent and environmental monitoringo (D g 4;OC6,A,k. d. Fire Protection Program implementatio nd 0
- e. All programs specified In Specification 5.5.
6.9.(.Al 6.8. I.9
- WOG STS. 5.4 - 1 Rev. 2, 04130101 Attachment 1, Volume 16, Rev. 1, Page 61 of 256
Attachment 1, Volume 16, Rev. 1, Page 62 of 256 JUSTIFICATION FOR DEVIATIONS ITS 5.4, PROCEDURES
- 1. The brackets are removed and the proper plant specific information/value is provided.
- 2. These punctuation corrections have been made consistent with the Writer's Guide for the Improved Standard Technical Specifications, NEI 01-03, Section 5.1.3.
- 3. Grammatical errors corrected.
CNP Units I and 2 Page 1 of 1 Attachment 1, Volume 16, Rev. 1, Page 62 of 256
Attachment 1, Volume 16, Rev. 1, Page 63 of 256 Specific No Significant Hazards Considerations (NSHCs) Attachment 1, Volume 16, Rev. 1, Page 63 of 256
Attachment 1, Volume 16, Rev. 1, Page 64 of 256 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 5.4, PROCEDURES There are no specific NSHC discussions for this Specification. CNP Units 1 and 2 Page 1 of 1 Attachment 1, Volume 16, Rev. 1, Page 64 of 256
, Volume 16, Rev. 1, Page 65 of 256 ATTACHMENT 5 ITS 5.5, Programs and Manuals , Volume 16, Rev. 1, Page 65 of 256 , Volume 16, Rev. 1, Page 66 of 256 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) , Volume 16, Rev. 1, Page 66 of 256
Attachment 1, Volume 16, Rev. 1, Page 67 of 256 ITS 5.5 ITS sLO jADIDix SM TI CONTROIS 5.5 6.8.4 The following program shall be establisd, Impemnted, and maintned: 5.5.3 aL Radioactiv
=1guf
- ok s zoam 5.5.3 A program ha be provided conirinlng with 10 CFR 50zu for the control of radioactive effluents and for mintainfng the doses to NMlMB OF THE PUBLIC from radloactireffluants as low as reasonably achloyehl& Toe program (1) saln be contained In the.ODCK, (2)shell be ipleemented by operating procedures, and (a) shaM Include remdi actions to be taken whenever the program limits as exceeded Theprogram sh in.d tfollwingthe eement:
5.5.3.a 1) limitations an the operability of radioactive liqdd and gueass monitoring inrnnetation Iclduding urneance tests and setpoint determination in accordance wi the methodology In the ODCM - 5.5.3.b 2) LItAtions On the concenfrations of radioactive matri released In liquid e9usna to UNRESTCTMD AREAS conforming to 10 CFR 20.1001-20.2402, Appendix B, Table 2, Column 2, I 5.5.3.c 3) Mznitorln sampling, and analysis of radioactive liquid and gaaeous
*ffuents pursuant to 10 CPR 20.1802 and with the methodolog and I parameters in the ODCM,.
5.5.3.d 4) Umitations on the annual and quarterly doses or dose commitment to'a IM MEPOF TEM PUBIC from radioactive materials In liquid effluenb released from each unit to UNRESTRCTED AREAS conforming to Appendix I to 10 CFRPart 60, 5.5.3.e 5) Determdnatn of cumulatvn and prqdected dose. conhOiutlona from radioactive efluents for ths current calear quarter amd current calendar yer in aordnce W the metioology and paRmeTs in the ODCM at leas tvery ldays,
- 6) Imitatioe on the operabt and use of the lih*id d gaseous effuent 5.5.3.f treatment s:tei to ensure that the appropriate portions of these systems Ar used to reduce releas f radioactivity when the projected dom In sa91-day period would exceed 2 percent of the guidelines for the annual doee or dose commimnt conforming to Appedix I to 10 CFE Part 50, ..
COOK NUCLEAR PLANT.UMIT Pape 6-7 AhMF4MENTUO,2a, 245 Page 1 of 69 Attachment 1, Volume 16, Rev. 1, Page 67 of 256
Attachment 1, Volume 16, Rev. 1, Page 68 of 256 ITS 5.5 ITS 6.0 ADMINLSrRATIVE CON'TROIS PROCEDURES AND PROGRAMS (Continued) 5.5.3.g 7) Limitations on the dose rate resulting from radioactive material released in gaseous effluents to areas beyond the SITE BOUNDARY halm be limited to the following-a) For noble gases Less than or equal to a dose rate of 500 mrem/year to thetotalbodyandlessthnorequaltoadoserateof 3000mremyear to the skin, and b) For lodine-131, Iodine-133, tritiuim, and for all radionuclides in particulate form with half-lives greater than 8 days: Less than or equal to a dose rate of 1500 mremlyear to any organ. 5.5.3.h 8) Limitations on the annual and quarterly air doses resulting from noble gases released in gaseous effluents from each unit to areas beyond the SITE BOUNDARY conforming to Appendix Ito 10 CFR Part 60, 5.5.3.1 9) LImitations on the annual and quarterly doses to a MEMBER OF THE PUBLIC from Iodine-131, Iodine-133, tritium, and all radionuclides in particulate form with balf-lives greater than 8 days in gaseous effluents released from each unit to areas beyond the SITE BOUNDARY conforming to Appendix I to 10 CFR Part 60, and 5.5.3.j 10) Limitations on the annual dose or dose commitment to any MEMBER OF THE PUBLIC due to releases of radioactivity ant radiation fro ua im. fuel cycle sources conforming to 40 CFR Part 190. The provisions of SR 3.0.2 and SR 3.0.3 are b applicable to the Radioacive Effluent Control I b. .adiolo ea] EnyJonmental M ri Poaml Program Surveillance Frequencies. A pro n s1l be provided to onitor the radiation and *onucides in the environ of the plant. The program all provide (1) representa measurements of raio h in the highest pa exposure pathways, and ) verification of the accura of the effluent monitorng and modeling of en inental exposure pathw . The program shall (I be ontained. in the 0 , (2) conform to the gui afAppendixI to 10 CFR at50, and (3) include the owing:
- 1) Monitoring, sampling, lthe ODCM,* l sin, and reporting of radtion and radionuclides in'the environment in adance with the methodo gy and parameters in 2) A Land Use Census to that changes in the of areas at and beyond the SITE BOUNDARY Identified and that m cations to the monitoring program are made if by the results of this sus, and
- 8) Participation in a iterlboratory Comparison gram to ensure that independent checks the precision and accura y of the measurements of radioactive materi in environmental sample atrices are performed as part of the quality urance program for environental monitoring.
COOK NUCLEAR PLANT-UNIT 1 Pae 6-8 AMENDMENT 1., a2U, 245 Page 2 of 69 Attachment 1, Volume 16, Rev. 1, Page 68 of 256
Attachment 1, Volume 16, Rev. 1, Page 69 of 256 ITS 5.5 ITS 6.0 ADMINISTRATIVE CONTROLS PROCEDURES AND PROGRAMS (Continued) 5.5.12 6.8.5 Technical Specifications Bases Control Prorraxn This program provides a means for processing changes to the Bases of these Technical Specifications. 5.5.12.a a. Changes to the Bases of the Technical Specification shall be made under appropriate administrative controls and reviews. 5.5.1 2.b b. Licensees may make changes to Bases without prior Nuclear Regulatory Commission approval provided the changes do not require either of the following:
- 1. A changes in the Technical Specification incorporated in the license or
- 2. A change to the Updated Final Safety Analysis Report or Bases that requires Nuclear Regulatory Commission approval pursuant to 10 CFR 50.59.
5.5.12.c c. The Bases Control Program shall contain provisions to ensure that the Bases are maintained consistent with the Updated Final Safety Analysis Report. 5.5.12.d d. Proposed changes that meet the criteria of Specification 6.8.5.b above shall be reviewed and approved by the Nuclear Regulatory Commission prior to implementation. Changes to the Bases Implemented without prior Nuclear Regulatory Commission approval shall be provided to the Nuclear Regulatory Commission on a frequency consistent with 10 CFR 50.71(e). 6.9 REPORTING REQUIREMENTS ROUTINE REPORTS 6.9.1 In addition to the applicable reporting requirements of Title 10. Code of Federal Regulations. the following reports shall be submitted to the Regional Administrator unless otherwise noted. STARTUP REPORT 6.9.1.1 A summary report of plant startup and power escalation testing shall be submitted following (1) See ITS1 5.6 ) receipt of an operating license, (2) amendment to the license involving a planned Increase in power level, (3)installation of fuel that has a different design or has been manufactured by a different fuel supplier, and (4) modifications that may have significantly altered the nuclear, thermal, or hydraulic performance of the plant. 6.9.1.2 The startup report shall address each of the tests identified in the FSAR and shall include a description of the measured values of the operating conditions or characteristics obtained during the test program and a comparison of these values with design predictions and specifications. Any corrective actions that were required to obtain satisfactory operation shall also be described. Any additional specific details required in license conditions based on other commitments shal be included in this report. COOK NUCLEAR PLANT-UNIT I Page 6-9 AMENDMENT 2. 4- 4,9. 226 , 281 I Page 3 of 69 Attachment 1, Volume 16, Rev. 1, Page 69 of 256
Attachment 1, Volume 16, Rev. 1, Page 70 of 256 ITS 5.5 ITS JAN 27 204 (2) Technical Specifications The Technical Specifications contained In Appendices A and B, as revised through Amendment No.4QQare hereby Incorporated Inthe license. The licensee shall operate the facilty In accordance with the Technical Specifications. (3) Less than Four Loop Operation The licensee shall not operate the reactor at power levels above P-7 (as defined In Table 3.3-1 of Specification 3.3.1.1 of Appendix A to this license) with less than four reactor coolant loops In operation until (a) safety analyses for less than I four loop operation have been submitted, and (b) approval for less than four loop operation at power levels above P-7 has been granted by the Commission by amendment of this license. (4) Indiana Michigan Power Company shall implement and maintain, In effect, all provisions of the approved Fire Protection Program as described In the Updated Final Safety Analysis Report for the facility and as approved In the SERs dated December 12, 1977, July31, 1979, January30, 1981, February7, 1983, November 22, 1983, December 23, 1983, March 16, 1984, August 27, 1985, June 30, 1986, January 28, 1987, May 26, 1987, June 16, 1988, June 17, 1988, June 7, 1989, February 1, 1990, February 9,1990, March 26, 1990, April 26, 1990, March 31, 1993, April 8, 1993, December 14,1994, January24, 1995, April 19, 1995, June 8, 1995, and March 11, 1996, subject to the following provision: The licensee may make changes to the approved fire protection program without prior approval of the Commission only if those changes would not adversely affect the ability to achieve and maintain safe shutdown In the event of a fire. (5) Deleted by Amendment No. 279 (6) Deleted by Amendment No. 80 5.5.8 (7) Secondary Water Chemistry Monitoring Program The licensee shall Implement a secondary water chemistry, monitoring program Ia to Inhibit steam generator tube degradation. This proisbe esc ed I Ithestton c5?fn strv mahual and shall indude: 5.5.8.a 1. Identification of a sampling schedule for the critical parameters and control points for these parameters; 5.5.8.b 2. Identification of the procedures used to measure the values of the critical parameters; Amendment No. 279 Page 4 of 69 Attachment 1, Volume 16, Rev. 1, Page 70 of 256
Attachment 1, Volume 16, Rev. 1, Page 71 of 256 ITS 5.5 ITS
.4e l 5.5.8.c 3. Identification of process sampling points; 5.5.8.d 4. Procedure for the recording and management of data; 5.5.8.e 5. Procedures defining corrective actions for off control point chemistry conditions; and 5.5.8.f 6. A procedure Identifying (a) the authority responsible for the Interpretation of the data, and (b)the sequence and timing of administrative events required to initiate corrective actions.
(8) Deleted by Amendment No. 279 I (9) Deleted by Amendment No. 279 (10) Deleted by Amendment No. 279 Il (11) Deleted byAmendment No. 279 D. Physieal Protection The licensee shall fully implement and maintain In effect all provisions of the Commission-approved physical security, guard training and qualification, and safeguards contingency plans Induding amendments made pursuant to provisions of the Miscellaneous Amendments and Search Requirements revisions to 10 CFR 73.55 (51 FR 27817 and 27822) and to the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The plans, which contain Safeguards Information protected under 10 CFR 73.21, are entitled: 'Donald C. Cook Nudear Plant Security Plan, with revisions submitted through July 21, 1988; 'Donald C. Cook Nudear Plant Training and Qualification Plan,' with revisions submitted through December 19,1986; and "Donald C. Cook Nuclear Plant Safeguards Contingency Plan,' with revisions submitted through June 10, 1988. Changes made In accordance with 10 CFR 73.55 shall be I..1 implemented in accordance with the schedule set forth therein. E. Deleted by Amendment No. 0 ddpoposedSystems F. Deleted by Amendment No.80
/
G. In all places of this license, the reference to the Indiana and Michigan Electric Company is amended to read Indiana Michigan Power Company. 5.5.2 H. System Integrity The licensee shall Implement a program to reduce leakage from systems outside I containment that would or could contain highly radioactive fluids during a serious I transient or accident to as low a practical levels.i The program shall include the following: Amendment No. 279 Page 5 of 69 Attachment 1, Volume 16, Rev. 1, Page 71 of 256
Attachment 1, Volume 16, Rev. 1, Page 72 of 256 ITS 0 ITS 5.5 I 5.5.2 1. Provisions establishing preventive maintenance and periodic visual Inspection requirements, and
- 2. Integrated leak test requirements for each system at a freauencv Ireueinngcycie ipyervaF4 I. Idn h rvsosofS .. r plcbe L1. Idin The licensee shall I plement a program which will ensure the pability to accurately determine the aelme concentration In vital areas under acci nt conditions. This program shall In e the following:
-- LAe
- 1. TrainIng ersonnel,
- 2. Procedu for monitoring, and
- 3. Provisits for maintenance of sampling and analys equipment.
J. The licensee is authorized to use digital signal processing Instrumentation in the reactor protection system.
- 3. This amended license is effective as of the date of Issuance and shall expire at midnight October 25, 2014.
FOR THE NUCLEAR REGULATORY COMMISSION Original signed by Roger S. Boyd Roger S. Boyd, Director Dvision of Project Management Office of Nuclear Reactor Regulation
Enclosure:
Appendix A - Technical Specificatons Date of Issuance: March 30, 1976 Amendment No. 279 Page 6 of 69 Attachment 1, Volume 16, Rev. 1, Page 72 of 256
Attachment 1, Volume 16, Rev. 1, Page 73 of 256 ITS 5.5 ITS I PnetSS COWrML Pno r(Pc 1.28 The RDCUSS CORL PRORM (PCP) shall contain the current formulas, See CTS sampling, analyses. tests, and determinations to be made to ensure that proesng and packaging of solid radioactive wastes based on demonstrated 6.0 proces Ing of actual or simulated vet solid asteas will be accomplished In such w assure compliance with 10 CM Parts 20. 61, and 71, stat. regulats
- burbal ground requirements, and other requireents governing the dispocal of solid radioactive waste.
1.19 Deleted.' 5.5.1 OMIJU DOSE cQx.mAYOff KANUAL (0Ca4Y 1.30 The orni sz DOSIo xumUAC K8L (o=Q) Cbl contai the 5.5.1a8 methodology and parameters used in the calculation of offaite doses resulting from radioactive gaseous and liquid effluents, in the calculation of gsCJOW and liquid effluent monitoring alarm/trip
- tpolnte Itnthe engpduet of the Environmenta 1 Ritlloglea I M-lelsLssez Sbc,ODCH contan (a)th eelve Xffusnt Controls and Radiological yadio 5.5.1.b Environmental Monitoring Pregr r quired by Section 6.8.4 and (2) dscriptLins of the Informs ton that should be included in the Ara-l Radiological Envirotmental Operating and Annual Radioactive fflusunt Rel ase Reports required by SpecificatLons 6.9.1.6 and 6.9.1.7.
CASKUS lPAMAM ?TMTKMD 3YSTF 1.31 A CGASOUS RADUSTZ TREATIMy BY$=hI is ny system desned and 4 Installed to reduce radioactive gaseous efflunts by collctng primary coolant system off-gaose from the primary system ad providing for delay or holdup for the purpose of reducing the total radioactivity prior to release SeeITS FS to the environment. Chapter 1.0 J VE!UTIAON MO1ADST TOPATEDT RSYD( 1.32 A V MTIflKO MI&HSM IREOAMhZUT hS11 is say system designed and Installed to reduce gaseous radioiodine or radioactive matarial In particulate form In effluents by passing ventilation or vent exhaust gsas through charcoal absorbers and/or EVtA filters for the purpose of removing iodines or particulate. froo tho gaseous exhaust stress prior to the release to the environment. Such a system IJ not considared to have any effect on noble gas effluents. Engineered Safety Feature (StU)atmospheric cleanup systm are not considered to be VDf11ATIW. =2ST TS ATED? SYSTCK components. l.S3 VIWC= or TUMSI=a Is the controlled process of discharging air or gsa from a conf nmant to maintain temperature, pressure, humidity, concentration or other operating condition. in such a manner that replacement air or gas Is required to purify the confinement. 1.34 VZNrrIu Is the controlled proess of discharging air or gas from a confinement to ailntain temperature, pressure, himidity, concentration or 4 other operating condition, in such mncamr that replacement air or gas Is not provided or required during VDT13O. Vent, used in yste Dams, does not imply a VDTIO process. COOK NIUCLEAR PLAM? - VNIT IL 1.6 AKENDKENr 90. ", 189' Page 7 of 69 Attachment 1, Volume 16, Rev. 1, Page 73 of 256
Attachment 1, Volume 16, Rev. 1, Page 74 of 256 ITS 5.5 ITS 0 3/4 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4.0 APPLICABILITY lThis provision shall not prevent entry Iwuo OPERATIONAL MODES or other speccified conditions In the l Sees ITS IApplicability that are required to comply with ACTIONs or that are part of a shutdown of the unit. I Section 3.01 5.5.6 4.0.5 Surveillance Requirements for inservice l io testing of ASME Code Class 1. 2. and 3; 3 Xbe applicable as follows: lumps nd valv SURVE ILLANCE REQUIREMENTS 5.5.6.a b. interasified on X SME Boiler and Pressure C aessel nd Addends for the testing activities required by the l H a n d applicatbie Addenda shall be applicable as follows in these LTechnical specifications: ler a Pressure V Ruired freqencies for performing and applicable Addenda ternioo or inservicebg o fd esting activities lnserv o tng criteria l[weekly// At leastgcpe7das _A.s plonthM At lsfnepe3 ay Quarterly or every 3 mon~ts At ecasxC2d A. I aninidy or every 6 rmots At~e Yearly or 2nnu ly At least once per 366 days a c e A l c 731days
- . t2 years I 5.5.6.b c. The provisions of Specification 4.0.2 are applicable to the above required frequencies for performing i f t s t n --vtis LA.3) id. Perfor~~ of the abv nevcl~ testin acdvtiv~ eI a dton to othr l swc~tledSurveillance Required <J 5.5.6.d e. Nothing In the ASME Boler and Pressure Vessel Cod shall be construed to supersede the A-15 requiredents of TicalSc aon.
4.0.6 Deleted 4.0.7 Deleted COOK NUCLEAR PLANT-UNIT I Page 3/4 0-3 AMENDMmENT o :,X42,44A, ;a, 281 I Page 8 of 69 Attachment 1, Volume 16, Rev. 1, Page 74 of 256
Attachment 1, Volume 16, Rev. 1, Page 75 of 256 ITS 5.5 ITS 1W4I LDInNG COlb4DOM FOR~ CrZRA1ON AND 81RVEUIcAz BRQUMEMNDA SI 3M4A WRArR 2OO1An SW 3A3 Ewh smm Smr aha be OPERABLE. See USr APPLIAMMl, MODES 1.2.3 na4. - 3.4.13 Wh an at Nme. G C, baphers reux do bkqM"b puerno*~)w OPERABLE amm berowng T., awn 207 4A3.0 Pachaem j mraha be iemmmsued OPERABLE by pufxma of ft 9"flwb
.ipemd kwvice bquefdi pUo p ftreqmewoOSpecIfMaf 4.OJ.5 end 5.5.7 U=m Garra Samneh Se&daf atvl Tmm - Bc a Sp -em a f be A- I OPERABDLE *zlg Iow- by wne dmlaspeedq K bsun de MU= umb of Sem
!E4 5.5.7 4AJ.1 Zmcrw s-m Oimam shm.
~o Tb7W
-mcd W
aemb Ad prpoed ITS 5.5.7 generic prog~ram des~criprion TierSamee Sek 1~ arAbioemlm, - Thesta dwff~dco ad do Kintm tbe W xqub abebe a
-G0 woMW~ In Tft 442. Ile kwavic bsetmc of mm pmmtubes sWabe I er - d as ft0eqzuek, -speid k Sped~atim 4.4J ud de d pb,,Ada be rnti scowibb f W , Iof pec~alm44.5 'betaubes elvesed hr ach 5.5.7.a -
Ltatubts seoied fo &=ea opdm daBbe allctd aon ard besLc 5.5.7.a.1 a. Wber Inedu rt Wt shaWasc sha w~theembauy Inw~a amwalm t be bIemr OM. kar bE0I s Of daVAt haeea be ftm tm eftd1a mm 5.5.7.a.2 peovz~ b&atpem) .1mb m s aWkbeim semm
- 2. Aluttube op ovlawy hd dwakkaN wdipiawo (pessrtoequal to 5.5.7.a.2.a) 20%)*akW wbC=Plize&
sfetd Candm d h ftSpe~nd Opraf tq nMde4wk1 - 3A11. Kamabeakad4.13 arde lustieg a Icog es Lialeg LSeeITS) COOM tALTI PLACr4 I hp 2 447 A)UIMbW4T i,4,WMU 238 Page 9 of 69 Attachment 1, Volume 16, Rev. 1, Page 75 of 256
Attachment 1, Volume 16, Rev. 1, Page 76 of 256 ITS 5.5 ITS 3M XBOMO~COKDMhOM FM OFRAMON AND EMVRVEANCE FXQUZDaNflI 3MA REACTR COOLAWXMTM 5.5.7.a.2.b) 2. Txdbes lo bu mm wbemo b bficd PONWla haer pov m. 5.5.7.a.2.c) 3. A wbe kx=AM 0 l Spedin 4A.IA~AJ iu be pI 11 -4ai n cubodecood wbc uUM autedd= w amonsppump of 6e eddy F- -pmbefritcbehw mdmdhss be IFdMduldwwcbe donD be aeiecodwd m1 ~ m matube bnpedon 5.5.7.a.3 C. Mmrnsedad =dw Wsqc Of equkeby Tdb4A-Z Owft eb mkbuW od~m
- be *cad wap.1W tab I ecdaaproIa:
5.5.7.a.3.a) 1. Mm rdmo misond fbr ft bcb~f de tua f fto stb wta of fat~r taM MM wh" moUf wI haI I wem uway ftnd 5.5.7.a.3.b) 2. 7be b epcdw bich s 6= poahar of dovi wb hprc in w 5.5.7.b 7bC-I Lm~wp d beSda taw M C= Ohesamdmft dep dd=ac u s f C4dLantws 5r ofwu to wW 10% b ecod man 6ev& Imf.Id as ofpdo C3M m tubu 10%c tb ma be hopecea wdca k aiu thamor ~ bu1l Offb ha se b sas debPOP sI WOONKH~zapwAMMGI hp39" Page 10 of 69 Attachment 1, Volume 16, Rev. 1, Page 76 of 256
Attachment 1, Volume 16, Rev. 1, Page 77 of 256 ITS 5.5 ITS 3/4 L~hdMNG cCO~MoMM FRM OMAMON AN XMW tVMIAJ RZQUMZ ID 3/4.A REACTOR COOIAPITS ITD 5.5.7.b Hm inahpections p..Wh* deailded tins gazi uzhM Slpiec (puCeM a or equal la 10%) tAuiuws winuz wbe facided hi ibe above,wcm.Calmhicm. 5.5.7.c 4A." EM*1 raunin-h Abov required Imp*I. kuipecdowom @se= m Okshalbe h 5.5.7.c.1 a. Th M bvI of W n NI.p~d stintW o am~drH FawerM b.wh hillb at kamra of = Im Man 12 w ngu24 caena awfte psweimow kqcaon If Mc Mealwi huea oI~lwbg =wl ader AWr I casW01 cc Vim Feo ewace I queectltsdmonw dw Peiouusly ml, cey depedifto bas ox cnond ai s sd~kmI depadedon lis ccfed kml Interva may be sawaW soa aiin atomndper 40 O. 5.5.7.c.2 b. fienosofb-tnod Iqm a sitsm Conuctdi smorwdme wI& Tsb 4A4.2 si 40 -- buask M h2 Cwpy 0-3, s Impaction ftequencV in be btemd to k mmpr20 mo& 7be kusaw In fiegmat &I Ul d iq mn ft arkkI d Ipecftcsdan 4AJJ~A de kmasvl sa dmbe edsd sadedai eo io 5.5.7.c.3 C. Addltna onchoiMke I urnt. edoinil be rmbmd an eah wmm m rh ecn.wrd a.!mftde fin donsi 'ffledhiT"h4A-26zagtft I -1avat1 5.5.7.c.3.a) 1. p~oyoscna 0*m lei ab (oxn tbob 180010 how tb"0,460 ohm 5.5.7.c.3.b) 2. A sd ccnacrur &gmsdon ft Opersdag Ruba NBu~b. 5.5.7 c.3.c) 3. A hosWcodsce sdma c re 1 wkman-to eta pgzute sftwds. 5.5.7.c.3.d) 4. A alomm 15onbdw1 1II1. keak. COOK NUCLX4R ?IANr4.IT I hp3/4 49AIBl IMITU, 1462W M O 238 Page 11 of 69 Attachment 1, Volume 16, Rev. 1, Page 77 of 256
Attachment 1, Volume 16, Rev. 11 Page 78 of 256 ITS 5.5 31 LvMT~GctrnN~oaoims ol mn~oN ANIDsuivnm LAHCEUE-QU3DZ1'I 31A RACWrR COOkANrMm4 5.5.7.d 4A.SA 5.5.7.d.1 L. AszWekdaSpdic 5.5.7.d.1 .a) I. j=[W muma emudo inIn dbzosbu.ramlor at img ofa %be*tm6i I- by bakkuft die of Ve~xtm Eddy-aCWf tft klgkd belo 2D% t t eke -w. ddda If daewcztie My be uiemm 5.5.7.d.1.b) 2. DL&t m a smvkabslo aalq wutqe. wwv or guow trd 5.5.7.d.1.c) 2. De e -w md k afftaVaagmr squwio 20%of 9 mbuwaU I ftkm mudby depbw. 5.5.7 d.l.d) 4. mmLC=9 do I= of6 dwv was ~Oomwcc , owdIby 5.5.7.d.1 .e) f-. mm =' b abdM .yd osuc xc f bb cLAtb 5.5.7.d.1.f) 6. i--11JW t k r I,I~6 &xb atorbyowd wbib t %60 6ft be I emund ftm Imco vmAzqa w~c. Wm kfte uig pI 5.5.7.d.1 .g) 7. Ize b ~t~ fkbbr *a~ibg coalua wudeg, or &m Dor atdw go bmak,a ed5W In4A353.c. *o,.. 5.5.7.d.1l.h) L bwA Imeto dma of do mm a Imuba=I A pow of tiny Cu kg Wbm) euFv~ly to~ 6Ubd b0t lop aqppon tOn mmlkg~. oi~i r~r Is. 3M44@ (DS~r~UUU238 Page 12 of 69 Attachment 1, Volume 16, Rev. 1, Page 78 of 256
Attachment 1, Volume 16, Rev. 1, Page 79 of 256 ITS 5.5 ITS 0 3M MNlfG COMDiON3 FM O1UATKIN AND PJRVMANCE RREQUfllEM 3/U REACTOR COOLANT SYTN 5.5.7.d.2 b. TI a S m ar D k dmtk OPEABLE asr compledu wmpaq zft d
*4ft aD aMt mdN ft pbqtn ih adAUDAs cm 11 th au amacb) xqJrid by TabXl 4A-2.
Theproisinsof SR 3.0.2 and SR 3.0.3 are applicabte to the SG Program test Frequencies. A.6 COO ?4NCLAR wANUM. 1*3/44-11 AMrUM 41iS MUkOKU 238 I 1* Page 13 of 69 Attachment 1, Volume 16, Rev. 1, Page 79 of 256
Attachment 1, Volume 16, Rev. 1, Page 80 of 256 ITS 0 ITS 5.5 RZACTOR COOLANT SYSTEMS Table 5.5.7-1 X1NIUK Ml!R 0O STEM GMURATORS TO B! TNSPECTED DURMO 1NSERVICR INSPECTION Preservice Inspection Yet No. of Steam Generators per Unit Four First Inservice Inspection TWo Second & Subsequent Inservice Inspections OnG2 Table Notation: Table 5.5.7-1 1. The inservice inspection nay be limited to one steam generator on a Footnote (a) rotating schedule encompassing 3 SX of the tubes (where N is the number of steam generators In the plant) if the results of the first or previous inspections indicate that all steam generators arm performing in a like manner. Note that under some circumstances, the operating conditions In one or more steam generators may be found to be more severe than those in other steam generators. Under such circumstances the sample sequance shall be modified to inspect the most severe conditions.
- 2. The third and fourth steam generators not inspected during the first Table 5.5.7-1 InservLco Inspection shall be inspected during the second and third Footnote (a) Inspections, respectively. The fourth and subsequent inspections shall follow the instructions described in I above.
COOC NUCLUX PLNT - UNIT I 3/4 4-13 £MD)!KT O. 10 166 Page 14 of 69 Attachment 1, Volume 16, Rev. 1, Page 80 of 256
ITS Table 5.5.7-2 I sm~i SM"AMzmac11tum REM* Aasm 1pd CU~MIIATOR TiMw REV WA INSPDCMoN 2MDSJAWLMECTON Aid=m*- WA 3!"SAMO1Z3QWt1rM keg* MIA' AaM=Req" WA A mbimu afS C-I "me I C2-i Pfdif A 1 mbA kMq, CI NO MA W/A wbm 23b ha I h £0. 0 .I C-2 Phg**cwuemAuui C-I Nane
-ap -kk~mw43 bAO in CD aCD C-2 Plug dcfizdve Kmb .I C-I Palmu OdImosUc-Imok ftft CII mcd= k K- l-3 to -u1 C-3 bae*
__________A M. isaUMC-&G _______2 CI Pla .s ri1 mm h WA WIA CD co tabu b edmb AhUS.0..1U A~..u Io 0, some S.b0ui23CC- Ptifm WA WAc.r 0 0 2
-4j cn S3m .amC-3 uom~m sCia 0, AAbsdcmhrSA kbfe A V= a ab .C WA NIA a)
M- wd ftu decahe mba. 1 Puip ufclmi SoNR
-- tonicUiI U
-U a)
I
£ S-3(Ntn)% N is die imbrof mum gmennlor In fte umn.and a Is die number of s'temenefuors urdding dmp an inspection.
1 CD.. CO. 01 0
-9
- 0) Cn (0
Attachment 1, Volume 16, Rev. 1, Page 82 of 256 ITS 5.5 ITS 0
,I 314 LIMITG CONDITIONS FOR OPERATION AND SURVEILANCE REQUIREMENTS 314A REACTOR COOLANT SYSTEM 3/4.4.10 STRUCTURAL INTEGRITY SeecTs~
31.4.10.1) ASME CODE CLASS 1. 2 and 3 COMPONENTS LIMIING CONDITION FOR OPERATION 3.4.10.1 The structural Integrity of the ASME Code Class 1. 2 and 3 components shall be maintained In accordance with Specification 4.4.10.1. APPLICABILITY: ALL MODES ACTION:
- a. With the structural integrity of any ASME Code Class I component(s) not conforming to the above requirements, restore the structural integrity of the affected component(s) to within its limit or isolate the affected component(s) prior to Increasing the Reactor Coolant System.temperature more than 50°F above the minimum temperature required by NDT considerations.
- b. With the structural integrity of any ASME Code Class 2 component(s) not conforming to the above requirements, restore the structural Integrity of the affected component(s) to within its limit or isolate the affected component(s) prior to increasing the Reactor Coolant System temperature above 200DF.
C. With the structural integrity of any ASME Code Class 3 component(s) not conforming to the above requirements, restore the structural integrity of the affected component(s) to within its limit or Isolate the affected conmponent(s) from sevice. SURVEILlANCE REOUIREMENTS 5.5.5 4A.10.1 I I In adSon to the re nents of Swocti 4o0n5T each reactor coolant pump flywheel sball be inspected by either qualified in-place UT examination over the volume from the Inner bore of the flywheel to the circle of one-bali the outer radius or a surface examination (magnetic particle l testing and/or penetant te ) of posed surfaces defuned bj the volume of the disassembled once every 10 years. pflywheels I (~Add proposed ITS 5.5.5 [generic program statement J rThe provisions of SR 3.0.2 and SR 3.0.3 are , applicable to the Reactor Coolant Pump Flywheel Inspection Program Surveillance Frequency. 11 COOK NUCLEAR PLANT-UNIT I Page 314 4-33 AMENDMENT 98,;, 281 I Page 16 of 69 Attachment 1, Volume 16, Rev. 1, Page 82 of 256
Attachment 1, Volume 16, Rev. 1, Page 83 of 256 ITS 5.5 ITS 314 LIMITNG CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREIEMS 314.6 CONTAINMENT SYSTMMS See rTS 3.6.1 J 5.5.14.b. 5.5.14.c, 5.5.14.d.1 5.5.14.d.1
- AM I With eidter (a) the measmed overall inteated contai t leakage rate exceeding 0.75 LJ or (b) with the measured 5.5.14.d.1 combined leakage rate for ill peetrations and valves subject to Types B and C tests exceeding 0.60 1.lreste the I overall integrated leakage rate to _ 0.75 La and to n e rate for tratow VAs su ect to TypeaB and C tats to0S0.60 th o rem MO actor Coortnt emp tuc abov 200°F. I 5.5.14.a 4.6.12 Perform leakage rate testing in accordance with 10 CFR 50 Appendix J Option B, except as wodifed by NRC-pproved e cpons, and RegulatoryGuide 1.163, dated September 1995.
See Notes I and 2. l
- a. Each ont- air lock shallbe verified to be in compliance with the requirements of See ITS 1
Spefiction3.6.1.3. 3.61 J
- b. jbe provisions ofpfwication 4.0.2 qmfot applicable. _L A.7e
[ Add proposed ITS 5.5.14.e Notes: A.?- 5.5.14.a.2 1 A one-time exception to the requirement to perform post-modification Type A testing is allowed for the steamn generators and associated piping as components of the contatiment barrier. For this case, ASME I Section XI leak testing will be used to verify the leak tightness of the repaired or modified portions of the containment barrier. Entry Into MODES 3 and 4 following the extended outage that commenced in 1997 maybe made to performthis testing. 5.5.14.a.1 2 ThcType A testing frequency specified in NEI 9401, Revision 0, Pazagraph 9.2.3, as ' ...at least once per 10 years based on acceptable performance bIst ' kI modified to be "...at least once per IS years based on acceptable performance history." Ihis change applies only to the interval following the Type A test performed In October 1992. COOK NUCLEAR PLANT-UNMT I Page 314 6-2 AMENDMENTW48,40,M,496,M0,2418, 274 Page 17 of 69 Attachment 1, Volume 16, Rev. 1, Page 83 of 256
Attachment 1, Volume 16, Rev. 1, Page 84 of 256 ITS 5.5 ITS 314 LDIlT1NG CONDMONS FOR OPERATnON AND SURVEnLLANCE REQUIRE:MIENTS 314.6 CONTALNMNTr SYSTENMS CONTAINMENT AIR LOCKS See ITS 1 3.6.2J LIMmNO CONDmON FOR OPERATION 3.6.13 Each containuent air lock sa be OPERABLE with:
- 4. Both doors dosed acept when the air lock Isbking used for wnor transit enaty and exit thnx4h the containmt. then au Se ow air lock door sl1 be dosed, and 5.5.14.d.2.a). b. An owair lockkrTe of S0.05 L. 1 P, 2 pilg.
5.5.14.b AEPEL1CAILl : MODES 1, 2. 3 and4. See ITS WIAh an ar lck Inoperable. restoe the al lock to OPERABLE MAia widt 24 hours or be in at lean HOT 3.6.2 STANDBY within the next 6 bours and In COLD SHUTDOWN within the folowing 30 bours. 4.6.1.3 Each eodatmem air lock Sa be denwstazed OPERABLEt s.5.14.a a. To acdI with 10 CFR 50 Appdlx J Opdea B and Regulazoq Guide 1.163, daed Szemr 1995, and
- b. At lent o per 6 as Aa attn.
I mIL by etrn oe oly door l bck drcan be opeaed l See ITS 3.6.2 1 COOK NUCLEAR P4LNT4JT 1 Fa. 314 44 AMEMIW 4UO 209 Page 18 of 69 Attachment 1, Volume 16, Rev. 1, Page 84 of 256
Attachment 1, Volume 16, Rev. 1, Page 85 of 256 ITS 5.5 ITS 314 UMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 314.7 PLANT SYSTEMS SURVEILLANC EREOUIREMENT Ad rpsdITS 5.5.9 generic .program statemen 4.7.5.1 The control room emergency ventilation system shall be demonstrated OPERABLE:
- a. Deleted See ITS 3.7.10
- b. At least once per 31 days on a STAGGERED TEST BASIS by initiating flow through HEPA filter and charcoal adsorber train and verifying that the train operates for at least Ior I L minutes. , (A 8 124J 5.5.9 c. At least once perg_60ih or () maintenance on the HEPA fier or charcoal adsorber housings, or (2) following painting, fire or chemical release in any ventilation zone conmunicatlng with the system. To while Isin operation op that could adversely affect the 5.5.9.b 1. Verifying that the charcoal adsorbers remove 2 99% of a halogenated filter bank or charcoal hydrocarbon refrigerant test gas when they are tested In-place in accordance with adsorber capability ANSI N510-1975 while operating the ventilation system at a flow rate of 6000 cfm+/- 10%.
5.5.9.a 2. Verifying that the HEPA filter banks remove 2 99% of the DOP when they are tested In-place In accordance with ANSI N510-1975 while operating the ventilation system at a flow rate of 6000 cfm i 10%. 5.5.9.c 3. Verifyinglwi-9S PI deta aftthat a laboratory analysis of a carbon sample from either at least one test canister or at least two carbon samples removed from one of the charcoal adsorbers shows a penetration of less than or equal to 1.0% radioactive methyl iodide when the sample is tested In accordance with ASTM D3803-1989. 30CC, 95% R.IlG The carbon samples not obtained from test canisters shall be prepared by either 5.5.9.0. a) Ernmyng one entire bed from a removed adsorber tray, rnixing the adsorbent thoroughly, and obtaining samples at least two inches In diameter and with a length equal to the thickness of thebed, or 5.5.9.c.2 b) EnvyIng a longitudinal sample from an adsorber tray. mriing the adsorbent thoroughly, and obtaining samples at least two Inches In diameter and with a length equal to the thickness of the bed. 5.5.9.a. 4. Verifying a system flow rate of 6000 cfh + 10% during system operation when 5.5.9.1b tested in accordance with ANSI N510-1975. 271 AMENDMENT 449. COOK NUCLEAR PLANT-UNIT I Page 314 7.20 314 7-20 APaENDMENT44g271 Page 19 of 69 Attachment 1, Volume 16, Rev. 1, Page 85 of 256
Attachment 1, Volume 16, Rev. 1, Page 86 of 256 ITS 0 ITS 5.5 34 LIMCG COND1MIONS FOR OPIZATION AND SURVEIANCE REQnU1M5 1f47 PLATSYSTEM 5.5.9 dAer evety 720 bows of chacbhnoddb oert by eit 5.5.9.c samp obuonAma Ust canls a papenln ofka dmuequn l to 1.0% (O*radeacdve meftfy lodkie wben tde sample Is wsued tot* ASlh4 D3WM31919, 3C 95% RA or 5.5.9.c 2. VaiftngWvl 31 da $Atufemeyal n a laboa*t1y analyis of at Uast two rao uipksshows apeatmdonoflsthanor eq tolfor zaoloctd methyl Iodide when th aamplea we testd In conlanco with ASfM D380919S. 30oCr 9 a; ad Om sampla we preared by ebdter. I 5.5.9.c.1 a) E6 ty Om nre d Ai= a removed idaobz try, mhlg f fe daimbent thoroughly and obtainin samples at leas two hnichs In disinerrand wlhakng~iheua to the h4--eso6ebed. r 5.5.9.c.2 b) Emptykn a lonwtuhna sample kme an sdawrbe bra. mixin do adri~t thcroughly. and abtaining samples at least two Inches In a and wth a kngth equal to the th s of the bed ma gtani arebo t sample. thie t be oosdidcOy Lbyaeo: /
*3 palfylnhatth adcxenuove41 tof2ab aloentd t
/ gas .hen ftey tested hinace In a#corac with ANSI 10-1975 whle th ntllado aysbm 1.4e ataflowratoOdOf0 *1 and b) Vefyls dtat te IA filtbanks rm t of dOe DOP wben twy wu lested IDacclo anc wit M5O-1975 WIOd INSI
/ope ng at asycm low uate cfit*10*
COOK NUCLEAR PLANr-UNIT 1 Pare314741 imENDME 257 Page 20 of 69 Attachment 1, Volume 16, Rev. 1, Page 86 of 256
Attachment 1, Volume 16, Rev. 1, Page 87 of 256 ITS 5.5 ITS 3/4 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4.7 PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) 5.5.9 e. At least once per by:
- 1. Verifying that the pressure drop across the combined HEPA filters and charcoal 5.5.9.d adsorber banks Is less than 6 Inches Water Gauge while operating the ventilation system at a flow rate of 6000cfm plus or minus 10%.
- 2. B. Vffifylbg that en a Safety Injetion Signal fromn Unit I, the system automatically operates In the pressurization/cleanup mode.
- b. Verifing that on a Safety Injection Signal from Unit 2, the system adz -fel - -ls n -ne^./
f SeeITS 3.3.7 and ITS 3.7.10 -
-- ubwmAuaAY n wc nt-ruruJMLLc3Ieup rnooc.
- 3. Verifying that the system maintains the control room envelope/pressure boundary at a positive pressure of greater than or equal to 1/16 Inch W. G. relative to the outside atmosphere at a system flow rate of 6000 cfm plus or minus 10%, with a SeeITS -
1 3.7.10 rnakcuzpairflowrateof< 1OO cfm. 5.5.9 fter each complete or partial replacement of a HEPA filter bank Iby verifying that the HEPA filter banks remove greater than or equal to 99% of the DOP when they are tested 5.5.9.a In-place in accordance with ANSI N510O1975 while operating the ventilation system at a now rate of 6000 cfm phis or minus 10%. 5.5.9 g. r each complete or partial replacement of a charcoal adsorber bankiby verifying that he charcoal adsorbers remove greater than or equal to 99% of a halogenated hydrocarbon 5.5.9.b refrigerant test gas when they are tested In-place In accordance with ANSI NSIO-1975 while operating the ventilation system at a flow rate of 6000 cfm plus or inus 10%. The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the VFTP test Frequencies. I COOK NUCLEAR PLANT-UNIT I Page 314 7-22 AMENDMENT0, U44,-18,271 Page 21 of 69 Attachment 1, Volume 16, Rev. 1, Page 87 of 256
Attachment 1, Volume 16, Rev. 1, Page 88 of 256 ITS 5.5 ITS uAr msiLMf 3.7.4.1 Two 1ndapsndaw irfTvmatilatian syst..a sabauc afr filter turans shal'be ICIALIU. See ITS] Apiu.ra MO 1. 2. and 4. . 1th *neo1be veon~latten *yseur exhaeus &t filter trea lqqerablai. retate the tenprerble train to DIEUAh status wvitb 7 days or be La at laar.MAT ItvINCt:lfqr STAIDIY within the next 6 hours and In COL SXitog within the following la wur,,i. 4 aAdd proposed ITS 5.5.9 generic program statement I A9
&.7.I.1 lath U3T V601l:1acon system "haul1air filter train shall, be deeaestratee OPUASLI:
UTC TI USAIM b7 Watlatt . 3.7.1
- a. At leuut cn0 per 31 days an a STI free the aenmio roan. flew thwq4A the RUA filter nehatea1 adgorbar train. end vertfyi4~that the train oepracs. fer at leant 13 miutvtei.
- 3) 2 (tA.. e
- b. At least me) per or1usteul tt a1intenanCe 5.5.9 on the XIA filter er ha l adseber holw oe r (2) follewing paintIng. fi. er shemual release ina ay vencif a coini acng with the sye teemb:
t O ht"tt
- heI Isn operation that
.ecould adversely affet the 5.5.9.b 2. Vrifying that 2. the Vlf charel ut adsarhers de 1egto ree &I 2 na e Ilter bank or charcoal dsorber capability balegoeated hydrienarbims refrigesant teat as Ubie& they aft tsted in-plase La aceordae with. A2t 1510-1USA w-i '
eperACIag thes westiLAtIem dytde St a fle r at .1 15,000 stU e 10. 5.5.9.a 3. Veriyng that the RUA fiter h_ re" 2I l o the D301 when bys-r tasted in-,luse In amerade. with AUX 110.-1110
'Ella eperacing the VmamL1uLOU system at a flw late f W000
*efm t leO.
- 0. C. COOZ . MQT L 3/4 7-.2 _ _ahe Ue.124 Page 22 of 69 Attachment 1, Volume 16, Rev. 1, Page 88 of 256
Attachment 1, Volume 16, Rev. 1, Page 89 of 256 ITS 5.5 ITS 314 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS VAI PLANT wrTFUMq SURVEII1ANCE REOUIREMNS (Continued) 5.5.9.c 4. Verifying vitbrn 31 rlis srrFFmoval that a laboratory analysis of a carbon sample from either at least one test canister or at least two carbon samples removed from one of the charcoal adsorbers shows a penetration of less than or equal to 5% for radioactive methyl iodide when the sample is tested in accordance with ASTM D3803-1989, 306C, 95% RH, and 2 45.5 fpm face velocity. The carbon samples not obtained from test canisters shall be prepared by either 5.5.9.c.1 a) Emptying one entire bed from a removed adsorber tray, mixing the adsorbent thoroughly, and obtaining samples at least two inches in diameter and with a length equal to the thickness of the bed, or 5.5.9.c.2 b) Emptying a longitudinal sample from an adsorber tray. mixing the adsorbent thoroughly, and obtaining samples at least two Inches In diameter and with a length equal to the thickness of the bed. Subsequent o reinstalling the adrr tray used for obt ning the carbon sample, the tem shall be demons~ated OPERABLE by verifying that the charcoal q(sorbers remove than or equal to 996 of a halogenated hydrocar~n refrigerant test gas Den they are tested in-plae in ccordanc with L4-- ANSI N}1I0-1980 while operad* the ventilation system a flow rate of 25,000 cfm plus or minus 10%. / 5.5.9.a, 5. Verifyng a system flow rate of 2S 000 cfm plus or minus 10%o during systemn 5.5.9.b operation when tested In accordance with ANSI NS10-1980. 5.5.9 c. After every 720 hours of charcoal adsorber operation by either
- 1. Verifyingiwitltn 31 ddys *fenovallthat a laboratory analysis or a cartoon 5.5.9.c sample obtained from a test canister shows a penetration of less than or equal to 5% for radioactive methyl iodide when the sample is tested in accordance with ASTM D3803-199, 300C. 95% R.H.. and 45.5 fpm face velocity; or I
---e 5.5.9.c 2. Verifyinglwitfin 31 dais afterAemovallthat laboratory analyses of at kast twu carbon samples shows a penetration of lcss than or equal to 5% for radioactive methyl iodide when the samples ae tested in accordance with ASTM D3803-1989, 301C 95% R.H.. and 2 45.5 fpm face velocity and the samples are
---e prepared by either.
5.5.9.c.1 a) Emptying one entire bed from a removed adsorber tray, mixing the adsorbent thoroughly, and obtaining samples at kast two inches in diameter and with a length equal to the thickness of the bed, or COOK NUCLEAR PLANT-UNIT I Page 3/47-24 AMENDMENT 124, i,257 Page 23 of 69 Attachment 1, Volume 16, Rev. 1, Page 89 of 256
Attachment 1, Volume 16, Rev. 1, Page 90 of 256 ITS 5.5 ITS
-PLANT lsnTc 7
'RVlILLANCA REOUTRmIgM (Continued -
5.5.9.c.2 b) mptyLng a longitudinal sample frou an ad*orber tray. mixing the adsorbent thoroughly. and obtaining samples at least two inches in dLameter and with a length equal to the thickness of the bed. Subsequan£ to riLnstalllng dhe r tray i d for obtain-
*dsorb:
ing the arbon sample, the Ayetea hall be deinstraced
=PECABI4 by also vrLfyingL that groea refri than or equal to I9*
charcoal Aorbers remove of a halogenae# hydrocarbon arent test gas whi they ars t sted i accot anco with ANSI NSaO 1980 whleoper ing th ventilation in i-plae -0G 5.5.9 d. floy rate At least once per DIfntonhs by: 2-00 ei la%!
-0G 5.5.9.d
- 1. Verifying that the pressure drop across the combined HEPA filters and charcoal adeorber banfIs is less than 6 inches Vater Cauge hile operating the ventilation system at a flow rate of 25,000 cfm plus or minus 10%.
- 2. Deleted.
- 3. Verifying that the standby fan itarts automaticlly on a Containment Pressure--igh-gHih Signal and dLre:ts its & See TS exhaust flow through the hEPA filters and charcoal adsorbs 3.7.12 banks on a Containment Pressure-HiLgh-HLgh Signal.
559 ter A each complete or partial replacement of HEPA filter band by eryLn C 01 KM 21lter banU reoe greater than or equal to 5.5.9.a 99% of thehoD when they are tested in-place in accordance with ANSI Nl0-19t0 vwhle operating the ventilation system at a flow rate of 25 000 cfn plus ot minus 10%. 5.5.9 f. tar each cooplet or partial replacement offa-charcoal adsorber bantk byrt x th eMCh charcoal dadorbera remoe greater than or equal 5.5.9.b to 99% of a halogenated hydrocarbon refrigerant test Sax when they are tosted lnaplace in accordance with ANSI I510-1980 while operating the tilatLon systea at a flow rate of 25,000 cfn plus or mids 10%. The provisions of SR 3.0.2 and SR 3.0.3 are z .applicable to the VFTP test Frequencies. COOK tUCLEAR PL.NT - UMSIT 1 3/4 7-25 AMMENT NNO.q, IUZ. Page 24 of 69 Attachment 1, Volume 16, Rev. 1, Page 90 of 256
Attachment 1, Volume 16, Rev. 1, Page 91 of 256 ITS 5.5 ITS 3t4 LIMITING CONDIIIONS FOR OPERATION AND SURVEIllANCE REQUIREMNrs 314.J ELECTRICAL POWER SYSTEMS SURVEIUNC E gOlthREMENTS C n 4.8.1.1.2 Each diesel erator shall be demonstrated OPERABLE: See ITS
- a. In accordance with th frequency specifled In Table 4.J.1 on aSTAGGERED TEST rIS 3.8.3 )
BASIS by: See UTSI It. YeriygthNd la n theLdank 3.8.1 12. I -- Verfiia
--- . - thate h- level In the fuel storse tank.
I
. . See ITS
--- 3.83 J
- 3. Verifying thatefhel transfer punp can be suted and tbeait uanseai fuel from the sotage system to the day tank,
- 4. Verifying that the diesel utarts fom standby conditions and achlieves in less thn orequalto 10 second, voltage - 4160i 420 V. and frequey - 60tl.2 Hz.
-k--f See ITS 3.8.1)
- 5. Verifyinl the diesel is saynclroeied and baled and operates for greater than or eqalo60miues ata of 350 kw*. and I
- 6. Verifying tha the diesel penetor if aligned to provide standby power to the " 3.8.1 and See ITS -
I associated cmaty bsses.i ITS 38.3 lb. By renitiving accumulated Visa"'r i See IS I) From tS day Lee at lu Oec pa 31 Gays n ar ec1 occaion wfll iMO 3.8.1 J diesel Isoverated fIr sieatr tbas I hour. and
- 12) Frovn the storag tankat leas once per 31 See ITS 3.8.3 )
5.5.11I.a 5.5.11.8
- The diesel rPatoc start (10 seconds) fhm standby conditions shall be performed at least oce I
*per 184 days Inthese surveilliace tsts. AJU otherengine sarts for the purpose of dis sureillance testing and compesazoty action may be at reduced ceratlon rae as recommended by the See ITS manufacturer so that mecanlcal stress and wear on the diesel engine a mInimied. 3.8.1 M meto od =entsdonotlnvaldate thls test.
- Te acti*ns to be ta= should any of thepropertIes be found outse of specified s ITS )
I
- dfiedi th a. 3.8,3 COOK NUCLEAR PLANT-UNTI 1 Pap 314 MS3 AM EN I 435, 207 Page 25 of 69 Attachment 1, Volume 16, Rev. 1, Page 91 of 256
Attachment 1, Volume 16, Rev. 1, Page 92 of 256 ITS 5.5 ITS 314 LS;CTING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 314.1 ELECTRICAL POWER SYSTEMS lwithin limits SUREILACE E fMEENT (oa I/ 5.5.11 .a.2 I less thafor ) to 4.1 cealt it 40-rltiven y. saybolt Vin I or 40.1 supplier*ceti a tf tywanmdo orb edbycomparisonwith .- / 5.5.11 .a.2 b) A Bash pointet o or 1(ta tt liis 5.5.11 .a.1 2) By vtdfla cewlhbctVdtnD~80 ~Prr )ghn irni to addlxte newvftud to the stanh. that the IC~lhusibraa
/
ht tYuof r d cto: 30 deat but "l. H Pt o401/ ld nq44 e.87 =or iwlto i ed. ce66spdicg v 1°t sotrW Wb =t compar to pgc's _
/ fnew fue oil, ote than those 5.5.1 1.a.3 3) ByverifwsIn =Wdfnn With the tclfled In MlD1bZadrc ddressed In to addInS hew ul to tbe st rat te ismn thuat cltrIihe Spedfication 5.5.1aabove, sppearze with proper color.
- 4) )Ii By verifiit within 31 dxyslof obtsini theatnll~ tbst 5.5.1 1.b l ifd in TA I of ASM D97S41 1are within the appropriate Iimetnh eIsta In Qyace wws=A3 a w5^1CC u y iyts lot low n^
5.5.1 1.c d. At least owre per 31 !AY
- H=
* ,ole GC oifc al tnnn fthcd=- ths in IiZD VM A4176 1Msad Yelfyiq that total pauilaz coommnation Is less than 10 nml/ter when teted In accordnee with ASTM D227tU3 Method See ITS C. At least once per 13 rnths, duiM shutdown, by: 1 3.8.3J
- 1. Subjectn fte diesd engtne to an In ectiottIn sc erdac with procedures - See ITS prepared In coejustdon with Its mfacturer's recommendations for thbis cla t 3.8.1 J of stnby service.
Th provisios ofS 3.0.2 and SR 3.0.3 are applicable to Diesel Fuel Oil Testing [he Programn test Frequencies. Mbe e to be take shul any of the pm fe bectm outidde of the Vacimfe kinthe dedarc See rTS
, . . 33.8.3 COOK NUCLEAR PLANT4MNIT 1 Pap 314 14 AMENDME3NT 125 Page 26 of 69 Attachment 1, Volume 16, Rev. 1, Page 92 of 256
Attachment 1, Volume 16, Rev. 1, Page 93 of 256 ITS 5.5 ITS 314 LIMTING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4.9 REFUELING OPERATIONS STORAGE POOL VENTILATION SYSTEM** LIMITING CONDITION FOR OPERATION 3.9.12 The spent fuel storage pool exhaust ventilation system shall be OPERABLE. SreeITS 3.7.13 J APPLICABILITY: Whenever irradiated fuel Is In the storage pool. ACTION
- a. With no fuel storage pool exhaust ventilation system OPERABLE, suspend all operations Involving movement of fuel within the storage pool or crane operation with loads over the storage pool until at least one spent fuel storage pool exhaust ventilation system Is restored to OPERABLE status.*
- b. The provisions of Specification 3.0.3 are not aplicable.
SUJRVFILL&_ACF 1RFAn1rRPu;:MTrq Add proposed ITS 5.5.9 generic program _ tA 4.9.12 The above required fuel storage pool ventilation system shall be demonstrated OPERABLE: See ITS
- a. At least once per 31 days by Initiating flow through the HEPA filter and charcoal adsorber train 3.7.13J and verifying that the train operates for at least 15 munaitn.
5.5.9 b. At least once per[a t charcoal adsorber housings, or (2) following painting, fire or chemical release in any ventilation zone communicating with the sy by: He ItIs In operation that
- 1. Deleted t ~~~adversely affect the fiher akoA8 charcoal adsorber capabli 5.5.9.b 2. Verifying that the charcoal adsorbers remove 2 99% of a halogenated hydrocarbon refrigerant test gas when they are tested in-place in accordance with ANSI N510-1980 while operating the exhaust ventilation system at a flow rate of 30,000 cfm +/- 10%.
- The crane bay roll-up door and the south door of the auxiliary building crane bay may be opened under administrative control during movement of fuel within the storage pool or crane operation with loads over the storage pool.
enShared system with D.C. COOK - UNIT 2. a sU See rrS 3.7.13 3 COOK NUCLEAR PLANT-UNIT I Page 314 9-13 AMENDENT4z4,A43, 29, 281 I Page 27 of 69 Attachment 1, Volume 16, Rev. 1, Page 93 of 256
Attachment 1, Volume 16, Rev. 1, Page 94 of 256 ITS 5.5 ITS 0 314 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3149 REFUELING OPERATIONS SURVEI'LLANCE REOUDDENTS-{Cotined 5.5.9.a 3. Verifying that the HEPA filter banks remove greater than or equal to 99% of the DOP when they are tested in-place In accordance with ANSI NS10-1980 while operating the exhaust ventilation system at a flow rate of 30.000 cfm plus or minus 10%. 5.5.9.c 4. Verifying withn 31 ds a fcmoyat aa tory analysis of a carbon sample from either at least one test canister or at least two carbon samples removed from one of the charcoal adsorbcrs shows a penetration of less than or equal to 5% for radioactive methyl iodide when the sample Is tested in accordance with ASTM D380341989, 30°C, 95% R.H., and 2 46.8 Ipm face velocity. The carbon samples not obtained from test canisters shall be prepared by either 5.5.9.c.1 (a) Emptying one entire bed from a removed adsorber tray, mixing the adsorbent thoroughly, and obtaining samples at least two Inches in diameter and with a length equal to the thickness of the bed, or 5.5.9.c.2 (b) Emptying a longitudinal sample from an adsorber tray, mixing the adsorbent thoroughly, and obtaining samples at least two inches in diameter and with a length equal to the thickness of the bed. Subsequent t installing the adsober used for obtaining the car&on sample, the system shall be de tnsrated OPERABLE by0also verifying that the ch od adsorber remove Am_ greater th or equal to 99% of a halo nated hydrocarbon refrige nt test gas when they are 1.4 tested in- lace In accordance with SI NSIO-1980 while ng the ventilation system at a flo rate of 30,000cfm c lus inus10%. 5.5.9.a, 5. Verifying a system flow rate of 30,000 cfm plus or minus 10% during system operation when 5.5.9.b tested in accordance with ANSI NS510-1980. 5.5.9 c. After every 720 hours of charcoal adsorber operation by either. 5.5.9.c I. Verifying 3 a t a laboratory nalysis ofa carbonsample obtained (i) from a test canister ows a penetration of less than or equa to 5% for radioactive methyl iodide when the sample is tested in occordance with ASTM D3803-1989. 3D°C. 95% RH and 2 46.8 fpm face velocity; or COOK NUCLEAR PLANT-UNIT 1 Page 34 9-14 AMENDMENT*,4,", 257 I Page 28 of 69 Attachment 1, Volume 16, Rev. 1, Page 94 of 256
Attachment 1, Volume 16, Rev. 1, Page 95 of 256 ITS 5.5 ITS
*314* LIPMGNG CONDMONS FOR OPRATnON AND SURVEILLANCE REQUIEMF 3Y43 REUEN OPEAONS I
=(C'n*i 5.5.9.c 2. VefIng withl 31 4K5% mlderllemovha of at lest two cabou samples showa penetxtion of less than er equal to 5%for radioactive methyl odide when the aMples are la wnand with ASTM D3303-199. 30 95% RJL and 24&S fpm face velocity and the samples ae prepared by either:
5.5.9.c.1 (a) Emptying oe entire bed from. removed adisorber ry. mixn the adsorbent t u y mid obtaining sample at les two Incwe In diameter and with a length equal to the thickness ofthe bed, or 5.5.9.c.2 (b) Emptying a loughtmlnal sample from an adsorber tray, mixing the adsorbent thorodghly, and obtainin samples at lests two Inches In diameter and wth a length equal to the thicknes of the bed. Subsequent io tn the ad ber for obtling thn cumplb the systemi
-0e shall be daqatrd OEB y vefing U ; tha adsoebe remov t cr e 99% test gas when they "a lested In accordance iso the ventilation system I at a f rf30.000mplorl0%.
- 5.5.9 d. At least once pesf t o2
- 1. Verlaflng tht fte pressure drop aros the combined HEPA filtersand dcharoal adsorber 5.5.9.d banks Is les than or equal to 6 inches Water Gawg while operating the exhaut ventilation system a a flow rate of 30000 cfn phus or minus 10%.
- 2. Deleted.
- 3. Vrifying that on a hligh-adlaulon sinl, the system automatically directs Its exhas flow through the cacoal adscrber bank and automatically shuts down the storage pool ventilation system suply fns.L
- 4. Verfidng tht the exhaust ventilation system maintains th spent fuel storage pool wea at a See ITS negative pressure of pester than or equl to 13 inches Water Gauge relative to the outside - ( 3.7.13J atmosphere during system operation.
COOICNUCLEARPLANT.UNrrl Page3 M 9415 I Page 29 of 69 Attachment 1, Volume 16, Rev. 1, Page 95 of 256
Attachment 1, Volume 16, Rev. 1, Page 96 of 256 ITS 5.5 ITS eiwt C Ia Otaf?~ gmettuG QMetmis 5.5.9 -_ . gfter each cenplate or partial rteplacemnt of a EVA filter beahn,* scITUY74 thal;t tbaik tea" k I a the Da iae o 5.5.9.a r are teoetaA in-plan in aceordaese tthAnt 1310-1980 while operating the veatilaieuo sysaes at a flow :ae .of 30000 fs t lo,. 559 r oach amplato or partial zqlecm %t of a ehastuml a£notbor 5.59 wuifying shat the charcoeal ads rba teasg Z 199 of a egnasted bydrocarhon refrigent teast Sa who they at* testce 5.5.9.b lin-pleo La accordanse with ANSI mI0-191o VWo serating the lanslalou qy s at a flow rte of 30.000 ets t 10 e provisions o tsR 3.0.2 and SR 3.0.3 are applicable to the VFTP test Frequencies.
- 0. C. COs . MM I 3/5. 9416 -MWAM xlo 124 Page 30 of 69 Attachment 1, Volume 16, Rev. 1, Page 96 of 256
Attachment 1, Volume 16, Rev. 1, Page 97 of 256 ITS 5.5 ITS 3/4 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 314.11 RADIOACTIVE EFFLUENTS II.IJID HOLDUP TANKS LIMMNG CONDITION FOR OPERATION Add proposed ITS 5.5.10 generic program statement 5.5.10, 3.11.1 The quantity of radctiv material contained in each of the foil tanks shall be limited to=les 5.5.10.c than or eqal to 10 curies. excludig tritumn and dissolved or entrained noble Sass.
- a. Outside temporary tanks.
APPLICABILTMY At all tines. ( ACMON:
/a. With the quantiy of ye material in any of the hove listed tanks exceeding thee above limit, wialldelay supend al additions of raoactive material to the tank and within 48 hours the tank COntents to within the t.
- b. The provisions of ifction 3.0.3 are notaaplica .
SURVEILLANCE REQUIREMENTS 5.5.10.c 4.11.1 The quantity of radioactive material contained in each of the above listed tanks shall be determined to be within the above limit by anal a resentatv the tank's co a: s Tohepeays when radiogc~e materials are bciaf added to the ta. l 4 [The provisions o SR 3.0.2 and SR 3.0.3 are applicable to the Storage nTank Radioactiity Monitoring Program Surveillance Frequencles. IAll 5.5.10.c
- Tanks included in this Specification are those outdoor tanks that are not surrounded by liers, dikes, or walls capable of holding the tank contents and that do not have tank overflows and surrounding area drains connected to the liquId radwaste treatment system.
COOK NUCLEAR PLANT-UNIT 1 Page 314 11-1 AMNDMAENT 69,154,489 281 I I Page 31 of 69 Attachment 1, Volume 16, Rev. 1, Page 97 of 256
Attachment 1, Volume 16, Rev. 1, Page 98 of 256 ITS 5.5 ITS 3/4 LMIITING CONDITIONS FOR OPERATION AND SURVEILTANCE REQUIREMENTS 3/4.11 RADIOACTIVE EFFLUENTS I 3/4.11.2 GASEOUS EFFLUENTS EXPLOSIVE GAS MIUXTURE MITING CONDmON FOR OPERATION Add proposed ITS 5.5.10 generic program statement 5.5.10, 3.11.2.1 Th concentration of oxygen inthe waste gas boldup system shallbbemtimite s rua 5.5.10.a 3% byvolus ilfthehydrgeninthe is reaterthanor to4 vo APPUCABILUTY: At all times. ACTION: With the concentration o in the waste gas holdup stem greater than 3% by Ioxygen volume but less than or hydrogen. restore the to 4% by volume and containin greater than or equal to 4% nof oxygen to less than hydrogen concentration less than 4% within 96 hours. equal to 3% or reduce the
-0e
- b. With the conccmatio ofoxygen in the waste gas holdup stemor tank greater than 4%
by volume and grea than 4% hydrogen by volume wi t delay suspend all additions of waste gases to the tem or tank and reduce the c of oxygen to less than or equal to 3% or the ceron of hydrogen to less or equal to 4% within 96 hours In the system or C. The provisions of 3.0.3 SURVEILLANCE REQUIREMENTS 5.5.10.a 4.11.2.1 The concentration of oxygen In the waste gas holdup system shall be determined to within the above limits by continuously monitoring the waste gases In the waste gas ho! rystem e LA ) oxygsh monito required OeEWdFEby Table 3.3-12 of ipcaton 3.33 The provisions of SR 3.0.2 and SR 3.0.3 are appicable to the Explosive Gas Radioactivity Monitoring Program Surveillance Frequencies. COOK NUCLEAR PLANT-UNIT 1 Page 314 11-2 AMENDNENT n2, 54,489, 281 l I Page 32 of 69 Attachment 1, Volume 16, Rev. 1, Page 98 of 256
Attachment 1, Volume 16, Rev. 1, Page 99 of 256 ITS 5.5 ITS 3/4 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4.11 RADIOACTIVE EFFLUENTS GAS STORArE TANICS LWIMTING CC INDmITON FOR OPERATION Add proposed ITS 5.5.10 generic program statement 5.5.10. 3.11.2.2 The quantity of radioactivity contained in each gas stora eank sl r~ 5.5.1 0.b noble gus (conidered a- Xe- 1
/
I _ APPLICABiLITY: At all times./l With the quantity of raterial maoactivein any gas stora tank exceeding the-above limit, without delay 48 hours reduce the t all additions of radioactive ccontents to within the limit. tonal to the tank and within
-0 The provisions of Sc iflcaton 3.0.3 are not applicable I SURVEILLANCE REQUIREMENTS 5.5.10.b 4.11.2.2 The quantity of radioactive material contained In each gas storage tank shall be determined to be within the above limit Wav once per 7 days w enever rad dactive materials are ad to the I tanc and at least once per 4 hours during primary coolant a stem degasin eratio The provisions of SR 3.0.2 and SR 3.0.3 are 4 applicable to the Storage Tank Radioactivity Monitoring Program Surveillance Frequencies.
T-Q COOK NUCLEAR PLANT - UNIT 1 Page 3/4 11.3 AMENDMENT 69, 14, 3, 3l, 281 Ul9, I Page 33 of 69 Attachment 1, Volume 16, Rev. 1, Page 99 of 256
Attachment 1, Volume 16, Rev. 1, Page 100 of 256 ITS 5.5 ITS
.0 ADMN RAT CONTROLS 6J.1 Wrifnte pr-c s be estbtishe4fhlend mad tmintahed coverig the acb2tes d-ene below.
AL The applicable prcedes recozmended imAppan& "'A of Regulatay Guide 133, Rev. 2, Fcbry 1971. . ITS 5.4
- b. Deeted C. Dtd
- d. PROCEM CONTROLPRoaA WeFI-do
- e. OFS1I DOSE CALCULA17ON MAUALimpleazntaiom.
I . Qualiy Anurvce Pro=i for ee nd e omncml soni using the dance in ReplatgyGoide 1.21,1R. 1. Im 1974, andRMltoryGuide 4.1,Rev. 1,Apra 1975. S. Cocki or.Txaziw 1 pr m povicies controls to tnck the 5.5.4 Section 4.1. cdic and turient ec to we thS cw re rintained with e laimn.
- h. Fire proteto Proto inieent_
6.S2 Eac procedu and .ad policy of Specification 68.1 above; and changs therto, ih eiistratve ftenoazy changes hail be reviewed prixe to uscet forth in QuAlity Assurance P Description App xt CX Setio 6.5. -{I See ITS 1 5.4 J 6.U3 Ded t* COOK NUCLEAR PLANT-UNIT 1 Put " AMZDMrV,U4, 194,
.S0401436,261 Page 34 of 69 Attachment 1, Volume 16, Rev. 1, Page 100 of 256
Attachment 1, Volume 16, Rev. 1, Page 101 of 256 ITS 5.5 ITS 6.0 ADMINSTRAT1VE CONTROLS 6.13 PROCESS CONTROL PROGRAM (PCP) 6.13.1 Changes to the PCP:
- a. Shall be documented and records of reviews performed shall be retained as required by the Quality Assurance Program Description. Appendix C, Section 6.10.2.n. This documentation shall contain:
- 1. Sufficient information to support the change together with the appropriate analyses or evaluations Justifying the change(s) and
- 2. A determination that the change will maintain the overall conformance of the solidified See CTS waste product to existing requirements of Federal, State. or other applicable regulations. 6.0 J
- b. Shall become effective after review and acceptance by the PORC and the approval of the Plant Manager.
6.14 OFFSITE DOSE CALCULATION MANUAL (ODCM) 5.5.1.c 6.14.1 Changes to the ODCM: 5.5.1 .c.1 a. Shall be documented and records of reviews performed shall be retained as~euired theali I m Prr~iogmm esrikin.Amndix G-Scction M.U.Z.nir Tis documentation shall contain: 5.5.1.c.1.a) I. Sufficient information to support the change together with the appropriate analyses or evaluations justifying the change(s) and 5.5.1 .c.1 .b) 2. A determination that the change will maintain the level of radioactive effluent control pursuant to 10 CFR 20.1302. 40 CFR Part 190, 10 CFR 50.36a, and AppendixI to 10 CFR Part 50 and not adversely impact the accuracy or reliability of effluent, dose, or setpoint calculations. 5.5.1.c.2 b. Shall become effective after! .nd aet4ance hylic POR CviGr land the approval of therlaaat I tanager. 5.5.1 .c.3 c. Shall be submitted to the Commission in the form of a complete, legible copy of the entire ODCM NI-11 as a part of or concurrent with the Annual Radioactive Effluent Release Report for the period of the report in which any change to the ODCM was made. Each change shall be identified by markings in the margin of the affected pages, clearly indicating the area of the page that was changed, and shall indicate the date (e.g., monthlyear) the change was implemented.
- _iAdd I proposed ITS 5.5.13 and ITS 5.5.15 l I COOK NUCLEAR PLANT-UNIT I Page 6-15 AMENDMENT 7, 454,489, 236, 24, 279 Page 35 of 69 Attachment 1, Volume 16, Rev. 1, Page 101 of 256
Attachment 1, Volume 16, Rev. 1, Page 102 of 256 ITS 5.5 ITS ica AMS1Irhfl-q Aft-VUm rV'nrOl1r a ; . . PRQU^ ANDM O}(Cnnd 5.5 6.A The wing progra ae be utahUlidViemented, and mintaind. 5.5.3 a. 5.5.3
- prognam saD be prided conforning with 10 CFR 5026a for the control of A
*adioactive efiluents and for maintabnig the doe to B OF TEE PUBLIC from radioactive diuente a. low es reasonably achievable. Mea program (1) shall be contained in the ODC (2) halld be imnplems by bperatlng prowdures. and (8) ash Include remedial actions to be taken whenever the program limit are exceeded.
nshall.ncude thefollowingelements: 5.5.3.a 1) .Lfimtations en the operability of radioactive liquid and gaseous moion inshuentation incuding surwelln tests and setpint determination in rdance with the methodolog in the ODCKM, 5.5.3.b 2) Timtation on the concentretions of radioactive matirlI released In liquid efluente to UNRESTRICED AREAS conforming to 10 CFR 20.1001.
. I 20.2402, AppendixB, Tale 2, Column 2, 5.5.3.c 9). Monitoring, sampling, and anaulys of radioactive liquid and gaseous effiuentspruant to 10CFR201302 and with the methodology and I parameter i the ODCU, 5.5.3.d 4) Limitations on the annual and quarterly doe or dose commitment to a MEBER OF THE PUBLIC from radioactive materils in liquid emuent.
releaged from sub unit to UNRES=RIC ED AREAS cnforming to Appendix I-
- ItolOCPRPartgo, 9 5.5.3.e 5) Determ d. onataumuladrv and prqected doe *contihbu= firo radloacv efuent. for the current clendar quarter and current calendar year in a.crance with the mnetodiof and parameters in the ODCM at leat every 31 days, 6). linitation an the operablit and uee of the liquid and gaeous effhuent 5.5.3.f treatment systems to ensure that the apprprat portins of thlee systs ar used to reduce release of radioactivity when the pWq e do in a 31.
day period would exceed 2 percent of the guidelines for the anusl doe r dose commitment onforming to Appendix I to 10 CFR Part 50. COOK NUCLEAR PLANlTUN 2 . page 6-7 ANMENMENT479, ate3 226 Page 36 of 69 Attachment 1, Volume 16, Rev. 1, Page 102 of 256
Attachment 1, Volume 16, Rev. 1, Page 103 of 256 ITS 5.5 ITS 0 6.0 ADMMSTRATVZ CONTROLS 5.5.3.g 7) .Inltaatlons on the dose rate resulting from radiative material released in geous efucats to amu beyond the IM BOUNDARY shal be limited to the fs awb. a) Per noble gaae Lea than or equal to a dos rate of 500 znrem/year to the total body and leamthan or equal to a doase rate of 8000 inremxL r to the akn, nd - b) Por Iodine-l18, Iodilne-83, tridum, and for all redilnlaides In partculsate form with half-lives greater than 8 days: 1en them or equal to a dose rate of 1500 mrvon/yar to any organ. 5.5.3.h 8) LindtatIons on the an11 and qurtrly sir doses reauling from noble gases released in gaseous euezits from each unit to ae beyond the STE BOUNDARY conroring to Appendix I to 10 CYR Part 50, 5.5.3.1 9) Tlimitago on the annual and quartery doses to a MEMB OF TM PUBLIC from lodine-Iai, Iodhw-W, tridum, and all radionudides in p form with half-lives greater than B days in gaseous eluents released from ach unit to areas beyond the Sm BOUNDARY confoming to Appeaft Ito l CFR Pat 50, and
-2) 5.5.3.J 10) LImitations on the anrual doe or dose commitment to any MEMBER OF THE PUBLIC due to releases of radioactivity and to radiaion fm uranu m ul ccle og d to 40 CFB Part 190. The provisions of SR 3.0.2 and SR 3.0.3 are
- b. applicable to the Radioadive Efluent Con I -/ Poram Surveillance Frequend-es. 1 sal be ided to oth raainand iid n the othe Plant. Tpoa provIde measurements of In tbhe higet potn eoaer pathwxt and ( ) ion of the theeuentmntoinnd aof modeling ofe l eposre ha. he p am ce of Appendix Ito 0Cs lonitorig, smpling, be co rt ed a in the 0 ad(8) 60, Inlude the f and reporting ofas (2) coorm to the in the. environment ti with the and prm in the ODC1, A Lnd Use Casuto toatangeu in the areafat and byond the BTME BOUNDARY Idetifed and tht to the monitoring prgram aremadeif bythe ranof't ad Paidpdation in Compario to ensure that k4epndent ciha e the uz and thmeasurements of rdio environmental mple a performed as part of the quity program for en entl monitring.
COOK NUCLEAR PLANT.UNlT 2 Pap 64 , AMENDMENT5,3, 226 Page 37 of 69 Attachment 1, Volume 16, Rev. 1, Page 103 of 256
Attachment 1, Volume 16, Rev. 1, Page 104 of 256 ITS 5.5 6.0 ADMINISTRATIVE CONTROLS PROCEDURES AND PROGRAMS (Continued) 5.5.12 6.8.5 Technical Specification Bases Control Proeram This program provides a means for processing changei to the Bases of these Technical Specifications. 5.5.12.a a. Changes to the Bases of the Technical Specification shall be made under appropriate administrative controls and reviews. 5.5.12.b b. Licensees may make changes to Bases without prior Nuclear Regulatory Commission approval provided the changes do not require either of the following:
- 1. . A change in the Technical Specification incorporated in the license or
- 2. A change to the Updated Final Safety Analysis Report of Bases that requires Nuclear Regulatory Commission approval pursuant to 10 CFR 50.59.
5.5.12.c C. The Bases Control Program shall contain provisions to ensure that the Bases are maintained consistent with the Updated Final Safety Analysis Report. 5.5.12.d d. Proposed changes that meet the criteria of Specification 6.8.5.b above shall be reviewed and approved by the Nuclear Regulatory Commission prior to implementation. Changes to the Bases implemented without prior Nuclear Regulatory Commission approval shall be provided to the Nuclear Regulatory Comnmission on a frequency consistent with 10 CFR 50.71(e). 6.9 REPORTING REQUIREMENTS ROUTINE REPORTS 6.9.1 In addition to the applicable reporting requirements of Title 10, Code of Federal Regulations, the following reports shall be submitted to the Regional Administrator unless otherwise noted. STARTUP REPORT 6.9.1.1 A summary report of plant startup and power escalation testing shal be submitted following (1) - See rTs receipt of an operating license, (2) amendment to the license involving a planned increase in t 5.6 J power level, (3) installation of fuel that has a different design or has been manufactured by a different fuel supplier, and (4) mnodifications that may have significantly altered the nuclear, thermal, or hydraulic performance of the plant. 6.9.1.2 The startup report shall address each of the tests identified in the FSAR and shall include a description of the measured values of the operating conditions or characteristics obtained during the test program and a comparison of these values with design predictions and specifications. Any corrective actions that were required to obtain satisfactory operation shall also be described. Any additional specific details required in license tonditions based on other commitments shall be included in this report. COOK NUCLEAR PLANT-UNIT 2 Page 6.9 AhMENDMENT 4, iM, -S, 1, 265 I Page 38 of 69 Attachment 1, Volume 16, Rev. 1, Page 104 of 256
Attachment 1, Volume 16, Rev. 1, Page 105 of 256 ITS 5.5
.- 2 (t) Deleted by Amendment No. 261 (t) Deleted by Amendment 63 (u) Deleted by Amendment No. 261 5.5.8 (v) Secondary Water Chemistry Monitoring Program The licensee shall implement a secondary waterrchemistryymonitoringg program to inhibit steam generator tube degradation; Thisporm'11 .
I e"ed in the staton chemisfF anual anq~s-alicue 5.5.8.a 1. Identification of a sampling schedule for the critical parameters and control points for these parameters; 5.5.8.b 2. Identification of the procedures used to measure the values of the critical parameters; 5.5.8.c 3. Identification of process sampling points; 5.5.8.d 4. Procedure for the recording and management of data; 5.5.8.e 5. Procedures defining corrective actions for off control point chemistry conditions; and 5.5.8.f 6. A procedure Identifying (a) the authority responsible for the interpretation of the data, and (b) the sequence and timing of administrative events required to initiate corrective actions. (w) Deleted by Amendment No. 261 (x) Deleted by Amendment No. 261 (y) Deleted by Amendment No. 261 (z) The 72-hour allowed outage time of Technical Specification 3.8.1.1 Action b which was entered at 0923. on December 7, 2003, may be extended one time by an additional 72 hours to complete repair and testing of the 2 AB diesel generator. D. Physical Protection The licensee shall fully implement and maintain in effect all provisions of the Commission-approved physical security, guard training and qualification, and safeguards contingency plans including amendments made pursuant to provisions of the Miscellaneous Amendments and Search Requirements revisions to 10 CFR 73.55 (51 FR 27817 and 27822) and to the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The plans, which contain Safeguards Information protected under 10 CFR 73.21, are entitled: 'Donald C. Cook Nuclear Plant Security Plan," with revisions submitted Amendment No. 264. 264 Page 39 of 69 Attachment 1, Volume 16, Rev. 1, Page 105 of 256
Attachment 1, Volume 16, Rev. 1, Page 106 of 256 ITS 5.5 "0' I through July 21, 1988; "Donald C.Cook Nuclear Plant Training and Qualification Plan,' with revisions submitted through December 19. 1986; and "Donald C. Cook Nuclear Plant Safeguards Contingency Plan," with revisions submitted through June 10, 1988. Changes made In accordance with 10 CFR 73.55 shall be Implemented in accordance with the schedule set forth therein. E Deleted by Amendment No. 63 F. In all places of this Acense, the reference to the Indiana and Michigan Electric Company is amended to read Indiana Michigan Power Company. 5.5.2 G. System Integrity The licensee shall implement a program to reduce leakage from systems outside containment that would or could contain highly radioactive fluids durinc a seriousL transient or a6cident to as low as practical levels.*This program shafl include the following: 24 months L.1
- 1. Provisions establshing preventive maintenance and periodic visual inspection requirements, and
- 2. Inte rated leak test requirements for each system at a frequencvfinvt1oxced I
,I r" nn ln Iriiervd n- nf co 2 A 9 M I
.-.- -1 H. ie in The Ii see shall Implement a p ram which will ensure the pability to accurately dete e the airborne Iodine con rntration Invital areas unde accident conditions.
This p gram'shall Include the folh n:
- 1. training of personnel, A.
- 2. Procedures for monitoringand 3 Provisions for maintena of sampling and analys quipment.
- 1. Deleted by Amendment No. 261 I (1) Deleted by Amendment No. 261 I (2) Deleted by Amendment No. 261 I J. The licensee is authorized to use digital signal processing instrumentation in the reactor protection system.
Amendment No. 261 Page 40 of 69 Attachment 1, Volume 16, Rev. 1, Page 106 of 256
Attachment 1, Volume 16, Rev. 1, Page 107 of 256 ITS 5.5 DEFTNIIONS 1.29 Deleted. 5.5.1 OFFS9TE DOSE CALCUIATION tKAWAI. ODC2Q 1.30 The OrFmi DOSE CALCOLATON NAL (ODo) shall contain the 5.5.1.a - methodology and parameters used In the calculation of offaite dose. resulting from radioactive gesous and liquid effluents, In the calculation of gaseous and liquid effluant monitoring alarn/trip xetpo Mand in the Lconduc tof the Evironmental Radiological Monitorinu progra. 1 ODCK sh contain (1) the Radloactive fluent Controls and Radiological cll 5.5.1.b - Environ3ental Monitoring Programs required by Section 6.8.4 and (2) descriptions of the information that should be included In the Annual Radiological Environmental Operating and Annual Radioactive Effluent Release Reports required by Specifications 6.9.1.6 and 6.9.1.7. CASEOUS 1AYIUAMrtTR1AWNTSYSTP= 1.31 A CASEOUS RADWAsM TREATHEZT SYSTEN Is any system designed and Installed to reduce radioactive gaseous effluents by collecting primary coolant system off-gases from the primary syitum and providing for delay or holdup for the purpose of reducing the total radioactivity prior to release to the environnent. 1.32 A VlNTILATION EXHAUST TREASXHN SYSTE is ny system designed and Installed to reduce gaseous radiciodine or radioactive material in particulate .foru in effluents by passing ventilation or vnt exhau t gases through charcoal absorbers and/or WA filters for the purpose of removing Iodines or particulates from the gaseous exhaust stream prior to the release to the environment. Such a systpm is not considered to bave ny effect on noble gas effluents. Engineered Safety Feature (ISF) atmospheric clenp See rlS systems are not considered to be VESIVLUTION UMHAST TREASDIMT SYSTEM -( Chapter 1.0 components. 1.33- FUBCE or PUZGlNO Is the controlled process of discharging air or gas from a confinement to maintain temperature. pressure, humidity,. concentration or other operating condition, In such a manner that replacement air or gas Is required to purify the confinement. 1.34 VITING ine the controlled process of discharging air or gas from a confinement to maintain tperature, pressure, humidity, concentration or other operating condition, in such a manner that replacement air or gas is not proavded or required during V127M;. Vent, used in system names, does not imply a VEtIEG process. COOK NUCLEAR PLANT - UNIT 2 1-7 AM)NIMT N0. Lb. 441 175 Page 41 of 69 Attachment 1, Volume 16, Rev. 1, Page 107 of 256
Attachment 1, Volume 16, Rev. 1, Page 108 of 256 ITS 5.5 314 LIaTIG CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4.0 APPLICABILITY I' 4.0.4 Entry Into an OPERATIONAL MODE or other specified condition in the Applicability of a Umiting Condition for Operation shall only be made when the Lniting Condition for Operation's Surveillances have been met within their specified frequency, except as provided by Specification 4.0.3. When a See ITS 1 Limiting Condition for Operation Is not met due to Surveillances not having been met, entry Into an Section 3.0J OPERATIONAL MODE or other specified condition In the Applicability shall.only be made in accordance with Specification 3.0.4. This provision shall not prevent entry into OPERATIONAL MODES or other specified conditions in the Applicability that are required to comply with ACTIONs or that are part of a shutdown of the unit. 5.5.6 4.0.3 Surveillance Requirements for InserviceI l Honk shall be applicable as follows: 5.5.6.a At least once per 366 days rBierr7e-ny or 4 5.5.6.b C. The provisions of Specification 4.0.2 are applicable to the above required frequences for performing inserviceLEjteio Ntestng acuvities. .. 3
- d. Perfoati of the above inservice o testing actvites I be in addition to other sclidSurveillance R aure . A.4
- e. Nothing in the LASME Boiler and Presure Vessel Codeishall be construed to supersede the requirements of any Technical Specification.'
4.0.6 Deleted Add prcposed IS 5.s.6.c 4.0.7 Deleted \ .5 COOK NUCLEAR PLANT.UNIT 2 Page 3/4 0-3 AM1ENDENT5 P7,4314 24, 265 I Page 42 of 69 Attachment 1, Volume 16, Rev. 1, Page 108 of 256
Attachment 1, Volume 16, Rev. 1, Page 109 of 256 ITS 5.5 ITS REACTOR C t SYSTEM STEM GErTORS LUNITtTI cOmQtON FOR OPERATtON 3.4.5 Each steam generator shall be OPERLE. (See US APPLICABILTY: MOES 1, 2, 3 and 4P I L 3.4.13) Wi1th one or more steam generators Imperable. restore the Inoperable generator(s) to OERA status prior to Increasing Tay9 above ZM*f. SUMRYLAEN R!OuRIMEMTS 4.4.5.0 Each stem generator shall be deonstrated OPERALE by performance of the following augmented toservice Inspection progm and the require-ment of Specification 4.0.5. 5.5.7 A.6 5.5.7 A.4.5.2 Steam G6nntor Tube Sala Selection and tnsvection - The steam generator tube m*ni1u sample size. inspection rosult classifIcation, and the corres ing action requird Shll be as speciffed In Table 4.42. The Insrvc inspectn of steam nratr tubes shall be perfore at the frequencies specified In Spe4.4.5.3 and the inspected tubes shall be verifled acceptable per the acceptance criteria of Spocificat1in 4.S4.5AA tubes selected for each Inservice Inspec-i l Includeat l east 31 of the total raber of tubes In all steam 5.5.7.a genraors; the tubes selected for these Inspections shall be selected on a randn basis except: 5.5.7.a.1 a. . Where experience In similar plants with similar water chemistry Indicates critical areas to be Inspected, then at least 50S of the tubes inspected shall be frou these critical areas. 5.5.7.a.2 b. The first sample of tubas selected for each Inservice inspection (Subsequent to the reservIce Inspection) of each stem ene ratar shall Includ:
- hm AOcifLCAtjno dom noMt apy in M'Ide 4 hille petoming,
%.vroefluhingr SS 10M a ipitIdig Conditlan roc OpeatlO& I_ rSee foc Apficatioa 3.4.1.3 oWe intAd. _ -4 3.4.13 J D.C. COOK - UNIT 2. 314 4-7 £mndmmnt go. 89 Page 43 of 69 Attachment 1, Volume 16, Rev. 1, Page 109 of 256
Attachment 1, Volume 16, Rev. 1, Page 110 of 256 ITS 5.5 ITS RCTOR COMLU SYSTEM SlRVEMLL'It RZSUMRDIENTS (Continutd) 5.5.7.a.2.a) 1. All nonplugged tubes that previously bad detectable wall penetrations (*20:). 5.5.7.a.2.b) 2. tubes in those areas where exportenco hs indicated potential problems. 5.5.7.a.2.c) 3. A tube inspection (pursuant to Specification 4.4.S.4.a.8) shall be performed an each selected tube. It 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 shtalt be selected and subjected to a tube inspect1op. 5.5.7.a.3 c. The tubes selectef as the second and third samples 7T requ~Tr d-by Table 4.42) during each Inservice Inspection my be subjected to a partial tube Inspection provided: 5.5.7.a.3.a) 1. The tubes selected for thet samples include the tubes from those arms of the tube sheot arraywhe tubes with imperfections wee previously found. 5.5.7.a.3.b) 2. The inspections include those portions of the tubes *where imperfections were previously found. 5.5.7.b Te results of each sample inspection shall be classified Into one of the following three categories: CztM!X Inspection Results C-1 Less than 5S of the total tubes inspected are degraded tubes and none of the inspected tubes are defective. C-Z One or more tubes, but not more than 1S of the total tubes Inspected are defective, or between 5S and 1Y- of the total tubes inspected are degraded tubes. C-3 Yore than 10 of the total tubes inspected are degraded tubes or more than 1S of the Inspected Vubes are defective. Note: In all inspections, previously degraded tubes must exhibit significant (>10:) further wall penetrStions to be Included In the above prcentage calculations. D.C. COOK - UNIT 2 314 48 Page 44 of 69 Attachment 1, Volume 16, Rev. 1, Page 110 of 256
Attachment 1, Volume 16, Rev. 1, Page 111 of 256 ITS 5.5 ITS (ID REACTOR COOLANT SYSTEM SURVEILCNCE REQUIREMENTS (Continued) 5.5.7.c 4.4.5.3 Inspection Frequencies - The above required inservice Inspections of steam generator tubes shil be performed at the following frequencies: 5.5.7.c.1 a. The first inservice inspection shall be performed after 6 Effective Full Power Months but within 24 calendar months of initial criticality. Subsequent Inservice inspections shall be perforned at intervals of not less than 12 nor rore than 24 calendar months after the previous Inspection. If tw consecu-tive inspections following service under AVT conditions, not including the preservice inspection, result in all Inspection results falling into the C-l 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 maximun of once per 40 months. 5.5.7.c.2 b. if the results of the 1nservice inspection of a steam generator conducted In accordance with Table 4.4-2 at 4C 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 4.4.S.3.a: the interval may then be extended to a maximum of once per 40oconths. 5.5.7.c.3 c. Additional, unscheduled inservice inspections shall be performed on each steam generator in accordance with the first sample Inspection specified In Table 4.4-2 during the shutdown subsequent to any of the following conditions: 5.5.7.c.3.a) 1. Primary-to-secondary tubes leaks (not including leaks originating from tube-to-tube sheet welds) inexcess of the limits of Specification 3.4.6.2. 5.5.7.c.3b) 2. A seismic occurrence greater than the Operating Basis Earthquake. 5.5.7.c.3.c) 3. A loss-of-coolant accident requiring actuation of the engineered safeguards. 5.5.7.c.3.d) 4. A min steam line or feedwater line break. D.C. COOK - UNIT 2 3/4 4-9 1P Page 45 of 69 Attachment 1, Volume 16, Rev. 1, Page 111 of 256
Attachment 1, Volume 16, Rev. 1, Page 112 of 256 ITS 5.5 ITS . (I3
- I1.
REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) 5.5.7.d 4.4.5.4 Acceptance Criteria 5.5.7.d.1 a. As used in this Specification: 5.5.7.d1.a) 1. IM!rfection 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, ifdetectable, may be considered as inperfections. 5.5.7.d.1.b) 2. Degradation means a service-induced cracking, wastage, wear or general corrosion occurring on either Inside or outside of a tube. 5.5.7.d.1.c) 3. Dearaded Tube means a tube containing imperfections s20% of the nominal wall thickness caused by degradation. 5.5.7 d.1.d) 4. Degradation means the percentage of the tube wall thickness affected or remuved by degradation. 5.5.7.d.1.e) 5. Defect means an Imperfection of such severity that it exceeds the plugging limit. A tube containing a defect. isdefective. 5.5.7.dtl.) 6. PlUoinA g Lmit means the imperfection depth at or beyond which the tube shall be removed from service because it may becoe unserviceable prior to the next inspection and is equal to 40% of the nominal tube wall thickness. 5.5.7.d.1.g) 7. 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 as specified in4.4.5.3.c, above.
5.5.7.d.1.h) 8. Tube nspection mans an inspection of the steam generator tobe from the point of entry (hot leg side) completely around the U-bend to the top support of the cold leg. D.C. COOK - UNIT 2 3/4 4-10 Page 46 of 69 Attachment 1, Volume 16, Rev. 1, Page 112 of 256
Attachment 1, Volume 16, Rev. 1, Page 113 of 256 ITS 5.5 ITS REACOR COOLANT SYSTEM aoveru I Amr Ort~to~ravw rt o.-A 5.5.7.d.1 .1) 9. presefcel nsctinmeans an inspection of the full length or each tub ingenerator performed by eddy current techniques prior to service establish a baul1ne condition of the tubng. This Inspection shalf be performed after the field byrostatic test and prior to Initial POVER OPERATION utsng th equipment and techniques expected to be used during susequent inservicc inspections. 5.5.7.d.2 b. The steam generator shall be determined OPERABLE after cmpleting the corresponding actions (plug all tubes exceeding the plucg ng limit and all tubes containing through-wall cracks) rquired by Table 4.4 2 .~ R. . The provisionsofSR3.02and SR303 are t.4. .5 Repot applicable to the SG Program test Frequencies. A6
- a. following each tnservice Inspection of steam generator tubes, the nider of tubes Plusged in each stem generator shall be reported to the Comission wthin 15 days.
- b. TMe coete results of the stem generator tube insarvice inspection shall be Included In the Annual Operating Report for the per1od In which this inspection was corpleted. This report shell Include:
- 1. NIler and extent of tubes inspected.
- 2. Location and percent of wall-thickness penetration for each Indication of an mpearfection.
- 3. Identification of tubes plugd.
- c. Results of stem generator tube Inspections which fall Into Caftory C-3 and require prompt notification of the CmisiWon tWlrbo rsported pursuant to Specification 6.9.1 prior to { See ITS 5.6 )
resption of plant operation. The written followup of this report snell provide a description of Investigations conducted to determine cause of the tube degradation and corrective measures taken to prevent recurrence. D.C. COOK - UNIT 2 3/4 4-11 Page 47 of 69 Attachment 1, Volume 16, Rev. 1, Page 113 of 256
.ITS a
nZA414 Table 5.5.7-1 mmmmm W fflM M '1 Im O.O w 03 0 ft"m~eift bme~tku, CD r, of mmmi ammatur VW tvt CD mwu F4. i:
'a p".
I&waidl= Alu Thqrnctia nmcu W 0 0 Po CD 0, a, CD -o 1. The kzsvim hsInoete mW be 113its! to asn st gm~ onaratobtim aidoM 0 CD
- 0) ammI~alj~m3bofrnam t" meuia N is the Inr of stumg~asI 0)
Table 5.5.7-1 the. teoad at thrn first ccptdmum lrepactims Um3w*) h If inlld that au su gmaut"s we
- a Footnote (a) pwftt*I In a lIk* mwm. Mf that S mm~b ciramkotm CD as y be ftyd to be mm swoes th I iIc¶patsnI cnkD mm atme gm pMr- the ceri~tias In 9wgtmu. 1i*d r mch *r sts=
ucbmmam the sQ sq
- dmUl be wadif 1. to lzrspmt trn most am~ cmdtiurm.
- 0) 3Table 5.5.7-1 bs UdOSd and bfzwth Wt6 IJutt i rd insebd azrim the i1rat. humimkei 0)
"Footnote (a) "I dIrpd &wmtheqU ctiat smm ad thr rpofameiey.Imfahm
-matiqrapcti"r d*ia fthewUr WI 1bstrwiU" demm'1*k InI a
-N 0, a) -u 0) (a' Zi CD Cn in
ITS Table 5.5.7-2 TABLE 4.4-2 STEAM GENERATOR TUBE INSPECMION
- 0) IST SAMPLE INSPECTION 2ND SAMPLE INSPECTION 3RD SAMPLE INSPIECION 0) 0 Sample Size Result Action Requied Reuit Action Required Result Action Required AtinhumotS Cl None NIA NYA NIA MA 2)
B Tube per S.G. C) CD
-I G2 PluX dedtive tubes snd mspect additional 2S tubes in daisS O.
Gl1 Nont NtA NIA C-2 Phtg defeive tubes nd C-l None inspect addidonil 4S tubes in this SO. 0 C.2 Plux defective tubes CD C-3 Perform Kom for C-3 0, _result offinst sape C-3 Perfbrm action for G3 CD resuhlof first sIpt N/A NIA G3 Inspect all tubes in this 5.0. plug efectve tuba and Al odter S. aM C-I o None WA NA K:) insp.ct 2S tubes in eachother Ca
- 0) S.G. CD
'ag CD Proerp" notification to NRC (71 pursuant to specifIcation 6.9.1 0)
Some S.G.s Perform action for C-2 C.2 but no result ofsecond sampke NIA NYA additional S.O. CD' re C-3. * . -9' Additional S.G. is 03 Inspect all tubes in eah S.a And plug defective NIA NYA la tube por notiftcaion CA to NRC pusant to 0, I
- ecification 9.1. _ -U S - 3(n)% Wbere N is the namber of stea generalors in the uit, and n is the nmber of steam generators inspected during an inspection.
I CD CD CD A.0 =c 0 C/)
- 0) L71 CD
Attachment 1, Volume 16, Rev. 1, Page 116 of 256 ITS 5.5 0 314 LIMITING CONDITIONS FOR OPERATION AND SURVE[LLANCE REQUIREtMTIS 314A REACTOR COOLANT SYSEM. 314.4.10 STRUCTURALINTEGRIrr Y34.4.10.1 ASME CODE CLASS 1.2 and 3 COMPONENTS See CTS Ii LIMTNG CONDMON FOR OPERATION 3.4.10.1 The strUctural Integriy of ASME Code Clas 1, 2 and 3 components shall be maintained In accordance with Specification 4.4.10.1. APPLICABILlTY: ALL MODES ACTION:
- a. With the structural integrity of any ASME Code Class I component(s) not conforming to the above requirements, restore the structural Integrity of the affected component(s) to within Its limit or Isolate the affected component(s) prior to Increasing the Reactor Coolant System temperature more than 50F above the minimum temperature required by NDT considerations.
- b. With the structural Integrity of any ASME Code Class 2 component(s) not conforming to the above requirements, restore the structurl Integrity of the affected component(s) to within its limit or isolate the affected component(s) prior to increasing the Reator Coolant System temperature above 200°F.
- c. With the stutua Interity of any ASME Code Class 3 comnponent(s) not coaflorning to the above requIrets, rcstorc thc strutualr integrity of the affected componerit(s) to within its limit or isolatc the affected component(s) from scr-vice.
SURVEILLANCE REQUIREMENTS 5.5.5 *.4.10.1 In tOter t of S tion 4.0.5 ractor coolant pump flywheel hall be inspected by either qualified in-place UT examination over the volume from the Inner bore of the flywheel to the circle of one-balf the outer radius or a surface examination (magnetic particle testing and/or penetrant testing) of exposed surtfces defined by the volume of the disassembled flywheels once every 10 years. I iAdd l DroDosed ITS 5.5.5 1 C> lgeneric program statement l 3 The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Reactor Coolant Pump Flyweel Inspection Program Surveillance Frequency. COOK NUCLEAR PLANT-UNIT 2 Page 3/4 4-31 AA1ENDMENTr39, 2`G, 265 I Page 50 of 69 Attachment 1, Volume 16, Rev. 1, Page 116 of 256
Attachment 1, Volume 16, Rev. 1, Page 117 of 256 ITS 5.5 ITS 5.5.14.b, 5.5.14.c, 5.5.14.d.1 5.5.14.d.1 i] 5.5.14.d.1 SWV8bF scvV^^ 1* 5.5.14.a 4.6.1:2 Perform lcakage rate testing in acordance with io a R 50 Appendix 1 Option B, except as modified byNRC-pproved ex ts, and Regulatory Guide 1.163. dated Sept9br1995. See Note 1. I Ia. Each containment air lock shall be verified to be In compliance with te requirtments of 1 - rSee ITS
.. .us .... 1 1 & I
- I lt 3.6.1 J Ib. -,Me provisiows of cificaflon 4.0.2 aeo t aplicable. I Q
Notes: Gc 5.5.14.a.1 1 The Type A testing fivqueny specified in NEM 94-01, Revision 0, Paragraph 92.3, a '...at least once per 10 yeas based on acceptable performnce history" is modified to be "...at least oncc per 15 years based on acceptable performance history" This change applies only to the interval following the Type A test performed inMay 1992. COOK NUCLEAR PLANT-UNrf 2 Page 34 6-2 AMENDMENT 443,173,1,9,254 Page 51 of 69 Attachment 1, Volume 16, Rev. 1, Page 117 of 256
Attachment 1, Volume 16, Rev. 1, Page 118 of 256 ITS 5.5 ITS 314 LLUMIG CONDTTIONS FOR OPERATION AND SURVEILLANCE REQUmEMEIrS 314.6 CO NTAINMENTSYSTEMS See Urs CONTAIMENT AIR LOCK 3.6.2 J LlMING CONDMON FOR OPERATION 3.6.1.3 Each comulnneotr lock shall be OPERABLE with:
- a. Both doors dosed except when the air lock Isbeing used for ormal transit d exit thruh the ctument. then at ea one air lock door h be dosed, and 5.5.14.d.2.a), b. An overall air locktage ofS 0.05 L, at P8, 12.0 pslg.
5.5.14.b APPlCAnLH : MODES 1, 2.3and 4. ACTION: See ITS Wkh an afr lock Inopesable,antain a l onedoor doed: r ee ar lock to OPERABLE m within 24 3.6.2 ) hours or be Is leastHOTSTANDBY within de next 6 hours and kn COLD SHUTDOWN within the foUowing 30 hours. SURVENCE REOw UMEMNTS 4.6.1.3 Each cor - air lok shall be _z u OPERABLE: 5.5.14.a L la Icdac wih 10 CFR 50 Appendix I Opdon B and Regulatoy Ouds 1.163. dated Sepwober 1995. and
- b. Atleomamper 6mothsbyverifying aonlyoneinahair kcanbeopened See nS 1 Iat a the. 3.6.2 J COOK NUCLEA PLANT-UNIT 2 Pap 314 4 AUENDMEW 193 Page 52 of 69 Attachment 1, Volume 16, Rev. 1, Page 118 of 256
Attachment 1, Volume 16, Rev. 1, Page 119 of 256 ITS 5.5 ITS 314 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 314.7 PLANT SYSTEMS SURVEILLANCE REQUIREM ENTS ITS 5.5.9 generic program statement 4.7.5.1 The control room emergency ventilation system shall be demonstrated OPERABLE:
- a. Deleted See ITS 3.710
- b. At least once per 31 days on a STAGGERED TEST BASIS by initiating flow throug the HEPA filter and charcoal adsorber train and verifying that the system operates o at least 15 minutes.3.L3j A8 5.5.9 C. At least once perLMinonzhs or (I) after a t or charcoal adsorber housilgs or (2) following painting. fire or chemical release In any wtile ItIs n operation that ventilation zone communicating with the system. - could adversely hafc the filter bank or charcoal 5.5.9.b I Verifying that the charcoal adsorbers remove 2 99% of a halogenated adsorber capability hydrocarbon refrigerant test gas when they are tested in-place in acordance with ANSI N5101975 while operating the ventilation system at a flow rate of 6000 cfm E 10%.
55.9.a 2. Verifying that the HEPA filter banks remove 2 99% of the DOP when they are tested in-place in accordance with ANSI N510-1975 while operating the ventilation system at a flow rate of 6000 cfm +/- 10%. 55.9.c 3. Verifyinglwiflvn at a laboratory analysis of a carbon sample from either at least one test canister or at least two carbon samples removed from one of the charcoal adsorbers shows a penetration of less than or equal to 1.0% for radioactive mncthyl iodide when the sample Is tested In accordance with ASTM D3S03-1989, 30°C 95% RH. The carbon samples not obtained from test canisters shall be prepared by either: 5.5.9.c. 1 a) Emptying one entire bed from a removed adsorber tray. mixing the adsorbent thoroughly, and obtaining smplcs at least two inches in diameter and with a length equal to the thickness of the bed, or 5.5.9.c.2 b) Emptying a longitudinal sample from an adsorber tray, mixing the adsorbent thoroughly, and obtaining samples at least two inches In diameter and with a length equal to the thickness of the bed. 5.5.9.a. 4. Verifying a system flow rate of 6000 cfim + 10% during system operation 5.5.9.b when tested in accordance with ANSI N510-1975. COOK NUCLEAR PLANT-UNIT 2 Page 314 7-15 AMENDMENT 340252 Page 53 of 69 Attachment 1, Volume 16, Rev. 1, Page 119 of 256
Attachment 1, Volume 16, Rev. 1, Page 120 of 256 ITS 5.5 ITS 3/4 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 314.7 PLANTSYSTEMS SURVEILLANCE REOUIREMENTS (Continued) 5.5.9 d. After every 720 hours of charcoal adsorber operation by either: 5.5.9.c 1. Verifyingl wtin 31 daes nfleciemoval that a laboratory analysis of a carbon sample obtained from a test canister shows a penetration of less than or equal to 1.0% for radioactive methyl iodide when the sample is tested in accordance with ASTM D3803-1989, 300 C, 95% R.H; or 5.5.9.c 2. Verifyingjwiin 31 ;vs after imoval that a laboratory analysis of at least two carbon samples shows a penetration of less than or equal to 1.0% for radioactive methyl iodide when the samples are tested in accordance with ASTM D3803-1989,30°C, 95% R.H. and the samples are prepared by either 5.5.9.c.1 a) Emptying one entire bed from a removed adsorber tray, mixing the adsorbent thoroughly, and obtaining samples at least two inches in diameter and with a length equal to the thickness of the bed, or 5.5.9.c.2 b) Emptying a longitudinal sample from an adsorber tray, mixing the adsorbent thoroughly, and obtaining samples at least two inches in diameter and with a length equal to the thickness of the bed. Subsequent t reinstalling the adso cr tray used for obtai ing the carbon sample, the ssm shall be demons ed OPERABLE by also: a) Ve fying that the charcoal adsorbers remove 2 99 of a halogenated hydrocarbon refigerant test gas when they are tested in-place in ordance with ANSI 510-1975 while operaing the ventilation stemataflowrateof cfm+/- 10%,and -0 b) erifying that the HEP filter banks remove 2 % of the DOP when y are tested in-plac in accordance with SI N510-1975 while operating the ventilatio system at a flow rate of cfm F 10%. COOK NUCLEAR PLANT-UNIT 2 Page 314 7.16 AMENDMENT24, 261 Page 54 of 69 Attachment 1, Volume 16, Rev. 1, Page 120 of 256
Attachment 1, Volume 16, Rev. 1, Page 121 of 256 ITS 5.5 ITS 3/4 LMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4.7 PLANT SYSTEMS 5.5.9 SURVEILLANCE REQUIREMENTS (Continued) C. At least once per[ Months by:
,e
- 1. Verifying that the pressure drop across the combined IIEPA filters and 5.5.9.d charcoal adsorber banks is less than 6 inches Water Gauge while operating the ventilation system at a flow rate of 6000 cfm plus or minus 10%.
4H
- 2. a. Verifying that on a Safety Injection Signal from Unit 1 the system automatically operates In the pressurizationtcleanup mode. See tTS 3.3.7 and ITS 3.7.10 J
- b. Verifying that on a Safety Injection Signal from Unit 2, the system automatically operates in the pressurization/cleanup mode.
- 3. Verifying that the system maintains the control room envelopelpressure boundary at a positive pressure of greater than or equal to 1116 inch W. 0.
relative to the outside atmosphere at a system flow rate of 6000 cfm plus or See rrS minus 10% with a makeup air flow rate of < 1000 cfm. 3.7.10 5.5.9 fter each complete or partial replacement of a IIEPA filter bank by verifying that the HEPA filter banks remove greater than or equal to 99% of the DOP when they are 5.5.9.a tested in-place in accordance with ANSI N510-1975 while operating the ventilation system at a flow rate of 6000 efmn plus or minus 10%. 5.5.9 g. [ter each complete or partial replacement of a charcoal adsorber bankrb-y verifying
,that the charcoal adsorbers remove greater than or equal to 99% of a halogenated 5.5.9.b hydrocarbon refrigerant test gas when they are tested in-place In accordance with ANSI N510-1975 while operating the ventilation system at a flow rate of 6000 cfm plus or minus 10%. nA9 The provisions of SR 3.0.2 and SR 3.0.3 are I applicable to the VFTP test Frequencies. J COOK NUCLEAR PLANT-UNIT 2 Page 3/4 7-16a AMENDMENT 9. 424, }1-58, 2.
424,252 Page 55 of 69 Attachment 1, Volume 16, Rev. 1, Page 121 of 256
Attachment 1, Volume 16, Rev. 1, Page 122 of 256 ITS 5.5 ITS 314 7.-6 VI? mrW .M trifTnnWO COtMtnQf TM otflA??OM 3.7.4.1 Two ludoperdmnt U? vmntilatlon ststam sxhaust air fllter trains
*sall'bs CIAL.
See ITS hZC.Iu Z: kIOCSU 1, 2. 3 asd 4. -I 3.7.12 J with m isr ventilsaion systue exhaust alr filter train lnaperable, resrt:. the ineperable train to OftU tZ status within 7 a or be in as lest NOT STAMY within she next £ houts and- CinOW Sjn = within the following 30 heurs. hos. . J Z p!Add proposed ITS 5.5.9 generic program statement 4.7.4.1 lash U? ventilation sytem exhaust air fliltr train sh1l be deonsteractd OVUAUIX: se.;rrs])
- a. At least .me per 31 dayg an a*lSUC GOMM SAM by initiatlng. , 13.712 5.5.9 b.
frod the control roe., now chrough the IM filter and charcoal
* *other train mnd verifying that the train operates for a* least t M. utta.
At least oWe per tUs a ).(l) or
--- monrnsMO
&fTEr ay astrutuual naintendiwe e ,e on tshe 31 filter at 4lazcaa1 adasober houings. or Cl) following painting. fire or chael niase in ay ventilation anea whiiE3 It Is In operation cossiuciaig with the orby:%"ere that could adversely affe( Atthe filter bank or
- 1. Delatad.. char coal adsorber cap; ability I
5.5.9.b 2. V rWii thou the ebaresal adorbers ream 2I P 1 a hala eutd bydrocarbon refrigerant tast gas Vmm they ar% tat .plse In aecordanza with AUZ 9310.1110 Adle operating the Veilatione syst e &: , flow race *t 23.000 esa I
+/- lot.
5.5.9.a 3. Verifying that ths XUA flltr bank. ream I Ph of the DO? when tu l are tated in plaae La a*cJrdance with AIlS 310-1910 while operating the ventilation scitaeat a flow rate of 25,000 I CIO
- 10l.
- 0. -C. CO . OM 2
- 3/4 7.17 Ament ItolllI Page 56 of 69 Attachment 1, Volume 16, Rev. 1, Page 122 of 256
Attachment 1, Volume 16, Rev. 1, Page 123 of 256 ITS 5.5 ITS 314 LIMITING CONDITIONS FOR OPERAIION AND SURVEILLANCE REQUIRMENTS 34.7 PLANTSYSTEMS I SURVEIANCEmaa a ratory analysis ofa
- 4. Verifying IW~ -1C4MV1 WAi f a carbon 5.5.9.c sample from either at least one test canister or at least two carbon samples removed from one of the charcoal adsorbers shows a penetration of less than or equal to 5% I for radioactive methyl iodide when the sample is tested in accordance with ASTM D3803-1989, 30°C, 95% Rl.H, and 2 45.5 fpm face velocity. The carbon samples not obtained frm test canister shall be prepared byeither 5.5.9.c.1 *a) Emptying one entire bed from a removed adsorber tray, mising the adsorbent thoroughly, and obtaining samples at least two Inches in diameter and with a length equal to the thickness of the bed, or 5.5.9.c.2 b) Emptying a nhitudinal sample from an adsorber tray, mixing the adsorbent tboroughly. and obtaining samples at least two inches in diameter and with a length equal to the thickness of the bed.
Subsequent A reinstalling the adsorb/b tray used for obtaining e carbon sample, the systemhall be demonstd OfERABLE by also verify g that the charcoal adsorbers cmove greatcr than or equal to 99% of a hal refri 19SO so minus/0%. test gas when they areksted In-place In ated hydrocarbon with ANSI N5I0-la operating the 'vendlaon systen a a fow uaty of .000 cfm plus or -0G J 5.5.9.a. S. Verifying a system flow rate of 25,000 cfm plus or minus 10% during system 5.5.9.b operation when tested in accordance with ANSI N5 10-1980. 5.5.9 C. rulcr every 7su noun or cnarmco adsorue uperu"a1 uy eiwma
- 1. Verifying d s31 uercxtmovati tat a laboratory analysis of a arbon 5.5.9.c sample obtainme a t[st shows a penetraon of less than or eualto5 for radioactive methyl iodide when the sample is tested In a e with ASTI D38O3-l9S9,30 C,95% Rl, and2 45.5 fm face veiocity, or If 5.5.9.c 2. Verifying IWithidn 31 dpOs aftcr-emoval thar laboratory analysis of at leat two I carbon samples shows a penetrtion of less than or equal to 5% for radioactive rnethyl Iodide wben the samples are tested Inaccordanee with ASTM D3S03-1939.
30rC 95% RL.and ; 455 fpm face velocity ard th samples are prepared by either I 5.5.9.c.1 a) Emptying one entire bed from a removed adsorber tray, mixing the adsorbent thoroughly, and obtaining samples at least two inches in dWmeter nd with a length equal to the thickness of the bed, or COOK NUCLEAR PLANT-UNIT 2 Page 3/47-18 ANMENDMENT -1,uo, 240 Page 57 of 69 Attachment 1, Volume 16, Rev. 1, Page 123 of 256
Attachment 1, Volume 16, Rev. 1, Page 124 of 256 ITS 5.5 ITS Ma~iM&, mleua 5.5.9.c.2 b) Zuptying a laegtudLul ample frau an adeOrber tray. af-ig the adsorbent thoroughly, and ebtalng amples at least two indwe In diamter and with a length equal to the Alaheua of tba bed. Sub eque to riaetalls fg efaer tray used/far obini t he u wmlo,*the etqh ebaih ibe dawatrad 0 b I ale. r? gem 32 te thaet tbe oh~el adaothe of they an tae While a
- hlontd 1 lNSI hro e
r gretetr tha grant test atertaa ft2a at liwwtiletn
-0G 5.5.9 4.
1. ra Azt last Due ef 23.000 ef pine 6r WWribauhe by: Verifying that the preu 10:. dnp uress the camobed MA I
--G 5.5.9.d filter eand charoal edsther banks Is 'se than 1 Inches Uater CGe whle operating the wentilAtLOn sy*tem at a flaw rate of 25,000 ofe plu osr weV 10:.
- 2. Deleted.
- 3. TeVa1ing that the stwoy fem scat anttaLeally en a flow thzu&i the l, lresuse--Iigb-]igb Signal and dirests Its ehuxt
-nutaint, cltersa fll tharcoal "exther-banks an a _ 4 See ITS I 3.7.12 J Cort-sItnt Freaeuse-Vagh.Kgh SlIg .t 5.5.9 each _o _o repiossuent of a A filter bank byb Torif ingEFULs ,rSvutwe baus m grate r ti or aqwl 5.5.9.a sn to of the DO? when they are est greet cseIn or eqalh S 1010 operating the entilation te at a flaw Ustft of 25,00 cfot1 pls ornian 1OX.
5.5.9 -* f. ftsr Gltet er rtal.t1t of/all adher Ba " ~ that tag dsereater Ssl than 5.5.9.b ore~jes eo *S.f
- hteloened yrcth lrefie rart ten gt tis t,, - k~.X Xe l-to whle orptis the etiai stm at
- flow rto 25,000 e.6 p
p or aims M. The provisions of SR 3.0.2 and SR 3.0.3 areI r f. . rapplicable to the VFTP test Frequencies. ln of 72u prelfS pdfec cint *.0.9 r Uplcable. l See ITS m It prvia , To --- [.7. CO= RULCLEA11 - tI? 2 3/4.7.19 AMMWI NO. &,15&U. I Page 58 of 69 Attachment 1, Volume 16, Rev. 1, Page 124 of 256
Attachment 1, Volume 16, Rev. 1, Page 125 of 256 ITS 5.5 ITS 34 LAtG CONDMTONS FOR OPERATION AND SuRvEILLANCE REQUmMENTSM 3/4.b ELECTRICAL POWER SYsItMS 4.8.1.1.2 Each diwgener at or hshil be deosrtdOPERABLE: seIS
- a. in accordax wt t frquency specitied in Table 4.831 on a STAGGERED TEST tTS 3.8.3.
I 1. rven the fuel kovd in the day tzak. I See rrS
- 12. Veritying tbe fuel Icvd [a tbe hael - Se rTS
- 3. Veriqna that the fel tur fer pump can be started and that It transfers hel{ 3.8.3 J fro the stoa system to the day tank.
- 4. Verin ta the diesd starts from standby conditions and achieves in less than See ITS]
orequltol0 secads,voltage -4160420Vmndfhequency -60tl.2Hz. 3.8.1
- 5. Verying the dieseL Is synchronizd and loaded and operates for greater than or equs to 60 nrn t*es a a lad of 3500 kw*, An I See ITS 3.8.1 and
- 6. Vetifty that the disel eerator is ained to provide snfdby power to the ITS 3.8.3 assocated emeny basses.
I h- By remving acted
- 1) From the day tank a least once per 31 days and after eacb occaion when the - ( See ITS I diesel Is opct for er thn I hour.ad r 3.8.1 J See rrs I
3.8.3 5.5.1 1.a 5.5.1 1.a
- The dies tener stan (10 seconds) fxom standby condetons sh be pertb at ls once I per 154 days Inte suveilance tests. ARl other eng starts for th pupoe of ths auelllanc See ITS testing ad compensatoty ato may be at redued scceraslon rat an reommded by the 1 3.8.1 J mafar so that caical stss and wear on the disl eaine ae inhlzdd
** Momentary load transet do cot Isn date hs tat. I -
*** The actions to be taken ould any of the properties be foud odside of siIe ts are I See rrs deftned In e BaSs. { 3.8.3 J COOK NUCLAR PLNT- 2 Pop 3/4 53 AWlDWM4. 191 Page 59 of 69 Attachment 1, Volume 16, Rev. 1, Page 125 of 256
Attachment 1, Volume 16, Rev. 1, Page 126 of 256 ITS 5.5 ITS 3/4 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS Y4.8 ELECTRICAL POWER SYSTEMS SURVEILLANCE REOUIREMENTS (Continued) 5.5.1 1.a.2 a) A kiD certifi 5.5.11.a.2 b) A flas 5.5.1 1.a.1 2) By verifying,Fk adding the new oran absolutei H60W1Pwhen c 5.5.11 .a.3 3) By verifying. P adding new fu appearance wid 5.5.1 1.b 4) y ii ski bAe nterfos ir
- d. At least once per 31 d; 5.5.1 1c
- e. At least once per 18months, during shutdown, by:
Subjecting the diesel engine to an inspection in accordance with procedures prepared in conjunction with its manufacturers recommendations for this class of tanvhv servi The provisions of SR 3.0.2 and SR 3 0re applicable to the Diesel Fuel Oil Testing A10 Program test Frequencies. The actions to be taken should any of the properties be found outside of the specified limits are defined in the See ITS Bases. l 3.8.3 COOK NUCLEAR PLANT.UNIT 2 Par 3Y4.84 AMENDMENT ArI-S, 261 Page 60 of 69 Attachment 1, Volume 16, Rev. 1, Page 126 of 256
Attachment 1, Volume 16, Rev. 1, Page 127 of 256 C) ITS 5.5 314 LIMITING CONDITIONS FOR OPERATION AND SURVEIMANCE 314.9 REFIUEING OPERATIONS REQUIREMENTS STORAGE POOL VENTILATION SYSTEM* LIMmNG CONDITION FOR OPERATION 3.9.12 The spent fuel storage pool exhaust ventilation system shall be OPERABLE. See ITS APPLICABILITY: Whenever Irradiated fuel is in the storage pool. ACTION:
- a. With no fuel storage pool exhaust ventilation system OPERABLE, involving movement of fuel within the storage pool or suspend all operations crane operation with loads over the storage pool until at least one spent fuel storage pool exhaust ventilation OPERABLE statu. system is restored to
- b. The provisions of Specificaion (3.0.3 re not aitplicable SURVEILLANCE REQUIREMENTS Adposed ITS 5.5.9 generic program statement 4.9.12 The above required fuel storage pool ventilawon system shall be demonstrated OPERABLE: tSee uTS
- a. At leat once per 31 days by Iniriating flow through the HEPA filter and charcoal absorber train 3.7.13 a2d verifying that the train operates for at least 15 miovte ofa 5.5.9 b. At least once peratingothe exhaust aentiatio stem its aloraten charcoal adsorber housingsaor (2) following paiuxiiga fire or cherica of 3 h0cfP fi10%.
aone communicating withoventelithin release in an e untildr.1 t st he p cnoperation with lo (7
- l. Deleted..chroa dobrapilt 5.5.9.b 2. Verifying that the charcoal adsorbers; remove 2 99X of a halogenated hydrocarbon refrigerant test g is when they are tested In-place in accordance with while operating the exhaust ventilatdon system at a flow rate ANSI N510-1980 of 30.000 cfin +/- 10%.
ove pool.itr therel storag an o eThe crane bay roll-up door and the south door of the auxiliary building crane bay mnay be opened under See rrS adminiistrative control during movement of fuel within the storage pool or crane operation with loads A { over the storage pool. 3.7.13
** Shared system with D. C. COOK - UNIT 1.
COOK NUCLEAR PLANT-UNIT 2 Page 314 9-12 AMIENDM.NEN rUl, 2n2, z, 265 Page 61 of 69 Attachment 1, Volume 16, Rev. 1, Page 127 of 256
Attachment 1, Volume 16, Rev. 1, Page 128 of 256 ITS 5.5 ITS 31 LUM1NG CONDITIONS FOR OPAION AND SURVEILIANCE REQUIIRIY
*SS REFIJEIG OPERAMIONS
- 3. Veifytdn that e I1EPA filtr bnks remove great thM or equal to 99% of the DOP 5.5.9.a when the gm re sted 6iplace In accordanee with ANSI 51s0-1980 while operang the exhanstVentlbaton Sy at a Hlow rate of 30,000 din plus or minus 10%.
- 4. Val bi efmnlS sblho ~ weS 5.5.9.c et eatInaen t canister or at least two carbon samples removed from one of th charcoal admods shos a peeutdon of les fta or equal to 5% for radioactive methyl iodide when th sample Is Wsted In aswrdace withb AM D3103-1999. 30CC 95%
RH.. d2 46. fpm fare veloiy. The carbon mples not obtained fom test canistes
- lh bepepaped by ether:
(a) Thqyg am endre bed from a removed alsorber tria. mxig the adsorbent 5.5.9.c.1 torougl, and btaning samples at lea two Inches In diameter an with a klgth equal to te thidnes of dte bed, or 5.5.9.c.2 (b) Emptyn a logltudlna sample from an adrber tray, mixing tbe adsorbet horoughly. ad obtahin samples a last two Inc In dmtr and wih a lth equal bathe tkness ofthe bed. Sb ln Ihe sbe wr t f Carbo
. ae f m swat o e d cPEpl.Bubq0 oy d tiIneuis ggoperatlOe duscd/bn ~ r ql to99*of a balogtued /
whenn te In a c r n e tje we lwthd i In accord wit 198 wlld o 0NS>X veafflatift *flow rat of 30,000 5.5.9.a, 5.5.9.b whe Jf in Wd with ANSI NSS10-1990. 5.5.9 . eAftr y720 bos of cha ol sob operaion by either 5.5.9.c 1. Vertfying 31 ao
!that a labonrtozy a ysis of
- carbon a ph ob ed Mu aV tr a PeUnteaon pof kl than or equal to 5% for ra*diowtive methl IOd when th sample is tested In aordanee with ASW D3803-1989, W0', 95% 9LH and 2:46. fm ae veockty.
COOKC NUCLEAR PLANT-UGT 2 Pag Y4 9-23 AMNDMENTU1I4Q,43 240 I Page 62 of 69 Attachment 1, Volume 16, Rev. 1, Page 128 of 256
Attachment 1, Volume 16, Rev. 1, Page 129 of 256 ITS 5.5 ITS 3M LbIffnNG CODPON3 FOR OPERATION AND SURVEILANCE REQURMENT 3M4.9 RMWIELING OPM~ATIONS EMMA=lE lIR Montiudc l 5.5.9.c 2. Veihws Wu 31 di dlcra IoltoqP aials of at kest two cabo aspics shows a o or equal to 5% for radioactive methyl iodide wbn the samples we sted In caccmance with ASM D3M0 989. M0C, 95% J., and246.8 fipm face elocityand lt samples me prepared by eidh. 5.5.9.c.1 (a) Emptyng oa endr bed ftm a moved adsober tray, iming tm adsor throghy, a dtaining b samples at leas two Inches In diameter and with a
- ngth equal btheiof the bd, or 5.5.9.c.2 (b) Emptying alongitudnal sample from an adsoDber tray. miing th adsbent
*troqfl. and abodnizg samples at last two IneHes in diameter and with a length equal io the ftbkess of'dhe bed.
SubAquent =us 0 a1 e uay aO=ot O= 111Czmoo me s=ple. th ! tm s: 9 6bedost OPERABLE by ulsoig that the charoal ot remre tan equal to 99%/of a hui g h)*oesrt 9N0.1930 cemnm10. / they am Ventilation system accordac with rate of 30,000 ,L. 5.5.9 d. A least oncpa 5.5.9.d 1. Vw ng ta the pressu dopW scruo the combined HEPA filtes and charcoal dsdrber banks is les than or equal to 6 inches Water Gauge while operatn the exhaust ventilaon syste ata flow rat of 30,000 cn plus or minus 10%.
- 2. Dleted.
- 3. Vitrityng *At on a nig marm sip" Oe system atw eaiey dquects U exu ust 11ow thtsgh the charcoalIad b banks and antomatically shuts down the storage pool Ventilation qsyt p f .
r See ITS
- 4. Verfi** dtha the ex as1ellstion system maintain the spent fue storage pcol aea at ( 3.7.13 J a neative pressue of grSater than or equal to 1/8 indie Water Gawp relaive to the ous atmspbers dwin SYstem operation COOK NUCLEAR PLANT-UNrr 2 Pasp 34 9-14 AMENDNMENM14,M4,240 I Page 63 of 69 Attachment 1, Volume 16, Rev. 1, Page 129 of 256
Attachment 1, Volume 16, Rev. 1, Page 130 of 256 ITS 5.5 ITS
- UETCZ 0 5.5.9 a. iktur each complete or partial replaemmnt of
- RUA filter baznkXby
~'WryL~ng that th. RUA ul1tar DSast reamo ? '9S of thu DO? when 5.5.9.a they are tasted La-place in accordance with ANSI 331041980 while I
_*ratiungs teb Ye1tlation system at a flow eatsf30.000 cf t 10. 55 Afer cohWletoer patial ceplacement of*a charcoal adsorbor 5.59 fying that the chsreal adsorbers rom k 39% of a halogeated y b rfrgerant taut Ps whe they are tsted 559b 1-place, in aCcorac with ANSI 1510.1930 while oerating the ventilation syastm at a flow rate of 30.000 cb
- 109.
e provisions o 3.0.2 and SR 3.0. e A. applicable to the VFTP test Frequenis I r. D. C. CoOc - MNIT 2 3/4 9-13 Amendent lo. 111 Page 64 of 69 Attachment 1, Volume 16, Rev. 1, Page 130 of 256
Attachment 1, Volume 16, Rev. 1, Page 131 of 256 ITS 5.5 3/4 3/4.11 LIMiTING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS RADIOACTIVE EFFLUENTS I LIQUIlD HOLDUP TANKS* LIMITING CONDITION FOR OPERATION IoS5.5.10 generic programa 5.5.10, 3.11.1 The quantiy of radioactive material contalned n each of the following tanks shall belimitdto 5.5.10.c Ithan or equal to 10 curies. excluding tritium and dissolved or enrained noble lases/
- a. Outside temporary tanks.
APPLICABILITY: At all times. ACTION: a./ With the quantity of radioactiv material In any of the above litks exceeding the above limit, without delay suspend I additions of radioactive ma to the tank and within 48 bour IA. reduce the tank content to ithin the limit. The provisions of Specific zion 3.0.3 are not aplicable. SURVEILLANCE REQUIREMENTS 5.5.10.c 4.11.1 The quantity of radioactive material contained In each of the above listed tanks shall be determined to be within the above limit b anali a entatie sample of the tank's contentstletc r 7A) 17 days %hen radioactive mfiterials are being MM to the [The provisions of SR 3.0.2 and SR 3.0.3 are-4 ~-applicable to the Storage Tank Radioactiit tMontoing Program Surveiffance Frequencies. 5.5.1 Oc Tanks included in this Specifications are those outdoor tanks that are not surrounded by liners, dikes, or wals capable of holding the tanks contents and that do not have tank over flows and surrounding area drains connected to the liquid radwaste treatment system. COOK NUCLEAR PLANT-UNIT 2 Page 314 ll1 AAU1NDNMENT 1,Wj,W7, 265 I I Page 65 of 69 Attachment 1, Volume 16, Rev. 1, Page 131 of 256
Attachment 1, Volume 16, Rev. 1, Page 132 of 256 ITS 5.5 314 314.11 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS RADIOACTIVE EFFLUENTS .1 3/4.11.2 GASEOUS EFFLUENTS E(PLOSIVE GAS MDffURE .* LIMITING CONDITION FOR OPERATION Add proposed rrs 5.510 generk program statement 5.5.10. 3.11.2.1 The concentradon ofoxyeninthewase a holdup system shall be liniitedrto or eo3% 5.5.10.a by volu fthe hydrogen inthe system i Atcr than or equal to 4% by vol APPLICABILITY: At all times. ACTION:
- a. hth the concentration of oMY in the waste gas holdup system ter than 3% by volume but less than or equal to 4% by vol and containing greater than or to 4% hydrogen, restore the concentration of oxygen to ithan or equal to 3% or reduce hydrogen concentration to less than4% within 96 hours.
With the concentration of oxgen inthe waste gas holdup sys or tank greater than 4% by volume and greater than 4% drogen by volume without delay al additions of waste gases aspend to the system or tank and tace the concentration of oxygen less than or equal to 3% or the concentration of hydrogen tess than or equal to 4% within 96 ours in the system or tank.
/ c. The provisions of SpecificajSon 3.0.3 are not applicable.
SURVEMIANCE REOUIREMS 5.5.10.a 4.11.2.1 The concentration of oxygen in the waste gas holdup system shall be determined to within the above limits by continuously monitoring the waste in iases the waste gas bolduosyste xyg montoRrequired OPERABLE by Table 3-12 of Speciticaton 3.3.3. The provtslons of SR 3.0.2 and SR 3.0.3 are
* - . applicable to the Explosive Gas Radioactivity Monitoring Program Surveillance Frequencies.
COOK NUCLEAR PLANT-UNIT 2 Pagp 3/4 112 AMENDMENT 54, P8, PS , 265 I Page 66 of 69 Attachment 1, Volume 16, Rev. 1, Page 132 of 256
Attachment 1, Volume 16, Rev. 1, Page 133 of 256 ITS 5.5 3/4 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUMIEMENTS 314.11 RADIOACTIVE EFFLUENTS GAS STORAGE TANKS LIMITING CONDTON FOR OPERATION Add proposed ITS 5.5.10 generic program statement 5.5.10. 3.11.2.2 The quantity of radioactivity contained In each gas storaie tank shall be limited to 43.800 curies noble 5.5.10.b gas (considered as Xe-3). I t-A.7J APPLICAB IITY: At all times. ACTION:
- a. With the quantity of radioactive delay suspend all additions of contents to within the limit.
terial in any gas storage tank e ig the above limit, without
*oactive material to the tank andwithin 48 hours reduce the tank '-0 The provisions of Specifi 3.0.3 are not applicable.
SURVEILLANCE REQUIREMENTS 5.5.10.b 4.11.2.2 The quantity of radioactive material contained In each a storage tk shal be determinfd in b wihhn the above limi at least once m days whenever radioactive sals are added to the tan nd at least LA.7 once per 24 hours during p96r coolant Y., dega ooxdmtions. ZV The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Storage Tank Radioactivity Monitoring Program Surveillance Frequencies. i
%I rorn r - Xw . A xwr% 15"U"Ir.AArL-ZAnA-ULNII A Page 3/4 11-3 AMENDMENT 61, A, {68, 265 I
Page 67 of 69 Attachment 1, Volume 16, Rev. 1, Page 133 of 256
Attachment 1, Volume 16, Rev. 1, Page 134 of 256 ITS 5.5 ITS
- 6. ADNMGSTRJUTM CONTROLS 6.1.Wejien irocam sUaal be estblab~ed, kv1ewnted and zmaimisned covedog te actiides zefuereced beow-.
-(See rrS L TIbe applicable procCeine gneo=ncdod in Appendi WAof Rezlaoi Guide 133, Rev. 2, Febeusy 1973.
- b. Dekeed.
C. DhelC&d
- d. PROCESS CONRMOL PROGRAM~h eomo C. OPFSFM DOSE CALCUI.ATIN MANUAL ih1amenudc.
f Quality Ammea~e Progra fo efflm and em'fronmcnW al m Ftcin aiog fi guidance in Reaubiatry Guide 1.21. Rev. 1.Jne 1974. and Reaulatoxv Guidde 4.1. Rev. 1.Andl 1975. 5.5.4
- b. Fire Protection Prop= implemeatedon.
6.3.2. Elc prCedu= sni adniise aiv j~o~cy of Speefi~ranc C.M. above, and ebanges dse*o iziluin
- See rTS teavoraiy chegee "~1 be meyiewed pdorl ftpkcw as act fw*~ in QuhA b=feao Auanmne f 5.4 )
IS45 6.8.3 Deleted. COOK NUCa.AR PXANT4JUn'2 Pagp 6- A WENDM Nr58,43, p,47,8 4#31O,444 Page 68 of 69 Attachment 1, Volume 16, Rev. 1, Page 134 of 256
Attachment 1, Volume 16, Rev. 1, Page 135 of 256 ITS 5.5 ITS 6.0 ADMINISTRATIVE CONTROLS 6.13 PROCESS CONTROL PROGRAM (PCP) 6.13.1 Changes to the PCP:
- a. Shall be documented and records of reviews performed shall be retained as required by the Quality Assurance Program Description, Appendix C. Section 6.10.2.n. This documentation shall contain:
See CTS
- 1. Sufficient Information to support the change together with the appropriate analyses or - { 6.0 J evaluations justifying the change(s) and
- 2. A determination that the change will maintain the overall conformance of the solidified waste product to existing requirements ofFederal, State, or other applicable regulations.
- b. Shall become effective after review and acceptance by the PORC and the approval of the Plant Manager.
6.14 OFFSITE DOSE CALCULATION MANUAL (ODCM) 5.5.1.c 6.14.1 Changes to the ODCM: 5.5.1 .c.1 a. Shall be documented and records of reviews rformed shall be retained as ul te us it lVsur-aWcWrgrMM Descifn A C'sediIetinn 6-0 v n I Thits documcntation shall contain: 5.5.1.c.1.a) l. Sufficient information to support the change together with the appropriate analyses or evaluations justifying the change(s) and 5.5.1.c.1.b) 2. A determination that the change will maintain the level of radioactive effluent control pursuant to 10 CFR 20.1302, 40 CFR Part 190, 10 CFR 5036a. and Appendix I to 10 CFR Part 50 and not adversely impact the accuracy or reliability of effluent, dose, or setpoint calculations. 5.5.1.c.2 b. Shall become effective afterri the approval of the .tlant 0 anager. 5.5.1 .c.3 C. Shall be submitted to the Commission In the form of a complete, legible copy of the entire ODCM as a part of or concurrent with the Annual Radioactive Effluent Release Report for the period of the report in which any change to the ODCM was made. Each change shall be identified by markings in the margin of the affected pages, clearly Indicating the area of the page that was changed, and shall Indicate the date (e.g., monthlyear) the change was Implemented. qAdd proposed ITS 5.5.13 and ITS 5.5.15
)--e COOK NUCLEAR PLANT.UNIT 2 Page 6-1s AMENDMENT74, I^,4-76,214,226,261 Page 69 of 69 Attachment 1, Volume 16, Rev. 1, Page 135 of 256
Attachment 1, Volume 16, Rev. 1, Page 136 of 256 DISCUSSION OF CHANGES ITS 5.5, PROGRAMS AND MANUALS ADMINISTRATIVE CHANGES A.1 In the conversion of the CNP Current Technical Specifications (CTS) to the plant specific Improved Technical Specifications (ITS), certain changes (wording preferences, editorial changes, reformatting, revised numbering, etc.) are made to obtain consistency with NUREG-1431, Rev. 2, "Standard Technical Specifications-Westinghouse Plants" (ISTS). These changes are designated as administrative changes and are acceptable because they do not result in technical changes to the CTS. A.2 CTS 6.8.4.a specifies the requirements for the Radioactive Effluent Controls Program, however there is no statement as to whether or not the provisions of CTS 4.0.2 and CTS 4.0.3 are applicable. ITS 5.5.3 states that the provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Radioactive Effluent Controls Program Surveillance Frequencies. This changes the CTS by adding the allowances of ITS SR 3.0.2 and SR 3.0.3 to the Radioactive Effluent Controls Program. This statement is needed to maintain allowances for Surveillance Frequency extensions contained in the ITS since ITS SR 3.0.2 and SR 3.0.3 are not normally applied to Frequencies identified in the Administrative Controls Chapter of the ITS. In addition, prior to Amendments 189 (Unit 1) and 175 (Unit 2), dated February 10, 1995, these requirements were located in the LCO sections of the Technical Specifications. Amendments 189 (Unit 1) and 175 (Unit 2) relocated the Radiological Effluents Technical Specification from the Technical Specifications to other plant controlled documents, and added CTS 6.8.4.a to the CTS. Since this change is a clarification required to maintain provisions that would be allowed in the LCO sections of the Technical Specifications, it is considered administrative in nature. This change is designated as administrative because it does not result in technical changes to the CTS. A.3 CTS 4.0.5.b does not include all of the required Surveillance Frequencies for performing inservice testing activities. ITS 5.5.6.a adds a new required Frequency of "Biennially or every 2 years." This changes the CTS by adding a new Frequency to the required Frequencies for performing inservice testing activities. This change is acceptable because the change does not include any new requirements, but only provides clarification of required Frequencies for performing inservice testing activities. Therefore, this change is considered administrative. This change is designated as administrative because it does not result in technical changes to the CTS. A.4 CTS 4.0.5.d states that the performance of the above testing activities shall be in addition to other specified Surveillance Requirements. ITS 5.5.6 does not include a similar statement. This changes the CTS by deleting the statement. CTS 4.0.5.d restates that all applicable requirements must be met. Repeating this overall requirement as a specific detail is redundant and unnecessary. Therefore, this detail can be omitted without any technical change in the CNP Units 1 and 2 Page 1 of 16 Attachment 1, Volume 16, Rev. 1, Page 136 of 256
Attachment 1, Volume 16, Rev. 1, Page 137 of 256 DISCUSSION OF CHANGES ITS 5.5, PROGRAMS AND MANUALS requirements and is considered administrative in nature. This change is designated as administrative because it does not result in technical changes to the CTS. A.5 CTS 4.0.5 specifies the requirements for the Inservice Testing Program, however there is no statement whether the provisions of CTS 4.0.3 are applicable. ITS 5.5.6.c states that the provisions of SR 3.0.3 are applicable to the inservice testing activities. This changes the CTS by adding the allowances of ITS SR 3.0.3 to the Technical Specification Inservice Testing Program requirements. This statement is needed to maintain allowances for Surveillance Frequency extensions contained in the ITS since ITS SR 3.0.3 is not normally applied to Frequencies identified in the Administrative Controls Chapter of the ITS. Since this change is a clarification required to maintain provisions that would be allowed in the LCO sections of the Technical Specifications, it is considered administrative in nature. This change is designated as administrative because it does not result in a technical change to the CTS. A.6 CTS 4.4.5.1, 4.4.5.2, 4.4.5.3, and 4.4.5.4, including Table 4.4-1 and 4.4-2, specify the requirements for the steam generator tube surveillance testing activities. Inthe ITS, these requirements are included as ITS 5.5.7, "Steam Generator (SG) Program," and a generic statement describing the program has been included. In addition, a statement has been added which states that the provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Steam Generator Program test Frequencies. This changes the CTS by adding a generic description of the program and specifically stating that the allowances of ITS SR 3.0.2 and SR 3.0.3 are applicable to the Steam Generator Program. The ITS SR 3.0.2 and SR 3.0.3 statement is needed to maintain allowances for Surveillance Frequency extensions contained in the ITS since ITS SR 3.0.2 and SR 3.0.3 are not normally applied to Frequencies identified in the Administrative Controls Chapter of the ITS. Since this change is a clarification required to maintain provisions that are allowed in the CTS (since CTS 4.0.2 and CTS 4.0.3 apply to the Surveillances of CTS 3/4.4.5), it is considered acceptable. In addition, the generic statement describing the program is also acceptable. This change is designated as administrative because it does not result in technical changes to the CTS. A.7 CTS 4.6.1.2 requires the performance of containment leakage rate testing in accordance with 10 CFR 50 Appendix J Option B, except as modified by NRC-approved exemptions, and Regulatory Guide 1.1.63, dated September 1995. CTS 4.6.1.2 is also modified by two exceptions. CTS 4.6.1.2.b states that the requirements of Specification 4.0.2 are not applicable. CTS 4.6.1.3.a contains a requirement to perform air lock testing in accordance with 10 CFR 50 Appendix J Option B and Regulatory Guide 1.163, dated September 1995. ITS 5.5.14.a requires a program to establish 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, as modified by the listed exceptions. ITS 5.5.14.e states that the provision of SR 3.0.3 are CNP Units 1 and 2 Page 2 of 16 Attachment 1, Volume 16, Rev. 1, Page 137 of 256
Attachment 1, Volume 16, Rev. 1, Page 138 of 256 DISCUSSION OF CHANGES ITS 5.5, PROGRAMS AND MANUALS applicable to the Containment Leakage Rate Testing Program. This changes the CTS by including the requirements of CTS 4.6.1.2 and 4.6.1.3 in a program, adding the statement that the provisions of SR 3.0.3 are applicable to the Containment Leakage Rate Testing Program, and deleting the statement that the provisions of Specification 4.0.2 are not applicable. This change is acceptable because no changes have been made to the existing requirements. The CTS and proposed ITS 5.5.14 continue to require the same testing to be performed. The statement associated with CTS 4.0.2 is not needed since the Frequency extensions of ITS SR 3.0.2 are not applied to Frequencies identified in the Administrative Controls Section of the ITS, unless specifically identified. The statement associated with ITS SR 3.0.3 is needed to maintain allowances for Surveillance Frequency extensions contained in the ITS since ITS SR 3.0.2 and SR 3.0.3 are not applied to Frequencies identified in the Administrative Controls Chapter of the ITS, unless specifically identified. Since these changes are clarifications required to maintain provisions that would be allowed in the LCO sections of the Technical Specifications, it is considered administrative in nature. This change is designated as administrative because it does not result in technical changes to the CTS. A.8 CTS 4.7.5.1.c, 4.7.6.1.b, and 4.9.12.b require the performance of ventilation filter testing "following painting, fire, or chemical release in any ventilation zone communicating with the system." ITS 5.5.9 requires the performance of the same ventilation filter testing "following painting, fire, or chemical release in any ventilation zone communicating with the system while it is in operation that could adversely affect the filter bank or charcoal adsorber capability." This changes the CTS by requiring the filter testing to be performed only if the associated system was in operation and the painting, fire, or chemical release is considered significant enough to adversely affect the filter bank or charcoal adsorber capability. The purpose of ITS 5.5.9 is to ensure that ventilation filter testing is only performed when there is a potential adverse impact on the affected filter. Current CNP practice is that not all painting, fire, or chemical release results in the need to perform certain ventilation filter tests. Only painting, fire, or chemical release that could affect the functional capability of the ventilation filter trains (i.e., that are significant) would require performance of the tests. The words "that could adversely affect the filter bank or charcoal adsorber capability" were added for clarity and consistency with current practice to avoid a misinterpretation that any painting, fire, or chemical release (such as using a small can of paint to do touch-up work) would result in the need to perform the tests. Similarly, the wording "while it is in operation" was added to clarify that this is the time when the painting, fire, or chemical release could be communicating with the system. This clarification is administrative, and is consistent with other ITS submittals. In addition, the NRC in a letter to Entergy Operations, Inc., dated September 11, 1997, supported the clarification that not all painting, fires, or chemical releases required the filter trains to be tested. Furthermore, this clarification is also consistent with Regulatory Guide 1.52, Revision 3. This change is designated as administrative because it does not result in technical changes to the CTS. CNP Units I and 2 Page 3 of 16 Attachment 1, Volume 16, Rev. 1, Page 138 of 256
Attachment 1, Volume 16, Rev. 1, Page 139 of 256 DISCUSSION OF CHANGES ITS 5.5, PROGRAMS AND MANUALS A.9 The Surveillances (CTS 4.7.5.1.c, 4.7.5.1.d, 4.7.5.1.e.1, 4.7.5.1.f, and 4.7.5.1.g) associated with the ventilation filter testing for the Control Room Emergency Ventilation (CREV) System, the Surveillances (CTS 4.7.6.1.b, 4.7.6.1.c, 4.7.6.1.d.1, 4.7.6.1.e, and 4.7.6.1.f) associated with the ventilation filter testing for the Engineered Safety Features (ESF) Ventilation System, and the Surveillances (CTS 4.9.12.b, 4.9.12.c, 4.9.12.d.1, 4.9.12.e, and 4.9.12.0 associated with the filter testing for the Fuel Handling Area Exhaust Ventilation (FHAEV) System have been placed in a program in the proposed Administrative Controls Chapter 5.0 (ITS 5.5.9). As such, a general program statement has been added as ITS 5.5.9. Also, a statement of the applicability of ITS SR 3.0.2 and SR 3.0.3 is needed to clarify that the allowances for Surveillance Frequency extension do apply. This changes the CTS by moving the ventilation filter testing Surveillances associated with the CREV, ESF Ventilation, and FHAEV Systems to a program in ITS 5.5 and specifically stating the applicability of ITS SR 3.0.2 and SR 3.0.3 in the program. The addition of the program statement is acceptable because it is describing the intent of the CTS Surveillances. The addition of the ITS SR 3.0.2 and SR 3.0.3 statement is a clarification needed to maintain provisions that are currently allowed in the LCO and SR sections of the CTS, therefore it is considered acceptable. This change is designated as administrative because it does not result in technical changes to the CTS. A.10 The Surveillances associated with diesel fuel oil testing (CTS 4.8.1.1.2.c and d) have been placed in a program in the proposed Administrative Controls Chapter 5.0 (ITS 5.5.11). As such, a general program statement has been added as ITS 5.5.11. Also, a statement of the applicability of ITS SR 3.0.2 and SR 3.0.3 is needed to clarify that the allowances for Surveillance Frequency extension do apply. This changes the CTS by moving the diesel fuel oil testing Surveillances to a program in ITS 5.5 and specifically stating the applicability of ITS SR 3.0.2 and SR 3.0.3 in the program. The addition of the program statement is acceptable because it is describing the intent of the CTS Surveillances. The addition of the ITS SR 3.0.2 and SR 3.0.3 statement is a clarification needed to maintain provisions that are currently allowed in the LCO and SR sections of the CTS, therefore i is considered acceptable. This change is designated as administrative because it does not result in technical changes to the CTS. A.1 1 The liquid holdup tank requirements in CTS 3/4.11.1, the explosive gas mixture requirements in CTS 3/4.11.2.1, and the gas storage tank requirements in CTS 3/4.11.2.2 have been placed in a program in the proposed Administrative Controls Chapter 5.0 (ITS 5.5.10). As such, a general program statement has been added. Also, a statement of applicability of ITS SR 3.0.2 and SR 3.0.3 is needed to clarify that the allowances for Surveillance Frequency extensions do apply. This changes the CTS by moving the liquid holdup tank, explosive gas mixture, and gas storage tank requirements to a program in ITS 5.5.10 and specifically stating the applicability of ITS SR 3.0.2 and SR 3.0.3 in the program. The addition of the program statement is acceptable because it is describing the intent of the CTS Specifications. The addition of the ITS SR 3.0.2 and SR 3.0.3 CNP Units I and 2 Page 4 of 16 Attachment 1, Volume 16, Rev. 1, Page 139 of 256
Attachment 1, Volume 16, Rev. 1, Page 140 of 256 DISCUSSION OF CHANGES ITS 5.5, PROGRAMS AND MANUALS statement is a clarification needed to maintain provisions that are currently allowed in the LCO and SR sections of the CTS, therefore it is considered acceptable. This change is designated as administrative because it does not result in technical changes to the CTS. A.12 CTS 6.8.1.g requires written procedures to be established, implemented and maintained covering the activities of the component cyclic or transient limits program, which provides controls to track the UFSAR Section 4.1, cyclic and transient occurrences to ensure that components are maintained within the limits. ITS 5.5.4 requires a program to track the UFSAR, Section 4.1 cyclic and transient occurrences to ensure that components are maintained within the design limits. This changes the CTS by placing the requirements of the Component Cyclic or Transient Limits Program currently located in the procedure section of the CTS Administration Controls Chapter into the Program section of the ITS Administrative Controls Chapter. One purpose of CTS 6.8.1.g is to ensure that there is a program to track the UFSAR, Section 4.1 cyclic and transient occurrences to ensure that components are maintained within the design limits. Since this change is a clarification that CTS 6.8.1.g also requires a program to be established, it is considered acceptable. This change is designated as administrative because it does not result in technical changes to the CTS. A.13 CTS 4.4.10.1 requires the inspection of each reactor coolant pump flywheel. ITS 5.5.5 requires a program to provide for the inspection of each reactor coolant pump flywheel. In addition, a statement has been added which states the provisions of ITS SR 3.0.2 and SR 3.0.3 are applicable to the Reactor Coolant Pump Flywheel Inspection Program Surveillance Frequency. This changes the CTS by including the requirements of CTS 4.4.10.1 in a program in the Administrative Controls Chapter of the Technical Specifications instead of as a Surveillance and specifically stating that the allowances of ITS SR 3.0.2 and SR 3.0.3 are applicable to the Reactor Coolant Pump Flywheel Inspection Program Surveillance Frequency. Other changes to 3/4.4.10.1 is discussed in the Discussion of Changes for CTS 3/4.4.10.1. This change is acceptable because no changes have been made to the existing requirements. The CTS and proposed ITS 5.5.5 continue to require the same reactor coolant pump flywheel inspections to be performed. The ITS SR 3.0.2 and SR 3.0.3 statement is needed to maintain allowances for Surveillance Frequency extensions contained in the CTS because ITS SR 3.0.2 and SR 3.0.3 are not normally applied to Frequencies identified in the Administrative Controls Chapter of the ITS. Since this change is a clarification required to maintain provisions that are allowed in the CTS (since CTS 4.0.2 and CTS 4.0.3 apply to the Surveillances of CTS 3/4.4.10), it is considered acceptable. This change is designated as administrative because it does not result in technical changes to the CTS. A.14 (Unit 1only) CTS 4.4.5.4.a does not contain a definition for Preservice Inspection. ITS 5.5.7.d.1.i) includes the definition. This changes the Unit 1 CTS by adding a definition for Preservice Inspection. CNP Units 1 and 2 Page 5 of 16 Attachment 1, Volume 16, Rev. 1, Page 140 of 256
Attachment 1, Volume 16, Rev. 1, Page 141 of 256 DISCUSSION OF CHANGES ITS 5.5, PROGRAMS AND MANUALS CTS 4.4.5.2.b, 4.4.5.3.a, and Table 4.4-1 (ITS 5.5.7.a.2, ITS 5.5.7.e.1, and ITS Table 5.5.7-1) refer to a preservice inspection. This proposed change is acceptable because the definition is consistent with the definition for preservice inspection in CTS 4.4.5.4.a.9 for Unit 2, and because ITS 5.5.7.a.2, 5.5.7.e.1, and Table 5.5.7-1 continue to refer to the preservice inspections. This change is designated as administrative because it does not result in technical changes to the CTS. A.15 CTS 4.0.5 requires pump and valve testing per the requirements of Section XI of the ASME Boiler and Pressure Vessel Code. ITS 5.5.6 requires pump and valve testing per the requirements of the ASME Operation and Maintenance Standards and Guides (OM Codes). This changes the CTS by referring to the ASME OM Codes instead of ASME Boiler and Pressure Code, Section XI. In the 1987 Addenda to the 1986 edition of ASME Boiler and Pressure Vessel Code, Section Xl, the requirements for Inservice Testing were removed and relocated to the ASME/ANSI OM Codes. This change was endorsed in 10CFR50.55a. 10CFR50.55a(f) now addresses the requirements for inservice testing using the ASME/ANSI OM Codes and 10CFR50.55a(g) addresses the requirements for inservice inspection using ASME Boiler and Pressure Vessel Code, Section XI. The ITS has been revised to incorporate the current Code requirements. In addition, the terms weekly, monthly, and semiannually are not used in the applicable ASME/ANSI OM Codes. Therefore, these Frequencies have been deleted. This change is designated as administrative because it does not result in technical changes to the CTS. MORE RESTRICTIVE CHANGES M.1 License Conditions 2.H (Unit 1) and 2.G (Unit 2) provide the requirements for a System Integrity program. The program is not explicit as to which systems outside containment must be monitored. ITS 5.5.2 includes the requirements for the Leakage Monitoring Program and provides a list of systems that should be monitored because they could contain highly radioactive fluids during a serious transient or accident. The purpose of the Leakage Monitoring Program is to minimize leakage from those portions of systems outside containment that could contain highly radioactive fluids during a serious transient or accident to levels as low as practicable. The systems added to the Specification include the Safety Injection System, Chemical and Volume Control System, Residual Heat Removal System, Containment Spray System, post accident sampling, and the boron injection tank injection flowpath of the Centrifugal Charging System. The change is acceptable because these systems are currently monitored to satisfy the current License Conditions and is a complete list of those systems that could contain highly radioactive fluids during a serious transient or accident. This change is designated as more restrictive because it adds an explicit list of systems to the Technical Specifications. M.2 The CTS does not include program requirements for a Safety Function Determination Program or Battery Monitoring and Maintenance Program. The CNP Units 1 and 2 Page 6 of 16 Attachment 1, Volume 16, Rev. 1, Page 141 of 256
Attachment 1, Volume 16, Rev. 1, Page 142 of 256 DISCUSSION OF CHANGES ITS 5.5, PROGRAMS AND MANUALS ITS includes programs for these activities. This changes the CTS be adding the following programs: ITS 5.5.13, "Safety Function Determination Program (SFDP)"; and ITS 5.5.15, "Battery Monitoring and Maintenance Program." The Safety Function Determination Program is included to support implementation of the support system OPERABILITY characteristics of the Technical Specifications. The Battery Monitoring and Maintenance Program is included to provide for battery restoration and maintenance. The specific wording associated with these two programs may be found in ITS 5.5.13 and ITS 5.5.15. The changes are acceptable because they support implementation of the requirements of the ITS and the UFSAR. This change is designated as more restrictive because it imposes additional programmatic requirements in the Technical Specifications. RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES LA. 1 (Type 6 - Removal of LCO, SR, or other TS requirement to the TRM, UFSAR, ODCM, QAPD, or I1P) CTS 6.8.4.b, "Radiological Environmental Monitoring Program," describes a program to monitor the radiation and radionuclides in the environs of the plant. ITS Chapter 5.0 does not require such a program. This changes the CTS by moving the requirements for the Radiological Environmental Monitoring Program to the Offsite Dose Calculation Manual (ODCM). The purpose of CTS 6.8.4.b is to provide representative measurements of radioactivity in the highest potential exposure pathways, and verification of the accuracy of the effluent monitoring program and modeling of environmental exposure pathways. The removal of the requirement for this program from the Technical Specifications is acceptable because this type of information is not necessary to be included in the Technical Specifications to provide adequate protection of public health and safety. ITS 5.6.2 still requires an annual report of the results of the "Radiological Environmental Monitoring Program." Also, this change is acceptable because these types of procedural details will be adequately controlled in the ODCM. Changes to the ODCM are controlled by the ODCM change control process in ITS 5.5.1, which ensures changes are properly evaluated. This change is designated as a less restrictive removal of requirement change because the requirements for a program are being removed from the Technical Specifications. LA.2 (Type 6 - Removal of LCO, SR, or other TS requirement to the TRM, UFSAR, ODCM, QAPD, or lIP) Operating License Conditions 2.1 (Unit 1) and 2.H (Unit 2) specify that the Iodine Monitoring Program shall be implemented and provides a description of what the program shall include. ITS 5.5 does not include this CNP Units 1 and 2 Page 7 of 16 Attachment 1, Volume 16, Rev. 1, Page 142 of 256
Attachment 1, Volume 16, Rev. 1, Page 143 of 256 DISCUSSION OF CHANGES ITS 5.5, PROGRAMS AND MANUALS program. This changes the CTS by moving the details of the Iodine Monitoring Program to the Technical Requirements Manual (TRM). The removal of this requirement from the Technical Specifications is acceptable because this type of information is not necessary to be included in the Technical Specifications to provide adequate protection of public health and safety. This program is required by the CNP Units 1 and 2 commitment to NUREG-0578, Item 2.1.8.c, as stated in a letter from R.S. Hunter (AEP) to Harold R. Denton (NRC) dated December 10, 1980. The program is designed to minimize radiation exposure to plant personnel in vital areas of the plant after an accident, and has no impact on nuclear safety or the health and safety of the public. The training aspect of the program is accomplished as part of the continual training program for personnel in the cognizant organizations, as well as during training for those individuals responsible for implementing the radiological emergency planning procedures. Provisions for monitoring and performing maintenance of the sampling and analysis equipment are addressed in chemistry and radiation protection procedures. This change is acceptable because the program requirements will be adequately controlled in the TRM. Any changes to the TRM are made under 10 CFR 50.59, which ensures changes are properly evaluated. This change is designated as a less restrictive removal of requirement change because requirements are being removed from the Technical Specifications. LA.3 (Type 6 - Removal of LCO, SR, or other TS requirement to the TRM, UFSAR, ODCM, QAPD, or lIP) CTS 4.0.5 provides requirements for the Inservice Inspection Program. The ITS does not include Inservice Inspection Program requirements. In addition, since the Inservice Testing Program is the only requirement remaining, the reference to ASME Code Class 1, 2, and 3 "components" has been changed to "pumps and valves" for clarity. Pumps and valves are the only components related to the Inservice Testing Program (as described in CTS 4.0.5.a). This changes the CTS by moving these requirements from the Technical Specifications to the Inservice Inspection Program (lIP). The removal of these requirements is acceptable because this type of information is not necessary to be included in the Technical Specifications to provide adequate protection of public health and safety. The Technical Specifications still retain requirements for the affected components to be OPERABLE. Also, this change is acceptable because these requirements will be adequately controlled by the IlP, which is required by 10 CFR 50.55a. Compliance with 10 CFR 50.55a is required by the CNP Units 1 and 2 Operating Licenses. This change is designated as a less restrictive removal of requirement change because requirements are being removed from the Technical Specifications. LA.4 (Type 3 - Removing Procedural Details for Meeting TS Requirements or Reporting Requirements) CTS 4.0.5.a specifies that the Inservice Testing of ASME Code Class 1, 2, and 3 pumps and valves shall be performed in accordance with Section Xl of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR 50, Section 50.55a. ITS 5.5.6 states that the Inservice Testing Program provides controls for inservice testing of ASME Code Class 1, 2, and 3 pumps and valves. This changes the CTS by CNP Units 1 and 2 Page 8 of 16 Attachment 1, Volume 16, Rev. 1, Page 143 of 256
Attachment 1, Volume 16, Rev. 1, Page 144 of 256 DISCUSSION OF CHANGES ITS 5.5, PROGRAMS AND MANUALS moving these procedural details from the Technical Specifications to the Inservice Testing Program. The removal of these details for meeting Technical Specification requirements from the Technical Specifications is acceptable because this type of information is not necessary to be included in the Technical Specifications to provide adequate protection of public health and safety. The ITS still retains requirements for the control for inservice testing of ASME Code Class 1, 2, and 3 pumps and valves. Also, this change is acceptable because these types of details will be adequately controlled in the plant controlled Inservice Testing Program. Changes to the Inservice Testing Program will be controlled by the provisions of 10 CFR 50.55a. This change is designated as a less restrictive removal of detail change because the details for meeting Technical Specification requirements are being removed from the Technical Specifications. LA.5 (Type 3 - Removing Procedural Details for Meeting TS Requirements or Reporting Requirements) CTS 4.7.5.1.c.3, 4.7.5.1.d.1, 4.7.5.1.d.2, 4.7.6.1.b.4, 4.7.6.1.c.1, 4.7.6.1.c.2, 4.9.12.b.4, 4.9.12.c.1, and 4.9.12.c.2 require that within 31 days after removal of a carbon sample the laboratory analysis results are shown to be within limit. ITS 5.5.9.c requires the same analysis to be performed however the detail of "within 31 days after removal of a carbon sample" is not included. This changes the CTS by moving these procedural details from the Technical Specifications to the Technical Requirements Manual (TRM). The removal of these details for performing Surveillance Requirements from the Technical Specifications is acceptable because this type of information is not necessary to be included in the Technical Specifications to provide adequate protection of public health and safety. The ITS still retains the requirement to perform the testing at the appropriate Frequencies. Also, this change is acceptable because these types of procedural details will be adequately controlled in the TRM. Any changes to the TRM are made under 10 CFR 50.59, which ensures changes are properly evaluated. This change is designated as a less restrictive removal of detail change because procedural details for meeting Technical Specification requirements are being removed from the Technical Specifications. LA.6 (Type 3 - Removing Procedural Details for Meeting TS Requirements or Reporting Requirements) CTS 4.8.1.1.2.c, 4.8.1.1.2.c.1), 4.8.1.1.2.c.1)a), 4.8.1.1.2.c.1)b), 4.8.1.1.2.c.2), 4.8.1.1.2.c.3), 4.8.1.1.2.c.4), and 4.8.1.1.2.d specify test and sampling requirements for new diesel fuel oil and diesel fuel oil in the storage tanks in accordance with certain ASTM standards (i.e., D4057-81, D975-81, D1298-80, D4176-82, D2622-82, and D2276-83) and provide limits for kinematic viscosity, flash point, API gravity, absolute specific gravity, and specific gravity. ITS 5.5.11 does not include either the explicit reference to the ASTM standards or the specific limits, but continues to require the verification that the new and stored diesel fuel oil is tested in accordance with the applicable standards and that the parameters are within limits. This changes the CTS by moving the procedural details on the testing requirements and the specific limits to the Bases of ITS 3.8.3. CNP Units 1 and 2 Page 9 of 16 Attachment 1, Volume 16, Rev. 1, Page 144 of 256
Attachment 1, Volume 16, Rev. 1, Page 145 of 256 DISCUSSION OF CHANGES ITS 5.5, PROGRAMS AND MANUALS The removal of these details for performing Surveillance Requirements from the CTS is acceptable because this type of information is not necessary to be included in the Technical Specifications to provide adequate protection of public health and safety. The ITS still retains requirement to determine that new and stored diesel fuel oil are within the applicable limits. Also, this change is acceptable because these types of procedural details will be adequately controlled in the ITS Bases. Changes to the Bases are controlled by the Technical Specification Bases Control Program in Chapter 5. This program provides for the evaluation of changes to ensure the Bases are properly controlled. This change is designated as a less restrictive removal of detail change because procedural details for meeting Technical Specification requirements are being removed from the CTS. LA.7 (Type 3 - Removing Procedural Details for Meeting TS Requirements or Reporting Requirements) CTS 3/4.11.1 includes the details for implementing the requirements for the liquid holdup tank. CTS 3/4.11.2.1 includes the details for implementing the requirements for the explosive gas mixture. CTS 3/4.11.2.2 includes the details for implementing the requirements for the gas storage tank. The details for implementing these requirements, including the specific limits, are not included in the ITS. The ITS only includes a requirement to maintain a program for these requirements. This changes the CTS by moving these procedural details for implementing the requirements, including the specific limits, from the Technical Specifications to the Technical Requirements Manual (TRM). The removal of these details for the specific limits, Applicability, Actions, and Surveillance Requirements from the Technical Specifications is acceptable because this type of information is not necessary to be included in the Technical Specifications to provide adequate protection of public health and safety. ITS 5.5.10 still retains the requirement to include a program, which 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 temporary liquid storage tanks. Also, this change is acceptable because these types of procedural details will be adequately controlled in the TRM. Any changes to the TRM are made under 10 CFR 50.59, which ensures changes are properly evaluated. This change is designated as a less restrictive removal of detail change because procedural details for meeting Technical Specification requirements are being removed from the Technical Specifications. LA.8 (Type 3 - Removing Procedural Details for Meeting TS Requirements or Reporting Requirements) CTS 6.14.1.a requires changes to the ODCM to be documented and records of reviews performed to be retained as required by the Quality Assurance Program Description, Appendix C, Section 6.10.2.n. CTS 6.14.1.b requires changes to the ODCM to be effective after review and acceptance by the PORC and the approval of the plant manager. ITS 5.5.1.c.1 requires changes to the ODCM to be documented and records of reviews performed to be retained. ITS 5.5.1.c.2 requires changes to the ODCM to become effective after the approval of the plant manager. This changes the CTS by moving the record retention requirement reference and the PORC review and approval requirement to the Quality Assurance Program Description (QAPD). CNP Units 1 and 2 Page 10 of 16 Attachment 1, Volume 16, Rev. 1, Page 145 of 256
Attachment 1, Volume 16, Rev. 1, Page 146 of 256 DISCUSSION OF CHANGES ITS 5.5, PROGRAMS AND MANUALS The removal of these details, which are related to meeting Technical Specification requirements, from the Technical Specifications is acceptable because this type of information is not necessary to be included in the Technical Specifications to provide adequate protection of public health and safety. ITS 5.5.1 still retains the requirement for changes to the ODCM. Also, this change is acceptable because these types of procedural details will be adequately controlled in the QAPD. Any changes to the QAPD are made under 10 CFR 50.54(a), which ensures changes are properly evaluated. This change is designated as a less restrictive removal of detail change because procedural details for meeting Technical Specification requirements are being removed from the Technical Specifications. LA.9 (Type 3 - Removing Procedural Details for Meeting TS Requirements or Reporting Requirements) CTS 6.14.1.b uses the title "Plant Manager." ITS 5.5.1.c.2 uses the generic title "plant manager." This changes the CTS by moving the specific CNP organizational title to the UFSAR and replacing it with a generic title. The removal of these details, which are related to meeting Technical Specification requirements, from the Technical Specifications is acceptable because this type of information is not necessary to be included in Technical Specifications to provide adequate protection of public health and safety. The allowance to relocate the specific CNP organizational title out of the Technical Specifications is consistent with the NRC letter from C. Grimes to the Owners Groups Technical Specification Committee Chairmen, dated November 10, 1994. The various requirements of the plant manager are still retained in the ITS. Also, this change is acceptable because the removed information will be adequately controlled in the UFSAR. Any changes to the UFSAR are made under 10 CFR 50.59 or 10 CFR 50.71(e), which ensures changes are properly evaluated. This change is designated as a less restrictive removal of detail change because information related to meeting Technical Specification requirements are being removed from the Technical Specifications. LESS RESTRICTIVE CHANGES L.1 (Category 10 - 18 to 24 Month Surveillance Frequency Change, Non-Channel Calibration Type) License Conditions 2.H (Unit 1)and 2.G (Unit 2) specify that the integrated leak test requirements for each system outside containment that would or could contain highly radioactive fluids during a serious transient or accident must be performed at a frequency not to exceed refueling cycle intervals. ITS 5.5.2 specifies that the same test must be performed at least once per 24 months and an allowance has been added which states that the provisions of ITS SR 3.0.2 are applicable. This changes the CTS by extending the Frequency of the Surveillance from 18 months (i.e., the current CNP normal refueling cycle interval) to 24 months (i.e., a maximum of 30 months accounting for the allowable grace period specified in ITS SR 3.0.2). The purpose of License Conditions 2.H (Unit 1) and 2.G (Unit 2) is to ensure the leakage from systems outside containment that would or could contain highly radioactive fluids during a serious transient or accident is reduced to as low as CNP Units 1 and 2 Page 11 of 16 Attachment 1, Volume 16, Rev. 1, Page 146 of 256
Attachment 1, Volume 16, Rev. 1, Page 147 of 256 DISCUSSION OF CHANGES ITS 5.5, PROGRAMS AND MANUALS practicable levels. This change was evaluated in accordance with the guidance provided in NRC Generic Letter No. 91-04, "Changes in Technical Specification Surveillance Intervals to Accommodate a 24-Month Fuel Cycle," dated April 2, 1991. Reviews of historical surveillance data and maintenance data sufficient to determine failure modes have shown that these tests normally pass their Surveillances at the current Frequency. An evaluation has been performed using this data, and it has been determined that the effect on safety due to the extended Surveillance Frequency will be minimal. Extending the Surveillance test interval for the System Integrity integrated leak test verification SR is acceptable because most portions of the subject systems included in this program are visually walked down, while the plant is operating, during plant testing, and/or operator/system engineer walkdowns. In addition, housekeeping/safety walkdowns also serve to detect any gross leakage. If leakage is observed from these systems, corrective actions will be taken to repair the leakage. Finally, the plant radiological surveys will also identify any potential sources of leakage. These visual walkdowns and surveys provide monitoring of the systems at a greater frequency than once per refueling cycle, and support the conclusion that the impact, if any, on safety is minimal as a result of the proposed changes. Based on the inherent system and component reliability and the testing performed during the operating cycle, the impact, if any, from this change on system availability is minimal. The review of historical surveillance data also demonstrated that there are no failures that would invalidate this conclusion. In addition, the proposed 24 month Surveillance Frequency, if performed at the maximum interval allowed by ITS SR 3.0.2 (30 months) does not invalidate any assumptions in the plant licensing basis. This change is designated as less restrictive because Surveillances will be performed less frequently under the ITS than under the CTS. L.2 (Category I - Relaxation of LCO Requirements) CTS 3.6.1.2.a specifies that the overall integrated leakage rate shall be limited to < L.. CTS 3.6.1.2.b specifies that combined leakage rate shall be limited to < 0.60 L. for all penetrations and valves subject to Types B and C tests. However, the CTS 3.6.1.2 Action does not allow the unit to increase Reactor Coolant System temperature above 2000 F if either the measured overall integrated leakage rate exceeds 0.75 L, or if the measured combined leakage rate for all penetrations and valves subject to Type B and C tests exceeds 0.60 L.. ITS 5.5.14 specifies that the containment leakage rate acceptance criterion is 1.0 La and that during the first unit startup following testing in accordance with this program, the leakage rate acceptance criteria are < 0.60 L, for the Type B and C tests and < 0.75 L, for Type A tests. This changes the CTS by only requiring the 0.60 La and 0.75 La limits to be met during the first unit startup following testing in accordance with the Containment Leakage Rate Testing Program. The purpose of ITS 5.5.14 is to ensure the appropriate limits are specified for the Containment Leakage Rate Testing Program. This change is acceptable because the acceptance limits continue to ensure the containment leakage is within the value assumed in the accident analysis. Currently, the overall integrated leakage rate of < La and the combined leakage rate of < 0.6 La applies in MODES 1, 2, 3, and 4. The CTS 3.6.1.2 Action will not allow the unit to enter MODE 4 from MODE 5 unless the integrated leakage rate is < 0.75 La and the combined leakage rate for all penetrations and valves subject to Types B and C CNP Units I and 2 Page 12 of 16 Attachment 1, Volume 16, Rev. 1, Page 147 of 256
Attachment 1, Volume 16, Rev. 1, Page 148 of 256 DISCUSSION OF CHANGES ITS 5.5, PROGRAMS AND MANUALS tests is < 0.60 La. In the ITS, the containment leakage rate acceptance criterion is < 1.0 L. and is applicable in MODES 1, 2, 3, and 4. The other limits (i.e.,
< 0.60 La and < 0.75 A) are only applicable during the first unit startup following testing in accordance with this program. This will allow subsequent unit startups (after the first unit startup following testing in accordance with the program) to proceed as long as the containment leakage rate acceptance criterion of < 1.0 L is met. This is acceptable because the leakage limit of La is assumed in the accident analysis. This change is designated as less restrictive because less stringent LCO requirements are being applied in the ITS than were applied in the CTS.
L.3 (Category 10-18 to 24 Month Surveillance Frequency Change, Non-Channel Calibration Type) CTS 4.7.5.1.c, 4.7.5.1.e.1, 4.7.6.1.b, 4.7.6.1.d.1, 4.9.12.b, and 4.9.12.d.1 require the performance of ventilation filter testing once per 18 months. ITS 5.5.9 requires these same Surveillances to be performed once per 24 months. This changes the CTS by extending the Frequency of the Surveillance from 18 months (i.e., a maximum of 22.5 months accounting for the allowable grace period specified in CTS 4.0.2 and ITS SR 3.0.2) to 24 months (i.e., a maximum of 30 months accounting for the allowable grace period specified in CTS 4.0.2 and ITS SR 3.0.2). The purpose of CTS 4.7.5.1.c, 4.7.5.1.e.1, 4.7.6.1.b, 4.7.6.1.d.1, 4.9.12.b, and 4.9.12.d.1 is to ensure that the Control Room Emergency Ventilation (CREV) System, the Engineered Safety Features (ESF) Ventilation System, and the Fuel Handling Area Exhaust Ventilation (FHAEV) System charcoal adsorbers and HEPA filters can perform their safety function. This change was evaluated in accordance with the guidance provided in NRC Generic Letter No. 91-04, "Changes in Technical Specification Surveillance Intervals to Accommodate a 24-Month Fuel Cycle," dated April 2, 1991. Reviews of historical surveillance data and maintenance data sufficient to determine failure modes have shown that these tests normally pass their Surveillances at the current Frequency. An evaluation has been performed using this data, and it has been determined that the effect on safety due to the extended Surveillance Frequency will be minimal. Extending the Surveillance test interval for the HEPA filter dioctyl phthalate (DOP) tests, the charcoal adsorber halogenated hydrocarbon refrigerant tests, the laboratory analysis test, and the flow test is acceptable since other tests may be required to be performed during the operating cycle. Tests described in ITS 5.5.9.a (the HEPA filter test) and 5.5.9.b (the charcoal adsorber halogenated hydrocarbon), shall be performed once per 24 months; after each complete or partial replacement of the HEPA filter bank or charcoal adsorber bank; subsequent to reinstalling the adsorber tray used for obtaining the carbon sample, after any structural maintenance on the HEPA filter bank or charcoal adsorber bank housing; and, following painting, fire, or chemical release in any ventilation zone communicating with the subsystem while it is in operation that could adversely affect the filter bank or charcoal adsorber capability. Tests described in ITS 5.5.9.c (laboratory test of the charcoal sample) shall be performed once per 24 months; after 720 hours of adsorber operation, after any structural maintenance on the HEPA filter or charcoal adsorber bank housing; and, following painting, fire, or chemical release in any ventilation zone communicating with the subsystem while it is in operation that could adversely affect the filter bank or charcoal adsorber capability. The additional Surveillance CNP Units 1 and 2 Page 13 of 16 Attachment 1, Volume 16, Rev. 1, Page 148 of 256
Attachment 1, Volume 16, Rev. 1, Page 149 of 256 DISCUSSION OF CHANGES ITS 5.5, PROGRAMS AND MANUALS Frequencies are adequate to ensure the filters remain OPERABLE during the cycle. Tests described in ITS 5.5.9.d (combined pressure drop across the combined HEPA filter and charcoal adsorbers) shall be performed once per 24 months. The CREV, ESF Ventilation, and FHAEV Systems are required to be tested every 46 days on a STAGGERED TEST BASIS (CREV and ESF Ventilation Systems) or 92 days (FHAEV System) for > 15 minutes. This testing ensures that a significant portion of the associated ventilation system is operating properly and will detect significant failures. Based on the inherent system and component reliability and the testing performed during the operating cycle, the impact, if any, from this change on system availability is minimal. The review of historical surveillance data also demonstrated that there are no failures that would invalidate this conclusion. In addition, the proposed 24 month Surveillance Frequency, if performed at the maximum interval allowed by ITS SR 3.0.2 (30 months) does not invalidate any assumptions in the plant licensing basis. This change is designated as less restrictive because Surveillances will be performed less frequently under the ITS than under the CTS. L.4 (Category 5- Deletion of Surveillance Requirement) CTS 4.7.5.1.d.2 requires the performance of a halogenated hydrocarbon refrigerant gas test on the CREV System charcoal adsorber and a DOP test on the CREV System HEPA filter banks after the reinstallation of the adsorber tray used for obtaining a carbon sample. CTS 4.7.6.1.b.4 and 4.7.6.1.c.2 require the performance of a halogenated hydrocarbon refrigerant gas test on the ESF Ventilation System charcoal adsorber after the reinstallation of the adsorber tray used for obtaining a carbon sample. CTS 4.9.12.b.4 and 4.9.12.c.2 require the performance of a halogenated hydrocarbon refrigerant gas test on the FHAEV System charcoal adsorber after the reinstallation of the adsorber tray used for obtaining a carbon sample. ITS 5.5.9 does not contain these explicit post maintenance testing requirements. This changes the CTS by deleting these explicit post maintenance requirements. The purpose of CTS 4.7.5.1.d.2,4.7.6.1.b.4,4.7.6.1.c.2,4.9.12.b.4, and 4.9.12.c.2 is to verify the OPERABILITY of the ventilation filter trains after the reinstallation of the adsorber tray used for taking a carbon sample. This change is acceptable because the deleted Surveillance Requirements are not necessary to verify the equipment used to meet the LCO can perform its required function. Thus, appropriate equipment continues to be tested in a manner and at a Frequency necessary to give confidence that the equipment can perform its assumed safety function. Any time the OPERABILITY of a system or component has been affected by repair, maintenance, modification, or replacement of a component, post maintenance testing is required to demonstrate the OPERABILITY of the system or component. This is described in the Bases for ITS SR 3.0.1 and required under ITS SR 3.0.1. The OPERABILITY requirements for the affected ventilation filter trains are described in the Bases for ITS 3.7.10, 3.7.12, and 3.7.13. In addition, the requirements of 10 CFR 50, Appendix B, Section Xl (Test Control) provide adequate controls for test programs to ensure that testing incorporates applicable acceptance criteria. Compliance with 10 CFR 50, Appendix B is required under the Units I and 2 Operating Licenses. As a result, post maintenance testing will continue to be performed and an explicit requirement in the Technical Specifications is not necessary. In addition, ITS 5.5.9 requires the performance of ITS 5.5.9.a (a halogenated hydrocarbon CNP Units 1 and 2 Page 14 of 16 Attachment 1, Volume 16, Rev. 1, Page 149 of 256
Attachment 1, Volume 16, Rev. 1, Page 150 of 256 DISCUSSION OF CHANGES ITS 5.5, PROGRAMS AND MANUALS refrigerant gas test on the charcoal adsorber) and ITS 5.5.9.b (a DOP test on the HEPA filter banks) after any structural maintenance on the HEPA filter bank or charcoal adsorber bank housing. Therefore, if after the reinstallation of the adsorber tray used for obtaining a carbon sample it is determined that ITS 5.5.9.a or 5.5.9.b are not met, the applicable ITS SRs must be declared not met and the appropriate Required Actions must be entered. Therefore, although the explicit Surveillance Frequency has been deleted, both ITS SR 3.0.1 and ITS 5.5.9 will require the performance of these tests if it is determined that the Surveillances may not be satisfied after reinstallation of the adsorber trays. This change is designated as less restrictive because Surveillances which are required in the CTS will not be required in the ITS. L.5 (Category 7 - Relaxation Of Surveillance Frequency, Non-24 Month Type Change) CTS 4.8.1.1.2.c.4 requires the evaluation that certain diesel fuel oil properties are within the appropriate limits within 31 days of obtaining the sample. ITS 5.5.11.b requires this same evaluation to be performed within 31 days following addition of the new fuel oil to the storage tanks. This changes the CTS by changing the time by which the evaluation for these properties must be completed. The purpose of ITS 5.5.11 .b is to ensure that the properties of the new diesel fuel oil added to the storage tanks are acceptable. This change is acceptable because the new Surveillance Frequency has been evaluated to ensure that it provides an acceptable level of equipment reliability. CTS 4.8.1.1.2.c.4 requires the evaluation that certain diesel fuel oil properties are within the appropriate limits within 31 days of obtaining the sample, while the ITS time limit begins after the fuel oil is added to the storage tanks. The new fuel oil can affect the stored fuel oil only when it is added to the storage tanks. Failure to meet the limit for these other fuel oil properties would not have an immediate effect on diesel generator operation because the oil added is normally only a small portion of the entire fuel oil storage volume. The 31 day period is also acceptable because the fuel oil properties of interest, even if they were not within stated limits, would not have an immediate effect on diesel generator operation. This change is designated as less restrictive because Surveillances will be performed less frequently under the ITS than under the CTS. L.6 (Category I - Relaxation of LCO Requirements) Operating License Conditions 2.C.(7) (Unit 1) and 2.C.(3)(v) (Unit 2) specify that the Secondary Water Chemistry Monitoring Program shall be described in the station chemistry manual and provides a description of what the manual should contain. ITS 5.5.8 does not specify that the program must be described in the station chemistry manual. It only states what shall be included in the Secondary Water Chemistry Program. This changes the CTS by deleting the details of where the description of the Secondary Water Chemistry Program shall reside from the Technical Specifications. The purpose of the Secondary Water Chemistry Program is to ensure proper controls are placed on monitoring secondary water chemistry in order to inhibit steam generator tube degradation. The change is acceptable because the Technical Specifications still retain the requirement to have a Secondary Water Chemistry Program and the Technical Specifications continue to describe the CNP Units 1 and 2 Page 15 of 16 Attachment 1, Volume 16, Rev. 1, Page 150 of 256
Attachment 1, Volume 16, Rev. 1, Page 151 of 256 DISCUSSION OF CHANGES ITS 5.5, PROGRAMS AND MANUALS contents of the program. Thus, the Technical Specifications continue to control the general content of the program and any changes will still require NRC approval. In addition, removal of this detail for meeting Technical Specification requirements (i.e., the actual location of the program) from the Technical Specifications is acceptable because this type of information is not necessary to be included in the Technical Specifications to provide adequate protection of public health and safety. This change is designated as less restrictive because less stringent Technical Specifications requirements are being applied in the ITS than were applied in the CTS. CNP Units 1 and 2 Page 16 of 16 Attachment 1, Volume 16, Rev. 1, Page 151 of 256
Attachment 1, Volume 16, Rev. 1, Page 152 of 256 Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs) Attachment 1, Volume 16, Rev. 1, Page 152 of 256
Attachment 1, Volurme 16, Rev. 1, Page 153 of 256 Programs and Manuals cr5 5.5 5.0 ADMINISTRATIVE CONTROLS 5.5 Programs and Manuals 769Lf The following programs shall be established, Implemented, and maintained. 5.5.1 Offsite Dose Calculation Manual 1ODCM)
- a. The ODCM shall contain the methodology and parameters used in the calculation of offsite doses resulting from radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring alarm and trip selpoints, and In the conduct of the radiological environmental monitoring prograro>nd
- b. The ODCM shall also contaln'the radioactive effluent controls and radiological environmental monitoring activIties, and descriptions of the Information that should be included in the Annual Radiological
. G (D' I
Environmental Operatinggand Radioactive Effluent Release Reports required by SpeoficationJ5.6.2band Specification 15.6.3
)Licensee initiated changes to the ODCM: 0 I
(.30 Shall be documented and records of reviews performed shall be retained. This documentation shall coritain: Sufficient Information to support the change(s) together with the appropriate analyses or evaluations Justifying the change(s)and A determination that the change(s) maintain the levels of radioactive effluent control required by 10 CFR 20.1302. 40 CFR 190, 10 CFR 50.36a, and 10 CFR 50. Appendix I. and not adiversely Impact the accuracy or reliability of effluent, dose, or setpoint calculationsi Shall become effective after the approval of the plant m anagj Shall be submitted to the NRC In the form of a complete, legible copy of th entire ODCM as a part of or concurrent with the Radioactive Effluent Release Report for the period of the report In which any change In the ODCM was made. Each change shall be Identified by markings In the margin of the affected pages, clearly Indicating the area of the page that was changed, and shall Indicate the date (i.e., month and year) the change was Implemented. WOG STS 5.5 - 1 Rev. 2. 04/30/01 Attachment 1, Volume.16, Rev. 1, Page 153 of 256
Attachment 1, Volume 16, Rev. 1, Page 154 of 256
.:A
_cTl Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.2 ts rv Coolant-snif) (14.4t1/ ,ceFe.1 L L;CeaJ)Se. This program provides controls to minimize leakage from those portions of Cco, A ,d, 4 ) 2. 6- systems outside containment that could contain highly radioactive fluids during a serious transient or accident to levels as low as practicable. The systems ,/j r1 include gna Safety Injectio Chemical and Volume Contro Th s uet
- a. Preventive maintenance and periodic visual inspection requirementsnd
- b. Integrated leak test requirements for each system at least once per months.
The provisions of SR 3.0.2 are applicable.
... 5.3 Post A I cident Sari;; /
- REVIEWER'S NOTE -
This program y be eliminated based on the implementation WCAP-14986, Rev. 1, "Post cident Sampling System Requirements: A Te nical Basis," and the associate NRC Safety Evaluation dated June 14, 2000. This pro ra provides controls that ensure the capability tobtain and analyze reactor c nt, radioactive gases, and particulates in plan gaseous effluents and conta ment atmosphere samples under accident co itions. The program shall inclde the following:
- a. Tr fling of personnel,
- b. Pocedures for sampling and analysis, and C. rovisions for maintenance of sampling and a is equipment.
i.5.Q Radioactive Effluent Controls Procram This program conforms to 10 CFR 50.36a for the control of radioactive effluents and for maintaining the doses to members of the public from radioactive effluents as low as reasonably achievable. The program shall be contained in the ODCMK shall be implemented by procedures, and shall include remedial actions to be taken whenever the program limits are exceeded. The program shall include the following elements: WOG STS 5.5-2 Rev. 2. 04130/01 Attachment 1, Volume 16, Rev. 1, Page 154 of 256
Attachment 1, Volume 16, Rev. 1, Page 155 of 256 5.5 0 INSERT I Residual Heat Removal System, Containment Spray System, post accident sampling, and the boron injection tank injection flowpath of the Centrifugal Charging System Insert Page 5.5-2 Attachment 1, Volume 16, Rev. 1, Page 155 of 256
Attachment 1, Volume 16, Rev. 1, Page 156 of 256 Programs and Manuals 5.5 C7S 5.5 Programs and Manuals N ma Ni.!5. Radioactive Effluent Controls Program (continued) 09 W a. Limitations on the functional capability of radioactive liquid and gaseous monitoring instrumentation Including surveillance tests and setpoint determination In accordance with the methodology in the ODC
- b. Limitations on the concentrations of radioactive material released In liquid effluents to unrestricted areas, conforming toetulhe concentration values In Appendix B, Table 2, Column 2 to 10 CFR 20.1001-20.2402
- c. Monitoring, sampling, and analysis of radioactive liquid and gaseous effluents In accordance with 10 CFR 20.1302 and with the methodology and parameters In the ODCA
- d. Limitations on the annual and quarterly doses or dose commitment to a member of the public from radioactive materials in liquid effluents released areas, conforming to 10 CFR 50, Appendix e.Determination of cumulativ ose contributions from radioactive effluents forIhecurentcalnda qurte an curen clendar year in accordance' with the methodolg an prmtsInhe ODCM at least every 31 day A -rmnao-np~r~ce ose contributions ry raioactiemunsm
%ztcgLdaagedththe mreAhndolQov in tho QQQkeatlst Aery 31 >
- f. Limitations on the functional capability and use of the liquid and gaseous effluent treatment systems to ensure that appropriate portions of these 6-tq. systems are used to reduce releases of radioactivity when the projected doses in a period of 31 days would exceed 2% of the guidelines for the annual dose or dose commitment, conforming to 10 CFR 50, Appendix
- g. Limitations on the dose rate resulting from radioactive material released in gaseous effluents from the site to areas beyond the site boundary shall be In accordance with the following:
- 1. For noble gases: a dose rate &500 mrem/yr to the whole body and a dose rate 5 3000 mrem/yr to the skiad
- 2. For iodine-131, iodine-133, tritium, and all radionuclides in particulate form with half-lives greater than 8 days: a dose rate s 1500 mremlyr to any orgarb.
- h. Limitations on the annual and quarterly air doses resulting from noble gases released In gaseous effluents from each unit to areas beyond the site boundary, conforming to 10 CFR 50, Appendix 3
WOG STS 5.5-3 Rev. 2, 04130/01 Attachment 1, Volume 16, Rev. 1, Page 156 of 256
Attachment 1, Volume 16, Rev. 1, Page 157 of 256 Programs and Manuals _cr 5.5 5.5 Programs and Manuals 5.5.42 Radioactive Effluent Controls Program (continued) 00 6.'.q A'-) I. Limitations on the annual and quarterly doses to a member of the public from lodine-131, iodine-133, tritium, and all radionuclides in particulate form with half lives > 8 days in gaseous effluents released from each unit to areas beyond the site boundary, conforming to 10 CFR 50. Appendix 9,X-cto) L1
- j. Limitations on the annual dose or dose commitment to any member of the public, beyond the site boundary, due to releases of radioactivity and to radiation from uranium fuel cycle sources, conforming to 40 CFR 190.
boc A,% The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Effluent Controls Program,. rveillance/equer*'-(quen 5.5@.i) This program provides controls to track theIFSAR transient occurrences to ensure that components design limits. m trovides controls for monitoring any tendon deg 4nete containments, induding effectiveness of its n dium, to ensure containment structural integrity.
§Iine measurements prior to initial operations. Tlw Program, g Inspection frequencies, and acceptafcE with [Regulatory Guide 1.35, Revision 3, 1989, ins of SR 3.0.2 and SR 3.0.3 are applicable ted the Tendon
>Program inspection frequencies.] /
Reactor Coolant Pump Flywheel Inspection Program This program shall provide for the Inspection of each reactor coolant D-c' A.s WOG STS 5.5 -4 Rev. 2. Attachment 1, Volume 16, Rev. 1, Page 157 of 256
Attachment 1, Volume 16, Rev. 1, Page 158 of 256 5.5 O INSERT 1A The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Reactor Coolant Pump Flywheel Inspection Program Surveillance Frequency. Insert Page 5.5-4 Attachment 1, Volume 16, Rev. 1, Page 158 of 256
Attachment 1, Volume 16, Rev. 1, Page 159 of 256 Programs and Manuals cr7-. 5.5 5.5 Programs and Manuals Reactor Coolant Pump Flywheel Inspe/n Program (continued) (5.5.7 _. / _ Prgacniud _ _ _
- EWER'S NOTES -
- 1. The inspection interval d scope for RCP flywheels stated above can be applied to plants that s sfy the staff requirements In the safety evaluation of Topical Report, W AP-14535A, 'Topical Report on Reactor Coolant Pump Flywheel Insp ction Elimination."
- 2. Ucensees shall c firm that the flywheels are made of SA 533 B material.
Further, licensee having Group-15 flywheels (as determined in WCAP-14535A, "Topi Report on Reactor Coolant Pump Flywheel Inspection Elimination") n ed to demonstrate that material properties of their A51 material is eq ivalent to SA 533 B material, and its reference temper ure, 2 RT, is less t n 30 'F.
- 1 3. For flywhe not made of SA 533 B orA516 material, license need to either de onstrate that the flywheel material properties are unded by those of A 533 B material, or provide the minimum specifid ultimate (1
tensile ress, the fracture toughness, and the reference mperature, RTNOrr that material. For the latter, the licensees s uld employ these mater I propertes, and use the methodology Inthe I ical report, as exte ed In the two responses to the staffs RAI, to rovide an assessment to j tify a change In Inspection schedule for their ants.
- 4. LI ensees with Group-10 flywheels need to co rm that their flywheels have adequate shrink fit to preclude loss of shri fit of the flywheel at the aximum overspeed, or to provide an evalu ion demonstrating that no etrimental effects would occur if the shrin it was lost as maximum overspeed.
- s. ,
5.5A Inservice Testing Proaram This program provides controls for inservice testing of ASME Code Class 1, 2, and 3 Q.he program sal iiudethe a Afo.' .L ) .e Testingquenciesspecified in nditheASME i) (i)@) a vestinge VessEo and applicable Addenda as follows:, S t
. y ndapplicabledded equired Frequencies for r/ Addenda terminology for performing Inservice testing (448'ke) l [inservice testing activities activities e Qikly s Ras once per 7 ))
.GO>
WOG STS 5.5-5 Rev. 2, 04130101 Attachment 1, Volume 16, Rev. 1, Page 159 of 256
Attachment 1, Volume 16, Rev. 1, Page 160 of 256 Programs and Manuals CJ5 5.5 Programs and Manuals', 5.54 Inservice Testing Program (continued) AS (i (2~ Co dpplcable Required Frequencies for Addenda terminology for pei forming Inservice testing Inservice testing activities ' act ivities I. CM-olwv -lAeast once per 311 Quarterly or every 3 months At least once per 92 days emiannualy or eve east onr er 184 days EVv 9 mnnthc Yearly or annually Lance e days/ At least once per 366 days 0 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 activitie ij s I£Y.OL A*C
- c. The provisions of SR 3.0.3 are applicable to Inservice testing activitiesj a IZ. e
- d. Nothing in the ASME Pssure"Vessel C shall be construed to supersede the requirements of any TS.
Steam Generator (S Pr o Program rt I --- q.4-5-% 4 1-5-2-) 7- REVIEWER'S NOTE -
'O .,r .-. . q-% 9'. .f.7- The Licensee's currenicensing basis steam generato ube surveillance . 1W., requirements shall b relocated from the LCO and In ded here. An appropriate adminis t'ive controls program format ould be used.
tI . Thegrvison o SR3.. re lcalo o heSG
- DOL- A- The rovisions of SR 3.0.&arepplicable to the SG lCutsuV en^Progr testequencies. A A3 Secondary Water Chemistry Proaram 6 (uLL,ce ire-4 This program provides controls for monitoring secondary water chemistry to inhibit SG tube degradation tnd low oresiure turfinedi-ir stab corrnsinon CwJ;.6oi ). C.Q)(Y) 42~r~he program shall Include:
- a. Identification of a sampling schedule for the critical variables and control points for these variabled 0
- b. Identification of the procedures used to measure the values of the critical variablesles? (D WOG STS '. 5.5- 6 Rev. 2. 04130101 Attachment 1, Volume 16, Rev. 1, Page 160 of 256
Attachment 1, Volume 16, Rev. 1, Page 161 of 256 5.5 Oi) INSERT 2 This program provides requirements for steam generator tube sample selection and inspection. Each steam generator shall be determined OPERABLE during shutdown by selecting and inspecting at least the minimum number of steam generators specified in Table 5.5.7-1. The steam generator tube minimum sample size, inspection result classification, and the corresponding action required shall be as specified in Table 5.5.7-2. The inservice inspection of steam generator tubes shall be performed at the Frequencies specified in Specification 5.5.7.c and the inspected tubes shall be verified acceptable per the acceptance criteria of Specification 5.5.7.d.
- a. The tubes selected for each inservice inspection shall include at least 3% of the total number of tubes in all 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 inservice inspection (subsequent to the preservice inspection) of each steam generator shall include:
a) All nonplugged tubes that previously had detectable wall penetrations greater than or equal to 20%; b) Tubes in those areas where experience has indicated potential problems; and c) A tube inspection pursuant to Specification 5.5.7.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. The tubes selected as the second and third samples (if required by Table 5.5.7-2) during each inservice inspection may be subjected to a partial tube 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; and b) The inspections include those portions of the tubes where imperfections were previously found.
- b. The results of each sample inspection shall be classified into one of the following three categories:
Insert Page 5.5-6a Attachment 1, Volume 16, Rev. 1, Page 161 of 256
Attachment 1, Volume 16, Rev. 1, Page 162 of 256 5.5 Q INSERT 2 (continued) Category Inspection Results C-I Less than 5% of the total tubes inspected are degraded tubes and none of the inspected tubes are defective. C-2 Greater than or equal to 5% and less than or equal to 10% of the total tubes inspected are degraded tubes or one or more tubes, but not more than 1% of the total tubes inspected, are defective. 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 (greater than or equal to 10%) further wall penetrations to be included in the above percentage calculations.
- c. The above required inservice inspections of steam generator tubes shall be performed at the following Frequencies:
- 1. The first inservice inspection shall be performed after 6 Effective Full Power Months but within 24 calendar months of initial criticality or replacement of steam generators. Subsequent inservice 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 All Volatile Treatment 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 Frequency may be extended to a maximum of once per 40 months.
- 2. If the results inservice inspection of a steam generator conducted in accordance with Table 5.5.7-2 at 40 month intervals fall in Category C-3, the inspection Frequency shall be increased to once per 20 months. The increase in inspection Frequency shall apply until a subsequent inspection satisfies the criteria of Specification 5.5.7.c.1, at which time the Frequency may be extended to a maximum of once per 40 months; and
- 3. Additional, unscheduled inservice inspections shall be performed on each steam generator in accordance with the first sample inspection specified in Table 5.5.7-2 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 LCO 3.4.13; Insert Page 5.5-6b Attachment 1, Volume 16, Rev. 1, Page 162 of 256
Attachment 1, Volume 16, Rev. 1, Page 163 of 256 5.5 O INSERT 2 (continued) b) A seismic occurrence greater than the Operating Basis Earthquake; c) A loss of coolant accident requiring actuation of the engineered safety features; or 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 an imperfection greater than or equal to 20% of the nominal wall thickness caused by degradation: d) Percent 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 plugging limit. A tube containing a defect is defective; f) Plugging Limit means the imperfection depth at or beyond which the tube shall be removed from service. Any tube which, upon inspection, exhibits tube wall degradation of 40% or more of the nominal tube wall thickness shall be plugged prior to returning the steam generator to service; 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 main steam line or feedwater line break, as specified in Specification 5.5.7.c.3 above; 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 to the cold leg; and i) Preservice Inspection means an inspection of the full length of each tube in the steam generator performed by eddy current techniques prior to service to establish a baseline condition of the tubing. This inspection shall be performed after the field hydrostatic test and prior to initial entry into MODE I using the equipment and techniques expected to be used during subsequent inservice inspections. Insert Page 5.5-6c Attachment 1, Volume 16, Rev. 1, Page 163 of 256
Attachment 1, Volume 16, Rev. 1, Page 164 of 256 5.5 O@ INSERT 2 (continued)
- 2. The steam generator shall be determined OPERABLE after completing the corresponding actions (plugging all tubes exceeding the plugging limit and all tubes containing through-wall cracks) required by Table 5.5.7-2.
Insert Page 5.5-6d Attachment 1, Volume 16, Rev. 1, Page 164 of 256
Attachment 1, Volume 16, Rev. 1, Page 165 of 256 5.5 INSERT2A Table 5.5.7-1 (page 1 of 1) Minimum Number of Steam Generators to be Inspected During Inservice Inspection Preservice Inspection Yes Number of Steam Generators per Unit 4 First Inservice Inspection 2 Second and Subsequent Inservice Inspections 1(a) (a) The third and fourth steam generators not inspected during the first inservice inspection shall be inspected during the second and third inspections, respectively. The fourth and subsequent inspections may be limited to one steam generator on a rotating schedule encompassing 3 N% of the tubes (where N is the number of steam generators in the plant) if the results of the first or previous inspections indicate that all steam generators are performing in a like manner. Note that under some circumstances, the operating conditions in one or more steam generators may be found to be more severe than those in other steam generators. Under such circumstances the sample sequence shall be modified to inspect the most severe conditions. Insert Page 5.5-6e Attachment 1, Volume 16, Rev. 1, Page 165 of 256
Attachment 1, Volume 16, Rev. 1, Page 166 of 256 5.5 INSERT 2A (continued) Table 5.5.7-2 (page 1 of 1) Steam Generator (SG) Tube Inspection First Sample Inspection Second Sample Inspection Third Sample Inspection Sample Size I Result Action IRequired Result j Action Required Result Action I Required A minimum of C-1 None NA NA NA NA S tubes per C-2 Plug defective C-1 None NA NA SG tubes and C-2 Plug defective C-1 None inspect tubes and C-2 Plug additional 2S inspect defective tubes in this additional 4S tubes SG tubes in this C-3 Perform SG action for C-3 result of first sample C-3 Perform action NA NA for C-3 result of first sample C-3 Inspect all All other None NA NA tubes in this SGs are C-SG, plug 1 defective Some SGs Perform action NA NA tubes, inspect are C-2, for C-2 result 2S tubes in but no for second each other additional sample SG, and notify SGs are NRC pursuant C-3 I I to Additional Inspect all NA NA Specification SG is C-3 tubes in each 5.6.7 SG, plug or repair defective tubes, and notify NRC pursuant to Specification
5.6.7 Where
S = 3 (N/n)%; N is the number of SGs in the unit; and n is the number of SGs inspected during an inspection. Insert Page 5.5-6f Attachment 1, Volume 16, Rev. 1, Page 166 of 256
Attachment 1, Volume 16, Rev. 1, Page 167 of 256 Programs and Manuals 5.5 Li,41 i ~er 5.5 .ogamend Manuals Pr .- &#ba - Ct~J~jO~ 2.(7) Scndaryj Water Chemsr Program (continued) i
- c. Identification of process sampling poin ir ude moniton
- d. Procedures for the recording and management of data
- e. Procedures defining corrective actions for all off control point chemistry )
condition~n
- f. A procedure Identifying the authority responsible for the Interpretation of the
~ . ~ 5.5. Vetlto ile etn Porm(FP hall( establish gtii~eihthe( ~required testing of
~'~lb,.. ngnere Sfety Feature (ESF) filter ventilatin systernim Ie)-
1 h ~j ,specee
~ nMeguaoulatoru 52
~~ ~J LJJ) Rvso M 5018, dA
.q.-Z z 4 7.$-L CN) 4/7-J-dj
.q.'7. 1,.1,31j -Z6.)J. C WA 0.
.;-jq,II. , - )IMIL
, t.!rj IYI,)L. 9 eI
$J I
. %-),51 1.C.1)17-Irlci 4,73-1-box)
- 1, 1,11-A-2." 1-1-10, c.
KiK Demonstrate for each of the ESF systems that a laboratory test of a ~- of the charcoal adsorber. when obtained s sr 4..-7S-1-C-3j4-7S'I-A) ,shows the methyl Iodide penetration less tha 0 11f) q.I 9.1. CA) - vale seciiedbelw wen tested in accordance with ASTM D3803-' V-1, it. 0) V1 C. aeatue tm o 30C (6 OF) and the relative humidityspecifled belc 4 WOG STS -5.5-7 Attachment 1, Volume 16, Rev. .1, Page'17o 167 of 256 5
Attachment 1, Volume 16, Rev. 1, Page 168 of 256 5.5 (i) INSERT 3 Tests described in Specifications 5.5.9.a and 5.5.9.b shall be performed once per 24 months; after each complete or partial replacement of the HEPA filter bank or charcoal adsorber bank; after any structural maintenance on the HEPA filter bank or charcoal adsorber bank housing; and, following painting, fire, or chemical release in any ventilation zone communicating with the subsystem while it is in operation that could adversely affect the filter bank or charcoal adsorber capability. Tests described in Specification 5.5.9.c shall be performed once per 24 months; after 720 hours of adsorber operation; after any structural maintenance on the HEPA filter bank or charcoal adsorber bank housing; and, following painting, fire, or chemical release in any ventilation zone communicating with the subsystem while it is in operation that could adversely affect the charcoal adsorber capability. Tests described in Specification 5.5.9.d shall be performed once per 24 months. I The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the VFTP test Frequencies. INSERT3A removal efficiency of > 99% of the dioctyl phthalate (DOP) C;) INSERT 4 ESF Ventilation ANSI Standard Flowrate (cfm) System CREV System N510-1975 > 5,400 and < 6,600 ESF Ventilation N510-1980 > 22,500 and < 27,500 System FHAEV System N510-1980 > 27,000 and < 33,000 INSERT4A removal efficiency of > 99% of a halogenated hydrocarbon refrigerant test gas Insert Page 5.5-7a Attachment 1, Volume 16, Rev. 1, Page 168 of 256
Attachment 1, Volume 16, Rev. i, Page 169 of 256 5.5 E(D INSERT 5 ESF Ventilation ANSI Standard Flowrate (cfm) System CREV System N510-1975 > 5,400 and < 6,600 ESF Ventilation -N510-1980 > 22,500 and < 27,500 System FHAEV.System N510-1980 .> 27,000 and c 33,000 . -, . ::-. ,,;INSERT5A :; from either at least one test canister or at least two carbon samples removed from one of the charcoal adsorbers Insert Page 5.5-7b Attachment 1, Volume 16, Rev. 1, Page 169 of 256
-Attachmrbent 1, Vo Iumef "16 Rv IPg 70o 5
.:.Programs and Manuals P.rograms and Manuals.. .
55.&Ventilation Fllter Testng Program '(continued) 7*St I.Wj K SF Yentilatin Sse Penetration .. -'R. FaceYoct IL : N teL Revi 'e rs<N t]N T
.- wte .
REIEWR NOTE-The use of any standard ot'P rthaAT D30 99to tef te aca sample may resultl anc rsimaion of the capabilt othchar o6l to adsorb ravldn sarsl th bityothe S charcoal filte'rs to perf In a manner cosstent with the'lies bas for the facility Is indeterminat ASTM D 3803 1989 Is ore'tige tetngsadard becaus It does not differentiate, between u dand new charcoal, It has a longer eq libratlon period performed atOtmear f3 C(6F) and a relativ h dity (RH) of 95/0 (r70%h RH with humid ycnrl; and It hasmore strringent tornces that. Improve repeatability o hb et
.Allowable Penetration [(1100%'- Methyl Iodide Effici!ency f Charcoal Credited
-inLicens ees Accide A Iayss) ISitfety Factor]
'When ASTM D3803 989 Isused with 30 'C (86 OF) and 9 IcRH (or '70%/ RH with humidity gontro Is uised,,the staff will accept the follow Safety factor 2rsytms with or without hJm'idity ontrol.
Hu1mIdity control c bprovIded by heaters or an NRC-a roved aiialysis thiat' demonstrates that e aie entering the charcoal will bern taied less than or' equal to 70 perce RH unrder worst-case design basi ditions.
'Ifthe system has fcveloclty greateerthani 11 0 per'ce tot 0.203 m/s (40 accident analys which wa~s reviewed an aprve the'staff In a safety 7' 5ZI-. d...Demonstrate
,.. for each of the ESF systems that the pressuredrpaosth*
combined HEPA fifte re *and the charcoal adsorbers Is less:. than the value specifl ew ntested Vn aordanc~ wi uo specified belowf+/-( 0 1 5 ,a essenlwae ;* :
*O .T .. :5.5- 8 . . Rev. 2, 04130/01 AtahetI oume 16,Rev`.1, Pa'ge 170 of 256
Attachment 1, Volume 16, Rev. 1, Page 171 of 256 5.5 INSERT 6 ESF Ventilation Face Velocity Penetration (%) RH (%) System (ffm) CREV System NA 1 95 ESF Ventilation 45.5 5 95 System FHAEV System 46.8 5 95 In addition, the carbon samples not obtained from test canisters shall be prepared by either:
- 1. Emptying one entire bed from a removed adsorber tray, mixing the adsorbent thoroughly, and obtaining samples at least two inches in diameter and with a length equal to the thickness of the bed; or
- 2. Emptying a longitudinal sample from an adsorber tray, mixing the adsorbent thoroughly, and obtaining samples at least two inches in diameter and with a length equal to the thickness of the bed.
Insert Page 5.5-8 Attachment 1, Volume 16, Rev. 1, Page 171 of 256
Attachment 1, Volume 16, Rev. 1, Page 172 of 256 ci Programs and Manuals 5.5 (s~Programs and Manuals
- 5. Ventilation Filter Testing Program (continued)
- 7. 5-a. et
.Ld ESFVentilationSystem Delta P Flowrate7 Y~zdl< [1] [ ] [ ] )
ie. Demonstrate that the Platers for each of the ESF systems dissipate the value specified belo _ 10%) when tested in acco ance with [ASME N510-1989 ESF V tilatlon System Witage I
*./ [ -] /[I
*{The provisis of SR 3.0.2 ar SR 3.0.3 are applica/e to the VFTP test A
* ~.(Irequendewt / aplce 5.5. Exglosive Gas and Storaae Tank Radioactivity Monitoring Program (I pa CA.1J This program provides controls for potentially explosive gas mixtures contained A I/. in the pWaste Gas Holdup Systems Dhe quantity of radioactivity contained In gas Z.v storage tanks r o trua .and the quantity of "gas
.3. ii. 2 *Jradioactivity contained In unprotected outdooriquid storage tank It 1.2- g aseous roavty quantities shall be dmeehil e following e ology in [Branch Techn al Postion (BTP) ETSB 1-5, 'Postulated Radioactiv Release due to Waste as System Leak or Fail e l. The liquid radwaste ntities shall be determin d in accordance with [St dard Review Plan, S VPostulat Radioactive Release du to Tank Failures"]. The program shall include: 3 a. The limits for concentrajons of hydrogen and oxygen In the Paste Gas 1%'Holdup Systemand aurveillance program to ensure the limits are maintained. Such limits shall be appropriate to the system's design criteria (i.e.. whether or not the system is designed to withstand a hydrogen explosion s6
- b. A/urvelllance program to ensure that the quantity of radioactivity contained vity
- 3. 11. . lUntach gas storage tankdind fed int, the oftg/asyeatment syteml Isless A)
<, Il, 7. 2-. than the amount that would result in a whole body exposure of 0.5 rem to any individual in an unrestricted area, in the event of slan uncontrolled c2 release of the tanks' contents&and
- c. A/urveillance program to ensure that the quantity of radioactivity contained LII I~In all outdoor liquid tanks that are not surrounded by liners, dikes, WOG STS Rev. 2, 04130101 Attachment 1, Volume 16, Rev. 1, Page 172 of 256
Attachment 1, Volume 16, Rev. 1, Page-173 of 256 5.5 ( INSERT 7 ESF Ventilation Delta P (inches Flowrate (cfm). System Water cauce) CFREV System 6 > 5,400 and < 6,600 SF Ventilation 6 > 22,500 and < 27,500 Fstem Fl--IAEV System 6 > 27,000 and < 33,000
* -..
- e . : . S ***
- S*- , , : .
Insert Page 5.5-9 Attachment 1, Volume 16, Rev. 1, Page 173 of 256 .-
Attachment 1, Volume 16, Rev. 1, Page 174 of 256 Programs and Manuals 5.5 5.5 Programs and Manuals - - '0'/
\ 5.5.R Explosive Gas and Storage Tank Radioactivity Monitoring Program (continued) 0 or wallqcapable of holding the tanks' contents and that do not have tank overflows and surrounding area drains connected to thet$.iquid Radwaste.
9.11.1 Treatment Systerr is less than the amount that would result in Y.It.1I concentrations less than the limits of 10 CFR 20, Appendix B, Table 2, Column 2, at the nearest potable water supply and the nearest surface water supply In an unrestricted area, in the event of an uncontrolled release of the tanks' contents. -00 Ai. I The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Explosive Gas and Storage Tank Radioactivity Monitoring Programiurveillance/'requencies. 5.5 Diesel Fuel Oil Testina Proaram A diesel fuel oil testing program to implement required testing of both new fuel oil ADot AS and stored fuel oil shall be established. The program shall include sampling and testing requirements, and acceptance criteria, all in accordance with applicable/ ASTM Standards. The purpose of the program Is to establish the following qI. 2 cv. I3.l2.) Accepabilityofnewfuel oil for use prior to addition to storage tanks by determining that the fuel oil has: I.z.l.2.C 1. An API gravityan abs eiicgravity imits y &i.1. CJ).*> 2. A flash poiBnd *nematic viscosity within limits s u a al.l.L
, c. a . and i
q,* 1.12 . c, i) 3. A clear and bright appearance with proper colo
- b. Within 31 days following addition of the new fuel oil to storage tanks, verify q..1, 1 . e-' Y) that the properties of the new fuel oil, other than those addressed In <
above, are within limitsfor AST)(4 2D f and
- c. Total particulate concentration of the fuel oil is s 10 mg/1 when tested every
'.8, . 4 0', 31 days In accordance with ASTM D-2276, Method r DOC A.8° The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Diesel Fuel Oil Testing Program test/requencies. 0 WOG STS 5.5- 10 Rev. 2. 04130/01 Attachment 1, Volume 16, Rev. 1, Page 174 of 256
Attachment 1, Volume 16, Rev. 1, Page 175 of 256 5.5 INSERT 8 if the gravity was not determined by comparison with the supp!ier's certification, a
- 1. ...
I. . I . I
. . I - -' . . 'l- . .
1 . ..... Insert Page 5.5-10 Attachment 1, Volume 16, Rev. 1, Page 175 of 256 I
Attachment 1, Volume 16, Rev. 1, Page 176 of 256 Programs and Manuals 5.5 5.5 Programs and Manuals t, 5.5. Technical Soecifications ITS) Bases Control Proaram &n 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.
6.8.P.L 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 A )
.1 2. A change to the updated FSAR or Bases that requires NRC approval pursuant to 10 CFR 50.59. .D6A
.1.
- c. The Bases Control Program shall contain provisions to ensure that the Bases are maintained consistent with theIE. c3: O
' {P.5.J d. Proposed changes that meet the criteria of Specification 5.5 a shall be reviewed end approved by the NRC prior to Implemen nttion. . J Changes to the Bases Implemented without prior NRC approval shall be provided to the NRC on a frequency consistent with 10 CFR 50.71(e).
boC_ tAFl Safety Function Determination Proaram (SFDP) (I) 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 Inoperabfity and corresponding exception to entering supported system Condition and Required Actions. This program Implements the requirements of LCO 3 .06.The SFDP shall contain the following: Provisions [or coss train checks to ensure a loss of the capability to perform the safety function assumed In the accident analysis does not go undeeteed m' Provisions for ensuring the function condition exist is maintained In a safe condition if a loss of Provisions to ensure that an Inoperable supported system's Completion Time is not Inappropriately extended as a result of multiple support system inoperabilitiea2nd WOG STS 5.5 - 11 Rev. 2 04/30/01 Attachment 1, Volume 16, Rev. 1, Page 176 of 256
Attachment 1, Volume 16, Rev. 1, Page 177 of 256 Programs and Manuals 5.5 Programs and Manuals Doe t1tI \s.5.9 Safety Function Determination Program (continued)
\i v Other appropriate limitations and remedial or compensatory actions.
0 (IY-A loss of safety function exists when, assuming no concurrent single failurem U'-) CmUEno concurrent loss of offsite power, or no concurrent loss of onsite diesel generator(s), a safety function assumed Inthe 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 required system redundant to the system(s) supported by the Inoperable 02 6 ( support system Isalso Inoperable _ A required system redundant to the system(s) n turn supported by the Inoperable supported system Is also inoperabl or
.
- A A required system redundant to the support system(s) for the supported systems(g) w above is also Inoperable.
(D 0 Odfunction DP identifies where a loss of safety function exists. If a loss of safety is determined to exist by this program, the appropriate Conditions and
.(1' 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.
DVo C-mA' 7 Containment Leakaae Rate Testing Program 5.5.f
- a. A program shall establi the leakage rate testing of Itl containment as required by 10 CFR 5 (5(o) and t0 CFI 50 , Appen fx J. Option A, as 1 modified by approve exemptions./
- b. The maximum allo ble containment leakage rate La, at P., shall be [I]% of containment air w ght per day.
- c. Leakage rate ac ptance criteria are:
- 1. Containm t leakage rate acceptance cri rion is s 1.0 La. During the first unit artup following testing in acco ance with this program, the leakage ate acceptance criteria are <c 0 L for the Type B and C
*1 tests aif < 0.75 L, for Type A tests.
- 2. Air bc testing acceptance criteria are WOG STS 5.5- 12 Rev. 2, 04130101 Attachment 1, Volume 16, Rev. 1, Page 177 of 256
Attachment 1, Volume 16, Rev. 1, Page 178 of 256 5.5 S INSERT 9 described in Specifications 5.5.13.b.1 and 5.5.13.b.2 I s Insert Page 5.5-12 Attachment 1, Volume 16, Rev. 1, Page 178 of 256
Attachment 1, Volume 16, Rev. 1, Page 179 of 256 Programs and Manuals 5.5 5.5 Programs and Manuals i.5.2 Containment Leak; age Rate Testing Progre_ ams (continued)(
^1 ff 7
r a) Overall ai lock leakage rate Is s[0.05 L.] when sted eat P b) For ea door, leakage rate Is s[0.01 La] wh pressurized to [O gI. i
- d. The provisions f SR 3.0.3 are applicable to the C tainment Leakage Rate Testing Prog m/
- e. Nothing in t se Technical Specifications shall constred to modify the testing Fre uencles required by 10 CFR 50. A endix J.
[OPTION B]
- a. A program shall establish 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, F-modified by the following exceptions:
13, 1 2.a, - J - IV /
- b. The calculated peak containment internal pressure for the design basis loss of coolant accident, P.; is apsif!Uhe cor iinment desi n pressu
,LLo3. 2o
- c. The maximum allowable containment leakage rate, L,, at P.. shall be of containment air weight per day.
- d. Leakage rate acceptance criteria are:
- 1. Containment leakage rate acceptance criterion is 1.0 L.. During the ILLo I k.I , 6. first unit startup following testing In accordance with this program, the leakage rate acceptance criteria are < 0.60 L, for the Type B and C tests and s 0.75 L. for Type A tests.
LCO , A"hob- 2. Air lock testing acceptance crite r (0 Overall air lock leakage rate is sJ0.05 Lhwhen tested at xP . .
*1,?.w -3 b) For each door, akage rate Is s[0.01 L when pressurized to
[a 10 psig]. WOG STS 5.5 - 13 Rev. 2, 04/30/01 Attachment 1, Volume 16, Rev. 1, Page 179 of 256
Attachment 1, Volume 16, Rev. 1, Page 180 of 256 5.5 INSERT 10 Unit 14.6.12 Note 2. Unit24.6.12Note 1 . The Type A testing Frequency specified in NEI 94-01, Revision 0, Paragraph 9.2.3, as 'at least once per 10 years based on acceptable performance history' is modified to be "at least once per 15 years based on acceptable performance history." This change applies only to the interval following the Type A test performed in October 1992 (Unit 1) and May 1992 (Unit 2). Unit 14.6.1.2 Note 1 2. (Unit 1 only) A one-time exception to the requirement to perform post-modification Type A testing is allowed for the steam generators and associated piping, as components of the containment barrier. For this case, ASME Section XI leak testing will be used to verify the leak tightness of the repaired or modified portions of the containment barrier. Entry into MODES 3 and 4 following the extended outage that commenced in 1997 may be made to perform this testing. Insert Page 5.5-13 Attachment 1, Volume 16, Rev. 1, Page 180 of 256
-Attachme'nt 1,'Volume 16, Rev. 1,-Pa 181 256' '. "
Programs and Manuals d Manuals 5.5 Program's
- 5.5.&)
an Containment Leakage Rate Testing Programn (continued).
'. o:'
7 . The provisions of SR 3.0.3 are applicable to theeContainment Leakage'Rate Testing Program. : . ' ' f.. NthingIn ti Tecnical Spec[fi ions shall be. cionstrue-d -tomdf te (1
'. testing Frequ desrequired by I 9 CFR 50, Appendix4. -j mined
- a. A program shall establish t aleakage rate testing of the contal ent as
'.. :: ' required by 10 CFR'50.54ilb and 10 CFR 50, Appendix J. ly F Aype a ~~~~and C] test requirements dIrin accordance with 10 CFR 50 Apni . Option A, as modifled by pproved exemptions. [Type B and (Type A] test
'requirements are In a dance with 10 CFR 50, Appendix J,ppion B. as.
requiredet byp1.Cvn a05 TpA~p J modified by approved e mptions. The 10 CFR 50. Appendi J Option B' test requirements shall e In accordance with the guidelines ntained In-
'Regulatory Guide 1.16 ,"Performance-Based ContainmenULeak-Test Program.' dated Sept mber, 1995 [,as modified by the foll ng
- exemptions '
*,: *, ' : l 1. .*. ] . . ' ' ' : ' '"' " ;/ ' ' :' '... ,.
;b. The calculated pea' containment Internal pressure for h design basis loss of coolant acciden P. 45 psig]. The containment d ign pressure Is
[50 psig].
- c. The maximum all wable containment leakage rate, at P.; shall be []YO of containment air eight per day.. . ' - *
- d. Leakage rate a ptance critarla are:. : /.
- 1. Containm nt leakage rate acceptance ction is s1.0 L. During'the first unIt artup following testing In accor nce with this program, the
..leakage te acceptance criteria are n 0. 0 1_ for the Type B and C s0.75 Lor Option A T ests(s 0.5 L. for Option B..
- Type A sts]. .;
- 2. Alrioc testing acceptance criteriaare
'a) verallair lock leakage rate' s .005Ljwhen tested at z'P b) o e door, leakage rate is [0 01 1L when pressurized to WOG STS . 5.5 14 Rev. 2, 4130/01 Attachment 1,Volume 16, Rev.;1, Page 181 of .256..
Attachment 1, Volume 16, Rev. 1, Page 182 of 256 Programs and Manuals 5.5 5.5 Programs and Manuals
.6Cntainment Leakage teTesting Program (continued)
- e. The provis'ns of SR 3.0.3 are applicable to the Cont nment Leakage Rat Testing Pogram. -
- f. Nothi in these Technical Specifications shall b construed to modify the testing Frequencies required by 10 CFR 50. Ap endix J.
Do.. w. 8attervonitoring and Maintenance Proaram This/rograr provides for battery restoration and maintenance, based onithe 63) 0 recommendations of IEEE Standard 450-1995. "IEEE Recommended Practice for Maintenance. Testing, and Replacement of Vented Lead-Acid Batteries for Stationary Applications," or of the battery manufacturer) including the following: U.)
- a. Actions to restore battery cells with float voltage c<2.1 3*veIn
- b. Actions to equalize and test battery cells that had been discovered with electrolyte level below the minimum established design limit.
I I I i WOG STS 5.5-15 Rev. 2, 04/30101 Attachment 1, Volume 16, Rev. 1, Page 182 of 256
Attachment 1, Volume 16, Rev. 1, Page 183 of 256 JUSTIFICATION FOR DEVIATIONS ITS 5.5, PROGRAMS AND MANUALS
- 1. The brackets are removed and the proper plant specific information/value is provided.
- 2. This Specification has been renumbered to be consistent with the ITS format and for clarity.
- 3. These punctuation corrections have been made consistent with the Writer's Guide for the Improved Standard Technical Specifications, NEI 01-03, Section 5.1.3.
- 4. The bracketed ISTS 5.5.3, Post Accident Sampling, is not included in the CNP Units 1 and 2 ITS. The requirements for Post Accident Sampling have been deleted from the CTS in License Amendments 261 (Unit 1)and 244 (Unit 2) dated January 16, 2002. Subsequent programs have been renumbered, as necessary.
- 5. Editorial changes made for enhanced clarity or to be consistent with the Writer's Guide.
- 6. Changes are made (additions, deletions, and/or changes) to the ISTS which reflect the plant specific nomenclature, number, reference, system description, analysis, or licensing basis description.
- 7. ISTS 5.5.6 provides requirements for the Pre-Stressed Concrete Containment Tendon Surveillance Program. There is no requirement for this program in the CTS.
Not including this ISTS program in the CNP Units I and 2 ITS is consistent with the CNP Units 1 and 2 licensing bases.
- 8. ISTS 5.5.7 (ITS 5.5.5) provides requirements for the Reactor Coolant Pump Flywheel Inspection Program. The allowance to perform the inspection per the recommendations of Regulatory Position C.4.b of Regulatory Guide 1.14, Revision 1, August 1975 has been deleted. This change is consistent with the CNP Units I and 2 licensing bases. The Surveillance Frequency has also been modified to be consistent with the CNP Units 1 and 2 licensing bases.
- 9. The Reviewer's Note has been deleted since it is not intended to be included in the ITS.
- 10. The Inservice Testing (IST) Program (ISTS 5.5.8) has been modified to state that the IST Program provides control for ASME Code Class 1, 2, and 3 "pumps and valves" in place of the current "components." 10 CFR 50.55a(f) provides the regulatory requirements for an IST Program. It specifies that ASME Code Class 1, 2, and 3 pumps and valves are the only components covered by an IST Program.
10 CFR 50.55a(g) provides regulatory requirements for an Inservice Inspection (ISI) Program. It specifies that ASME Code Class 1, 2, and 3 components are covered by the ISI Program, and that pumps and valves are covered by the IST Program in 10 CFR 50.55a(f. The ISTS does not include ISI Program requirements as these requirements have been relocated to a plant specific document. Therefore, the components to which the IST Program applies (i.e., pumps and valves) have been added for clarity. In addition, the statement 'The program shall include the following:" has been deleted because not all of the statements that follow are really part of the program requirements. Also, in the 1987 Addenda to the 1986 edition of ASME Boiler and Pressure Vessel Code, Section Xl, the requirements for Inservice CNP Units I and 2 Page 1 of 3 Attachment 1, Volume 16, Rev. 1, Page 183 of 256
Attachment 1, Volume 16, Rev. 1, Page 184 of 256 JUSTIFICATION FOR DEVIATIONS ITS 5.5, PROGRAMS AND MANUALS Testing were removed and relocated to the ASME/ANSI OM Code. This change was endorsed in 10 CFR 50.55a. 10 CFR 50.55a(f) now addresses the requirements for inservice testing using the ASME/ANSI OM Code and 10 CFR 50.55a(g) addresses the requirements for inservice inspection using ASME Boiler and Pressure Vessel Code, Section XI. The ITS has been revised to incorporate the current ASME/ANSI OM Code requirements. In addition, the terms weekly, monthly, semiannually, and every 9 months are not used in the ASME/ANSI OM Code and have been deleted.
- 11. Typographical/grammatical error corrected.
- 12. ISTS 5.5.9 (ITS 5.5.7) provides the requirements for the Steam Generator (SG)
Program. Consistent with the associated Reviewer's Note, the CNP Units 1 and 2 current licensing basis, reflected in CTS 4.4.5.1, 4.4.5.2, 4.4.5.3, and 4.4.5.4, for SG tube inspections are included in this program. The corresponding ISTS Reviewer's Note is deleted. The Reviewer's Note provides information for the NRC to identify acceptable methods to meet the requirements. The Reviewer's Note is not meant to be retained in the final version of the plant-specific submittal.
- 13. ISTS 5.5.10 (ITS 5.5.8) provides the requirements for the Secondary Water Chemistry Program. The program in the ISTS includes requirements to provide controls for monitoring secondary water chemistry to inhibit SG tube degradation and low pressure turbine disc stress corrosion. ITS 5.5.8 provides controls for monitoring secondary water chemistry only to inhibit SG tube degradation. This modification is consistent with the current requirements in License Condition 2.C.(7) (Unit 1) and 2.C.(3)(v) (Unit 2).
- 14. ISTS 5.5.1 0.c includes a requirement that the Secondary Water Chemistry Program identify process sampling points, "which shall include monitoring the discharge of the condensate pumps for evidence of condenser in leakage." ITS 5.5.8.c only includes the requirement that the Secondary Chemistry Program identify process sampling points and does not provide any explicit monitoring points. This change is consistent with current Operating Licensing Conditions 2.C.(7).3 (Unit 1)and 2.C.(3)(v)3 (Unit 2).
- 15. ISTS 5.5.11 (ITS 5.5.9) provides requirements for the Ventilation Filter Testing Program. ITS 5.5.9 is revised to reflect the CNP Units 1 and 2 licensing bases. The 18 month Frequencies in the CTS have been changed to 24 months in the ITS.
- 16. The following changes have been made to ISTS 5.5.13 (ITS 5.5.11):
- a. Specific gravity has been added as an option to API gravity or absolute specific gravity consistent with the current licensing basis;
- b. Saybolt viscosity has been added as an option to kinematic viscosity and the viscosity check is only required if the gravity was not determined by comparison with the suppliers certification, consistent with current licensing basis;
- c. The type of fuel oil, Type 20, has been deleted consistent with current licensing basis; and CNP Units 1 and 2 Page 2 of 3 Attachment I, Volume 16, Rev. 1, Page 184 of 256
Attachment 1, Volume 16, Rev. 1, Page 185 of 256 JUSTIFICATION FOR DEVIATIONS ITS 5.5, PROGRAMS AND MANUALS
- d. The words "ASTM D-2276 Method A-2 or A-3" in ISTS 5.5.13.c (ITS 5.5.11.c) have been changed to "ASTM D-2276 Method A" in ITS 5.5.11.c to be consistent with current licensing basis.
- 17. ISTS 5.5.16 (ITS 5.5.14) provides requirements for the Containment Leakage Rate Testing Program. The requirements of the ISTS are revised to reflect the Containment Leakage Rate Testing Program requirements of CTS 3/4.6.1.2 and 3/4.6.1.3. The containment design pressure limit specified in ISTS 5.5.16.b was not included because it currently does not exist in the CTS, and because this limit does not provide any useful input to the Containment Leakage Rate Testing Program.
The air lock door leakage test of ISTS 5.5.16.d.2.b) is not included because it is not required by the CTS. In addition, the statement in ISTS 5.5.16.f that "Nothing in these Technical Specifications shall be construed to modify the testing Frequencies required by 10 CFR 50, Appendix J" has been deleted. This phrase is not consistent with the allowances in ISTS 5.5.16.a (ITS 5.5.14.a), which states that the "program shall be in accordance with the guidelines contained in Regulatory Guide 1.163, "Performance- Based Containment Leak-Test Program," dated September, 1995, as modified by the following exceptions:" These exceptions stated in ITS 5.5.14.a are modifications to the testing Frequencies required by 10 CFR 50, Appendix J. In addition, there is no need to state any specific exception to any of the other requirements of the Specifications that discuss testing Frequencies, because the convention of application of requirements in the sections of ISTS 5.5 is that no other Specification requirements apply unless otherwise stated. For example, ISTS SR 3.0.2 does not apply to any of the ISTS 5.5 sections, unless specifically noted. Therefore, there is no need to include a statement that ITS SR 3.0.2 does not apply to the Frequencies of ITS 5.5.14.
- 18. The program details of the Explosive Gas and Storage Tank Radioactivity Monitoring Program are described in ISTS 5.5.12 (ITS 5.5.10) parts a, b, and c. Therefore, the sentence in the introductory paragraph that specifies a method to determine the explosive gas and storage tank radioactivity is not necessary.
- 19. Changes are made to ISTS 5.5.12.c (ITS 5.5.10.c) to be consistent with the first paragraph in ISTS 5.5.12 (ITS 5.5.10).
- 20. ISTS 5.5.1 1.d demonstrates that the pressure drop across the combined HEPA filters, prefilters, and charcoal adsorbers is less than the specified pressure drop when tested at the specified system flow rate. The referenced methods for performing the test, Regulatory Guide 1.52, Revision 2 and ASME N510-1989, do not provide the methods for performing this test. As a result, these test method references have been deleted in ITS 5.5.9.d.
- 21. The requirement of ISTS 5.5.7 (ITS 5.5.5) is currently located in an individual Specification in the CTS (CTS 4.4.10.1). Thus, CTS 4.0.2 (ITS SR 3.0.2) and CTS 4.0.3 (ITS SR 3.0.3) apply to the CTS Surveillance Frequency. To maintain consistency with the current licensing basis requirements, an allowance that ITS SR 3.0.2 and SR 3.0.3 are applicable to the Reactor Coolant Pump Flywheel Inspection Program Surveillance Frequency has been included in ITS 5.5.5. In addition, approved TSTF-421, which extends the Frequency to 20 years has not been adopted.
CNP Units I and 2 Page 3 of 3 Attachment 1, Volume 16, Rev. 1, Page 185 of 256
Attachment 1, Volume 16, Rev. 1, Page 186 of 256 Specific No Significant Hazards Considerations (NSHCs) Attachment 1, Volume 16, Rev. 1, Page 186 of 256
Attachment 1, Volume 16, Rev. 1, Page 187 of 256 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 5.5, PROGRAMS AND MANUALS There are no specific NSHC discussions for this Specification. CNP Units 1 and 2 Page 1 of I Attachment 1, Volume 16, Rev. 1, Page 187 of 256
, Volume 16, Rev. 1, Page 188 of 256 ATTACHMENT 6 ITS 5.6, Reporting Requirements , Volume 16, Rev. 1, Page 188 of 256 , Volume 16, Rev. 1, Page 189 of 256 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) , Volume 16, Rev. 1, Page 189 of 256
Attachment 1, Volume 16, Rev. 1, Page 190 of 256 ITS 5.6 ITS 6.0 ADMINISTRATIVE CONTROLS PROCEDURES AND PROGRAMS (Continued) 6.8.5 Technical Specifications Bases Control Proyram This programrprovides a means for processing changes to the Bases of these Technical Specifications.
- a. Changes to the Bases of the Technical Specification shall be made under appropriate administrative controls and reviews.
- b. Licensees may make changes to Bases without prior Nuclear Regulatory Commission approval See rrs]
provided the changes do not require either of the following: I. A changes in the Technical Specification incorporated in the license or
- 2. A change to the Updated Final Safety Analysis Report or Bases that requires Nuclear Regulatory Commission approval pursuant to 10 CFR 50.59.
- c. The Bases Control Program shaU contain provisions to ensure that the Bases are maintained consistent with the Updated Final Safety Analysis Report.
- d. Proposed changes that meet the criteria of Specification 6.8.5.b above shall be reviewed and approved by the Nuclear Regulatory Commission prior to implementation. Changes to the Bases implemented without prior Nuclear Regulatory Commission approval shall be provided to the Nuclear Regulatory Commission on a frequency consistent with 10 CFR 50.71(e).
5.6 6.9 REPORTING REQUIREMENTS ROUTINE REPORTS 5.6 6.9.1 In addition to the'applicable reporting requirements of Title 10. Code of Federal Regulations. the following reports shall be submitted to we egiotrator un rWe n In accordance with IO 10 CFR 50.4 STARTUP REPOR 6.9.1.1 summar report of plant st0arpand power escalation testing be submitted following (1) eipt of an operating license ) amendment to the license invlving a planned increase in wer level, (3) instalatikon of el that has a different design or been manufactured by a different fuel supplier and (4) dons that may have sigfcanrly altered the nuclear, thermal, or hydraulic perfo of the plant. 6.9.1.2 The startup report shall ad each of the tests identified in the FSAR and shall include a description of the measured vIues of the operating conditions ox characteristics obtained during the test program and a compuison of these values with desigl predictions and specifications. Any corrective actions that required to obtain satisfactory eration shall also be described. Any additional specific deqils required in license conditions faed on other commitments shall be included in this report. I COOK NUCLEAR PLANT.UNIT I Page 6-9 AMENDMENT P24-54.489,226 , 281 I Page 1 of 16 Attachment 1, Volume 16, Rev. 1, Page 190 of 256
Attachment 1, Volume 16, Rev. 1, Page 191 of 256 ITS 5.6 ITS 6.0 ADMINISTRAMiVE CONTROLS 76.9 SlARTJP Rft! RT (Continued} 6.9.L3 tartup reports shall be wthin wtid (1)90 day fo9owing completion of the startup test program, (2) 0 days following aumpn or commencement of commercial power operati ,or (3) 9 months following criticality, whichever is a. earliest.- If the Startup does not cover a he ents ( initial criticality, completion of startup te program, and resumption or mmencement of commercial power operation), e upp mentary reports be mitted at least every three months until all three ant have been completed.
.OTby April 30 (for Occupational Radiation Exposure Report) 5.6.1. 6.9.1.4 Annual reports covering the activities of the unit agtescribed below for the evious 5.6.7 calendar year shall be submitted of each year. (e i report ff shallbe iii tte priorto c h1 oftthyear owing i critcality./
6.9.1.6 Reports required on an annual basis shall include: 5.6.1 a. A tabulation on an annual basis of the number of station, utility and other personnel (including contractors) receiving annual exposures greater than Mi-). 100 mirem according to work and job functions', e.g., reactor operations and surveillance, in-service inspection, routine maintenance, special maintenance (describe maintenance), waste processing and refueling. Also included is a tabulation of the total person rem exposures for station, utility, and other personnel associated with each work and job function. The dose assignment to various 'duty functions may be estimates based on pocket dosimeter, electronic dosimeter, TLD, or film badge measurements. Small exposures totaling less than 20% of the individual total dose need not be accounted for. In the aggregate, at least 80% of the total deep dose received shall be assigned. to specific major work functions. 5.6.7 b. The complete results of steam generator tube in-service inspections performed during the report period (reference Specification 4.4.5.5.b). le. Doqpilentation of all challepes to the pressurize per operated reliefI l vives (PORVe) or safety Oales. 7/
- J ld. WI716-tion regarding la o~nes when the I-1 icific activity lmt I yvis exceeded. / /" `
5.6.1 Note ' A single submittal may be made for a multiple unit station. The submittal should combine those sections that are common to all units at the station. 5.6.1 ' Ts tabulation supplements the requirements of20.2206 of 10 CPR Part 20. I I COOK NUCLEAR PLANT-UNTr I Page 6-10 AMEDMENT&S,I3,4S4,4V,4U, 245 Page 2 of 16 Attachment 1, Volume 16, Rev. 1, Page 191 of 256
Attachment 1, Volume 16, Rev. 1, Page 192 of 256 ITS 5.6 ITS 6.0 ADMNRVE CONTROLS ANwlALERAtOLOGICAL ehmmowwm opERAT17NG REPORT1 5.6.2 6.9.1.6 lbe Annual Raioloa En mel Opsdg Re ri the qation of dw inkdung 6e prVcf alda 7 r ha be Itof euch yar. Ibe rpt sha Wde auznries,. hptdloc. and of a of ibeRadlotoa ERoeal Maeorics Progra fr di IEpnuIg pmIod lb. na p shaal be coictsn wkh tie objectes odined bh(1)the ODCM and (2)Seca IV.B.2, IV.83. and IY.C of Appendix I to 10 CPR Pan 50. t lINSERT 1 M.1 A.1 5.6.3 6.9.1.7 Ml. RIj Rado Efflcm Rese Rep wvaig Oe oprion of the unk Airim t pvOM 12 of opeat shallbe atmbed %ift90 das allr Juy I of each year. T1frep sall I- a ommy of Me qauodea of radacti & pwa ai! e efftlbea and sold 1waFte re d ftm do uwit. Tll narW prvided af be (1) couinwth tef objctde oudlmd In the ODCM and PCP and CZ) In caoone wih 10 CFR 50.36a ad Secton V.B.t of Appendix Ito 10 CFR Part 50. 5.6.2 Note. A earodcmy de ubfx Sr a Coft satio. lb. of tali i qombin tho ae(toA9 5.6.3 Note that areC to AU all rnIt ftW
- I .
I pwwimapa* shaU o Unf. COOX NUCLAR PLArrr I ae 6-li AMMMMMr A 441,154, 44,489, 22 I Page 3 of 16 Attachment 1, Volume 16, Rev. 1, Page 192 of 256
Attachment 1, Volume 16, Rev. 1, Page 193 of 256 ITS 5.6 ITS O@ INSERT I 5.6.2 The Annual Radiological Environmental Operating Report shall include the results of analyses of all radiological environmental samples and of all environmental radiation measurements taken during the period pursuant to the locations specified in the table and figures in the ODCM, as well as summarized and tabulated results of these analyses and measurements in the format of the table in the Radiological Assessment Branch Technical Position, Revision 1, November 1979. 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. Insert Page 6-11 Page 4 of 16 Attachment 1, Volume 16, Rev. 1, Page 193 of 256
Attachment 1, Volume 16, Rev. 1, Page 194 of 256 ITS 5.6 ITS
&D ADMVISTRAT11VE CONTROLS MOR7WEACKIROPERATI~IMPRWI 5.6.4 6.9.1.3 Routinreowi of operatinit sdrtistives AM -
cha tth RVs or valvesbal be submitted on a monthly basis ffQ"ffW U.YSL Nuc ar Reu ry i on n ci onrwl ME). WashnCont. dc).W u55. with a_ copy to th egional Offlc noolae than the 15th of each month following the ealendar month A.4 covered by the report. 5.6.5 CORE OPERATING LIMITS REPORT 5.6.5.a 6.9.1.9.1 Core operating limits shall be established and documented in the CORE OPERATING LIMITS REPORT before each reload cycle or any remaining purt of a reload cycle for the follow
*.7 ModeratorTcnperature Coefficnt Limits for Specircation 314.1.1 /A Reartor Core Safety Limits; SHUTDOWN MARGIN; b. /Rod D T pnmits forS cifiion 34.1.33 C. Shutdown Rod InseirtionUmits forSpecificaton 3/4.13.4.
RTS Instrumentation Overpressure AT and Overpower AT Allowable Value e. parameter values; RCS
- d. Control Rod Insertion Lknits for Specificadon 3/4.1.3.5.
Axial FIIX Diffr for Specificaion 3/4.2 1. G Pressure. Temperature, and Flow DNB Limits; and Boron f. Heat Flux Hot Channel Factor for Specification 31422. Concentration. C r , - Nuclear Enthalpy Rise Hot hannel Factor for Specification 3t4.23. and
\h Allowable Power Level for Specification 3141.6.
5.6.5.b 6.9.1.9.2 The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC In:
- a. WCAP-9272-P-A. "Westinghouse Reload Safety Evaluation Methodology. ,"
b. (Westinghouse Proprietary), WCAP-'3 r.*Pg staibution Control and Load Followng Procedures - Topical G
*Report. 8Set~pze1'J74I(Yes~inghouse Proprietary).
- c. WCAP-lO2l6-P-A. ' Relaxation of Constant Axial Offset ContbolFQ Surveillance Technical Specification,[f 1 4(WWesdnghousc Proprietary).
- d. WCAP-10266-P-A "he 1981 Version of Wcstighouse Evaluation Mode Using BASH Code,"' MRTestmnghousc Proprietary).
7k t4 WCAP-12610-P.A. "VANTAGE+ Fuel Assembly Reference Core Report.'u f li (Westinghouse Proprietary). WCAP48745-P-A, 'Design Bases for the Thermal Overpower AT and Thermal Overtemperature AT Trip Functlons,- (Westinghouse Proprietary). COOK NUCLEAR PLANT-UNIT 1 Page 6.12 AMENDMENT S°,14,4",4 20C 235. 279 Page 5 of 16 Attachment 1, Volume 16, Rev. 1, Page 194 of 256
Attachment 1, Volume 16, Rev. 1, Page 195 of 256 ITS 5.6 ITS 6.0 AD&MNSMATEVE COMIROS . CORE OPERATING LIFMS REPORT (Cownpe) 5.6.5.c 6.9.1.93 The coem penting sd *11be dutmI Isod t all oppicable bl (e.g.. ie ammrm-Mechanil Ibin, con i=MalydranC lmb, ECCS imi,. nuclar limbs socd " hutbwn m-gi~n. and tumleot and wcd&- nabyss Wumk) of fte ufafy analysis agm
=et 5.6.5.d 6.9.1.9.4 TM CORE OPATG LIM REPORT. lludleg my mWd-yvl tte&km ,
hberm lbsll be panv~ o lmae, for dscrdoad cyce. the NRCI A4 re VV5.idt tRgadP~iUMu4da p=tT--
- I I COOK NUCLEAR PLANTA-r I Pap 4.13 AMSMNMU , W4 489, 226 I Page 6 of 16 Attachment 1, Volume 16, Rev. 1, Page 195 of 256
Attachment 1, Volume 16, Rev. 1, Page 196 of 256 ITS 5.6 ITS 0 3/4 3/4.3 LIMITING CONDMONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS INSTRUMENTATION I. TABLE 3.3.6 (Continmed) TABLE NOTATION With the number of channels OPERABLE less than required by the Minimum Channels Operable See ITS LscoN0. requirement, comply with the ACTION requirements of Specification 3.A.6.1. 3.4.15 IACTION 22-With the number of channels OPERABLE less than required by the Minimum Channels Operable requirement, perform area surveys of the monitored area with portable monitoring Instrumentation at kast once per day. With the number of channels OPERABLE less than required by the Minimum Channels Operable See CTS 314.3.3.1 J S ITS requirements, comply with the ACTION requirements of Specification 3.9.9. This ACTION is not 3.36J required during the performance of containment integrated leak rate test ACTION 22A- With tbe nmber of OPERABLE Channels less than required by the Minimum Channels OPERABLE requirements: see ITS S
- 1. either restore the inoperable Channel(sj to OPERABLE status within 7 days of the event,J L 3.3.3 I 2. prepare and submit a Special Report to the Commission pursuant to Specification 6.9.2 i 5.6.6 within 14 days following the event outlining the action taken, the cause of the inoperability and the plans and schedule for restoring the system to OPERABLE status.
13- Technical Specdfica ton Section 3.0.3 is Not-App blc]be I See ITS 3.3.3 J ACTION 22B- With the number of OPERABLE Channels less than required by the Minimum Channels OPERABLE requirements.
- 1. either restore the inoperable Chann(s) to OPERABLE status within 7 days of the event, or
- 2. prepare and submit a Special Report to the Commission pursuant to Specification 6.9.2 See CTS within 14 days following the event outlining the action taken, the cause of the inoperability 3/4.3.3.1 J and the plans and schedule for restoring the system to OPERABLE status.
- 3. In the event of an accident Involving radiological releases initiate the preplanned alternate method of monitoring the appropriate parameter(s) within 72 hours.
- 4. Technical Specification Section 3.0.3 is Not Applicable.
I I I COOK NUCLEAR PLANT-UNIT I Page 314 3-37 ANIDMENT 94, 3, 68, 28 1 1 Page 7 of 16 Attachment 1, Volume 16, Rev. 1, Page 196 of 256
Attachment 1, Volume 16, Rev. 1, Page 197 of 256 ITS 5.6 ITS 3/4. REACMO COOLAN7 EWI~ 5.6.7 JL PFalowins esch bes v Invactim of sum ,atum~ n~bes. ft mtmabe of vAes planed Je each I
$at ax shaDbe repcujd
- Ca Conk wkbh15days.
- b. ?be teatS of to= p Wek vice sh&U heIhdofdA.
C- g o- f pew s . au speumWu opr n teonO BUaD
- 1. N?1uad aumofmtubesiasc I
- 2. LocaIe as!vP - of wat-dcbh1 vPeuIm 9xeach b~cad cofat hipedactdm
- 3. Idenfiif of Tes. Owegd I C. Rmb of i; 1 fts booed= which M ho !M C-3 P;4, I In NPOCWCOMO.Y.1 b I of pt= qxndm ran v. Mgmef gm -
COOMMd 10 I I dftvte ad ecoadre menam tdm to pxvc; PI AMPM.F I WOOMMUCLR WANMI 'us 2144U AMMIUM M 944 £MM~ am,aK 238 Page 8 of 16 Attachment 1, Volume 16, Rev. 1, Page 197 of 256
Attachment 1, Volume 16, Rev. 1, Page 198 of 256 ITS 5.6 ITS 6.0 ADMINISTRATIVE CONTROLS I PROCEDURES AND PROGRAMS (Continued) 6.8.5 Technical Specification Bases Control Proyram This program provides a means for processing changes to the Bases of these Technical Specifications.
- a. Changes to the Bases of the Technical Specification shall be made under appropriate administrative controls and reviews.
- b. Licensees may make changes to Bases without prior Nuclear Regulatory Commission approval provided the changes do not require either of the following: - See UrSl 5.5 )
I. . A change in the Technical Specification incorporated in the license or
- 2. A change to the Updated Final Safety Analysis Report of Bases that requires Nuclear Regulatory Commission 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 Updated Final Safety Analysis Report.
- d. Proposed changes that meet the criteria of Specification 6.8.5.b above shall be reviewed and approved by the Nuclear Regulatory Commission prior to implementation. Changes to the Bases implemented without prior Nuclear Regulatory Commission approval shall be provided to the Nuclear Regulatory Commission on a frequency consistent with 10 CFR 50.71(e).
5.6 6.9 REPORTING REQUIREMENTS ROUTINE REPORTS 5.6 6.9.1 In addition to the applicable reporting requirements of Title 10, Code of Federal Regulations. the following reports shall be submitedlto the Retials trator se In accordance with 10 CFR 50.4 UP REPOR summary report of plant and power escalation testing be submitted following (1) eceipt of an operating license, ) amendment to the license ling a planned increase iv in power level (3) installation of FI that has a different design or been manufactured by a
* / different fuel supplier, and (4)/ndiflcations that may have si ficantly altered the nuclear thermal or hydraulic prf e of the plant. . L
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a description of the measured 'alues of the operating conditons jr charcteristics obtained during the test program and a cot4arison of these values with design predictions and specifications. Any corrective actions that *ere required to obtain satisfactor) operation shall also be described. Any additional specific de/ails required in license condition ased on other commitments shall be included in this report. COOK NUCLEAR PLANT-UNIT 2 Page 6-9 AMENDMENT 54,4, S, , 265 . I Page 9 of 16 Attachment 1, Volume 16, Rev. 1, Page 198 of 256
Attachment 1, Volume 16, Rev. 1, Page 199 of 256 ITS 5.6 ITS 6.0 ADMINISTRATIVE CONTROLS STARTUP REPoT montinued) 6.9.1.3 ^tartup reports shall be su tted within (1) 90 days fo g completion of the startup test program, (2)/0 days following resump n or commencement of commercial power operati ,or (3) 9 months following inial criticality, whichever is 1. earliest. If the Startup rt does not cover all three ants (i.e., initial criticality, completion of startup tea program, and resumption or mmencement of commercial power operation), supp entary reports shall be mitted at least every three months until all three ants have been completed. Eby April 30 (for Occupational Radiation Exposure Report) 1. 5.6.1. 6.9,1.4 Annual reports covering the activities of thalunit as described below for the evious 5.6.7 calendar year shall be submitted nor to MArch of each year. Me iti report shall orto Io e ar criticaniil A.3 6.9.1.5 Reports required on an annual badis shall include: 5.6.1 a. A tabulation on an annual basis of the number of station, utility and other personnel (including contractors) receiving annual exposures greater than. I 100 mrem according to work and job functions', e.g., reactor operations and surveillance, in-service inspection, routine maintenance, special maintenance (describe maintenance), waste processing and refueling. Also included is a tabulation of the total person rem exposures for station, utility, and other personnel associated with each work and job function. The dose assignment to various duty functions may be estimates based on. pocket dosimeter, electronic dosimeter, TLD, or film badge measurements. Small exposures totaling less than 20% of the individual total dose need not be accounted for. In the aggregate, at least 80% of the total deep dose received shall be assigned to specific major work functions. 5.6.7 b. The complete results of steam generator tube in-service inspections performed during the report period (reference Specification 4.4.56.b). ig D ntation of anl ch~e-se to the pressurize pner operated relieWI l ' I Xvez (OVe) or safety vibves. 71 = 5 [d. Inf ation regarding any tances when the l-131 c activity limit I exceeded. . I 5.6.1 Note IAsingle submittal may be made for a multiple unit station. The submittal should combine those sections that are common to all units at the station. 5.6.1 a This tabulation supplements the requirements of 20.2206 of 10 CFR Part 20. I COOK NUCLEAR PLANT-UNIT2 Page 6-10 AMENDMENT;X, W4.47, 10, 226 Page 10 of 16 Attachment 1, Volume 16, Rev. 1, Page 199 of 256
Attachment 1, Volume 16, Rev. 1, Page 200 of 256 ITS 5.6 ITS 6.0 ADMINISTRATIVE CONI'ROLS ANNUAL RADIOLOGICAL ENVIRONMENTAL OPERATING REPORT' 5.6.2 6.9.1.6 The Annual Radiological Environmental Operatin R the operation of the unit during the previous calendar year shall be submitted 11of each year. The report shall include summaries. inrerpretadlons, and analysis 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 (1) the ODCM and (2) Sections IV.B.2, V.B.3. and IV.C of Appendix I to 10 CFR Pan 50. DIOR OACTIVE EFFLUENT RELEASE REPO~ 5.6.3 6.9.1.7
- Ike [ua Radioactive Effluent Release Report covering the operation of the unit during the prevluIi12 months of operation shall be submitted within 90 days after Januaxy I of each year.
The report shall Include a surunnay of the quantities of radioactive liquid and gaseous effluents and solid waste released from the unit. The material provided shall be (I) consistent with the objectives outlined in the ODCM and PCP and (2) in conformance with 10 CFR 50.36a and Section IV.B.I of iippendix I to 10 CFR Pan 50. 5.6.2 Note. 3 A single submittal may be made for a mtple unit station. The submittal should combine those sections 5.6.3 Note that are common to al units at th station howeveror units with separate radwaste system. the stminal shall specify the rel o rad oactive materal for eao unit. / I COOK NUCLEAR PLANT-UNff 2 Page 6411 AMENDMET 58,49,438, 45;7,475,210 I Page 11 of 16 Attachment 1, Volume 16, Rev. 1, Page 200 of 256
Attachment 1, Volume 16, Rev. 1, Page 201 of 256 ITS 5.6 ITS O INSERT I 5.6.2 The Annual Radiological Environmental Operating Report shall include the results of analyses of all radiological environmental samples and of all environmental radiation measurements taken during the period pursuant to the locations specified in the table and figures in the ODCM, as well as summarized and tabulated results of these analyses and measurements in the format of the table in the Radiological Assessment Branch Technical Position, Revision 1, November 1979. 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. Insert Page 6-11 Page 12 of 16 Attachment 1, Volume 16, Rev. 1, Page 201 of 256
Attachment 1, Volume 16, Rev. 1, Page 202 of 256 ITS 5.6 ITS 6.0 ADMINISTRATIVE CONTROLS MONTHLY REACTOR OPERATING REPO 5.6.4 6.9.1.8 Routine reports of operatin RchaE TINOEto th R-PORVs Lor e statistics and ohcume vales shall be submitted on Rcgulatoqy Ugfrunission (Attn: DoculnenConul's) a ronthly basislto the Washington, D.C. 20555, with a U.S/ I of all coyto te_ Region I Offce nq latrthn onth telt oliowjng re the calenuar month covered Dy te: 5.6.5 CORE OPERATING LlMITS REPOR 5.6.5.a 6.9.1.9.1 Core operating limits shall be established and documented in the CORE OPERATING LIMITS REPORT before each reload cycle or any remaining part of a reload cycle for the following:
/a. Moderator Temperature Coefficient Limits for Specification 3/4.1.1.A, I Reactor Cone Safety its, SHUTDOV rvNMARGIN l b. WDropTime lj tsforSpecifi ion3/4.l.3.4,
- c. Shutdown Rod Insertion Limits for Specification 314.1.3.5.
RTS Instrurnentation d. Control Rod Insertion Umits for Specification 3/4.13.6, Overprei ssure AT and Overpower A:TAllowable Value e. Axial Flux Difference for Specification 3142.1, paramete rvalues; RCS Pressure. Temperature. and f. Heat Flux Hot Channel Factor for Specification 3/4.22, Flow DNB LImits; and Boron Conc entration. g. Nuclear Enthalpy Rise Hot Channel Factor for Specification 3/4.2.3, and Allowable Power Level for Specification 3/4.2.6. 5.6.5.b 6.9.1.92 The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC in:
- a. WCAP-9272-P-A. 'Westinghouse Reload Safety Evaluation Methodology,' J 1 (Westinghouse Proprietary),
- b. WCAP-8385. 'Power Distribution Control and Load Following Procedures - Topical Report,'
Setep r 7 (Westinghouse Proprietary).
- c. WCAP-10216-P-A Rev on IA "Relaxation
' of Constant Axial Offset Control/FQ Surveillance Technical Specification, "Ifa 9 (Westinghouse Proprietary),
- d. WCAP1 "Phe 1981 Version of Westinghouse Evaluation Mode Using BASH Code,' rbE (Westinghouse Proprietary).
- e. WCAP-12610-P-A. 'VANTAGE+ Fuel Assembly Reference Core Report," Uutv'1991 r (Westinghouse Proprietary).
-0 WCAP-8745-P-A, 'Design Bases for the Thermal Overpower AT and Thermal Overtemperature AT Trip Functions,- (Wesunghouse Proprietary).
COOK NUCLEAR PLANT-UNIT 2 Page 6-12 AMENDMENTS1, 13, 57, 175,190, aii, 261 Page 13 of 16 Attachment 1, Volume 16, Rev. 1, Page 202 of 256
Attachment 1, Volume 16, Rev. 1, Page 203 of 256 ITS 5.6 ITS 6.0 ADMJIrATIVE CONTROLS COR [G5UMITS5 REPORT MwtCef 5.6.5.c 6.9.1.93 The core eperatglimitshalbe~dcrmhmd to thai all applcable lims(e.g., 0*lihermu-mecualca lmt. core 6crmayxdUcx hmk ECCS lhM, mlear lit such uashudown margin, and
=raeswt and scc d811eyu limbs) of eafey analysis are me.
5.6.5.d *A9.1.94 The CORE OPERATING LOAM REPORT. lehdng wy mid-c rviblocs or n lp ertshal peo lmman
- breahqrl6dcycle. to tNe NRC Specia sha be vnbfld go dto of de document crEol U.S. NuclRegulaR w C ht D.C. 20t555). to he Re H and die ReaSlmeo the N= Plt wkbin N perbodpeclfcd for e1s p Ba be eedveskkedled penIa to o' 0 cable reece
- a. bSebmcMoeml S3.3.3.3.
- b. un SpecFcao 4.33.32.
- c. 3 A.7
- d. Spedlc A My InRCS rdfScado.43A.S.
. CS Pr Tru; MlWgd RHR Szqy Valve or RCS V ), Specae 3A.9.3.
- f. Teuainre SPCoc 3.1.lA l
- g. Seal SOxe l ge x kt, Speclcdo 4.7.7.13.
- h. ECCS ,A Sp l 2 w5 .3a3.53 d
L Vixbanof SafetyLiitc.S 6.7.1. I COOK NUCLEAR PLANE-UNrr 2 Pae 6-13 A gD w, M8, *6, 210 'I Page 14 of 16 Attachment 1, Volume 16, Rev. 1, Page 203 of 256
Attachment 1, Volume 16, Rev. 1, Page 204 of 256 ITS 5.6 ITS 3/4 314.3 LIMTING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS INSTRUMENTATION .1 TABLE 3.3-6 (Contined) TABLE NOTATION ACTION 20 - With the number of channels OPERABLE less than required by the Minimum Channels Operable lSee ITS 1 requirement, comply with the ACTION requirements of Specification 3.4.6.1. 3.4.15 J ACTION 21 - With the number of channels OPERABLE less than required by the Minimum Channels Operable See CTS 1 requirement, perform area surveys of the monitored area with portable monitoring instnumenration L 3/4.3.3.1 J at least once per day. ACTION 22 - With the number of channels OPERABLE less than required by the Minimum Channels Operable requirement, comply with the ACTION requirements of Specification 3.9.9. This ACTION Is not SeeJrTS 3
- required during the performance of co_ imnt inegrated leak rate test. i 3.3.6 ACTION 22A- With the number of OPERABLE Channels less than required by the Minimum Channels OPERABLE requirements: See iTS
- 1. either restore the Inoperable Channel(s) to OPERABLE status within 7 days of the event, /
or a l2.
- prepare and submit a Special Report to the Cormmission pursuant to Specification 6.9.2 l 5.6.6 within 14 days following the event outlining the action taken, the cause of the inoperability and the plans and schedule for restoring the systemrto OPERABLE stants.
I Ia. 1, e e I ccnncal Speccicanon -ecoon 3.03a INO ptcoe
. n ma_.
- he ELM_ (( See3.3.3ITS 1J ACTION 22B- With the number. of OPERABLE Channels kss than required by the Minimum Channels OPERABLE requirements.
- 1. either restore the Inoperable annel(s) to OPERABLE status within 7 days of the event, or
- 2. . prepare and submit a Special Report to the Commission pursuant to Specification 6.9.2 See CTS]
within 14 days following the event outlining the action taken, the cause of the inoperabilqi and the plans and schedule for restoring the system to OPERABLE stus.
- 3. In the event of an accident involving radiological releases initiate the preplanned alternate method of monitoring the appropriate parameter(s) within 72 hours.
- 4. Techncal Specification Section 3.0.3 Not Applicable.
COOK NUCLEAR PLANT-UNIT 2 Page 314 3-36 AIENDMENT IM,1M, 151, 265 I Page 15 of 16 Attachment 1, Volume 16, Rev. 1, Page 204 of 256
Attachment 1, Volume 16, Rev. 1, Page 205 of 256 ITS 5.6 ITS REACTMR COOLAN SYSM~ S~MEULLANCE REOUIRDOMT (Continued)
- 9. Prmrvca Inspect10n ns an inspection of the full length of hi tube in tci steam generator performed by eddy current techniques prior to service establish a baseline condition of the tubing. This Inspection shall be performed after the field drostatic test and prior to Initial POWR Seers M
OPtATIONI usi the equipment and techniques expected to be 5.5 ) used during s sequent inservice inspections.
- b. Te stem generator shall be dateruined OPERABLE after copl ting the corresponding actions (plug all tubes exceding the DIlwina limit and all tubes containina throuahwall cracksl
*Irequlr ijbyTabl 4.4-2.
5.6.7 4.4.5.5 R"rts
- a. Following each faservice Inspection of steas, generator tubes, the rnumer of tubes pluged in each stem generator shall be reported to the Comission within 15 days.
- b. The coplete results of the ste enerator tube Inservice Inspection shall be included iMth ar T r ts p rtod In which this inspection was compled. Ths report shall include:
- 1. Numer and extent of tubes inspected.
- 2. Loation and percnt of wull-thfckness penetration for each Indication of an imperfection.
- 3. Identification of tubes plugged.
- c. Results of stem enerator tube is tions which fall Into C-3 andIrequ re prompt not fica ton of sson A10 hall be reported pursuant to Specification 6.9.1 prior to resUotitn of plnt operation. The written follow f this rM tmrnl mI prOwIeua icription our 1nvwu ry ns Gu" WU to de rnse cause of the tube degradation and corrective measures taken to prevent recurrence.
D.C. COOK - UNIT 2 3/4 4-11 Page 16 of 16 Attachment 1, Volume 16, Rev. 1, Page 205 of 256
Attachment 1, Volume 16, Rev. 1, Page 206 of 256 DISCUSSION OF CHANGES ITS 5.6, REPORTING REQUIREMENTS ADMINISTRATIVE CHANGES A.1 In the conversion of the CNP Current Technical Specifications (CTS) to the plant specific Improved Technical Specifications (ITS), certain changes (wording preferences, editorial changes, reformatting, revised numbering, etc.) are made to obtain consistency with NUREG-1431, Rev. 2, "Standard Technical Specifications-Westinghouse Plants" (ISTS). These changes are designated as administrative changes and are acceptable because they do not result in technical changes to the CTS. A.2 CTS 6.9.1 requires, in addition to the requirements of 10 CFR, reports be submitted to the Regional Administrator. ITS 5.6 requires that the reports be submitted in accordance with 10 CFR 50.4. This changes the CTS by removing the explicit requirement to send reports to the Regional Administrator. 10 CFR 50.4 provides distribution requirements for written communications to the NRC. This change is acceptable because the requirements deleted from the Technical Specifications are already required by 10 CFR 50.4. This change is designated as administrative because it does not result in technical changes to the CTS. A.3 CTS 6.9.1.4 regarding annual reports requires the initial report to be submitted prior to March 1 of the year following initial criticality. The ITS does not include such a statement. This changes the CTS by deleting a requirement for report submissions that have already occurred and will not be repeated. This change is acceptable because the one time reporting requirement has already been met and no longer needs to be specified. This change is designated as administrative because it does not result in technical changes to the CTS. A.4 CTS 6.9.1.8 requires the Monthly Reactor Operating Report be submitted to the U.S. Nuclear Regulatory Commission with a copy to the Regional Office. CTS 6.9.1.9.4 requires the CORE OPERATING LIMITS REPORT (COLR) to be provided to the NRC document control desk with copies to the Regional Administrator and Resident Inspector. ITS 5.6.4 requires the Monthly Operating Report to be submitted and ITS 5.6.5.d requires the COLR to be provided to the NRC. This changes the CTS by removing the specifics regarding distribution of the reports to the NRC. 10 CFR 50.4 provides distribution requirements for written communications to the NRC. This change is acceptable because the requirements deleted from the Technical Specifications are already required by 10 CFR 50.4. This change is designated as administrative because it does not result in technical changes to the CTS. A.5 CTS 6.9.1.9.1 requires, in part, that core operating limits be established and documented in the COLR for the rod drop time limits in CTS 3/4.1.3.3. ITS 5.6.5.a does not include a reference to rod drop time limits. This changes CNP Units 1 and 2 Page 1 of 6 Attachment 1, Volume 16, Rev. 1, Page 206 of 256
Attachment 1, Volume 16, Rev. 1, Page 207 of 256 DISCUSSION OF CHANGES ITS 5.6, REPORTING REQUIREMENTS the CTS eliminating the reference to rod drop time limits being core operating limits that are included in the COLR. Rod drop time limits are included in the CTS and the ITS, not the COLR. The information that CTS 3/4.1.3.3 is referring to in the COLR is the definition of what constitutes the full withdrawn position for the purposes of performing the rod drop time Surveillance. This information is not a core operating limit and is therefore not included in the list of individual Specifications that address core operating limits in ITS 5.6.5. This change is acceptable because the information that was moved to the COLR and is referenced in CTS 3/4.1.3.3 (i.e., what constitutes the full withdrawn position) remains in the COLR. This change is designated as administrative because it does not result in technical changes to the CTS. A.6 CTS 6.9.1.9.1 contains a list of the core operating limits established and documented in the COLR and CTS 6.9.1.9.2 contains a list of the locations for the analytical methods used to determine the core operating limits. ITS 5.6.5.a includes additional core operating limits established and documented in the COLR. These are Reactor Core Safety Limits; SHUTDOWN MARGIN; Reactor Trip System Instrumentation Functions 6 and 7 (Overtemperature AT and Overpower AT; respectively) Allowable Value parameter values; RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling Limits; and Boron Concentration. These limits had previously been addressed in other parts of the CTS, but are being moved to the COLR in the ITS, and because of this are listed in ITS 5.6.5.a. ITS 5.6.5.b.6 includes the document describing the analytical methods for the Overtemperature AT and Overpower AT Allowable Value parameter values. This changes the CTS by adding core operating limits established and documented in the COLR (and applicable methodology) because they are being moved there as part of changes to other parts of the CTS. Technical aspects of the changes are addressed in the Discussion of Changes for the respective individual ITS Specifications. This change is acceptable because it administratively documents changes made to other parts of the CTS and the COLR. This change is designated as administrative because it does not result in technical changes to the CTS. A.7 CTS 6.9.2 requires special reports be submitted to the NRC and lists the CTS Specifications that require special reports to be submitted. The ITS does not require these special reports to be prepared and submitted. This changes the CTS by deleting the references to the CTS Specifications requiring special reports. Justification for disposition of each of the special report requirements is addressed by the Discussion of Changes for the respective ITS or CTS Specification. The purpose of CTS 6.9.2 is to identify the Specifications that require special reports to be submitted. This change is acceptable because the special reports are no longer required by the respective Specifications. Justification for disposition of each of the special report requirements is addressed by the Discussion of Changes for the respective ITS or CTS Specification. This change is designated as administrative because it does not result in technical changes to the CTS. CNP Units 1 and 2 Page 2 of 6 Attachment 1, Volume 16, Rev. 1, Page 207 of 256
Attachment 1, Volume 16, Rev. 1, Page 208 of 256 DISCUSSION OF CHANGES ITS 5.6, REPORTING REQUIREMENTS A.8 CTS 4.4.5.5.b requires the complete results of the steam generator tube inservice inspection to be included in the Annual Operating Report. ITS 5.6.7 requires these same results to be submitted on an annual basis (i.e., prior to March 1 for the inspection that was completed in the previous calendar year). This changes the CTS by eliminating the requirement to include the steam generator tube inservice inspection results in the Annual Operating Report. The purpose of CTS 4.4.5.5.b is to ensure the results of the steam generator tube inservice inspection are provided to the NRC. It is not necessary to specify the report that will include the results. This change is acceptable because the steam generator tube inservice inspection results will still be required to be provided to the NRC at the same Frequency as in the CTS. This change is designated as administrative because it does not result in technical changes to the CTS. A.9 CTS 6.9.1.6 and 6.9.1.7 Footnote 3 states that, for these reports, the submittal should combine those sections that are common to all units at the station; however, for units with separate radwaste systems, the submittal shall specify the releases of radioactive material for each unit. ITS 5.6.2 and 5.6.3 does not include the portion of the statement concerning units with separate radwaste systems. This changes the CTS by deleting the reference to units with separate radwaste systems. This change is acceptable because CNP Units 1 and 2 share a radwaste system; they do not have separate radwaste systems. This change is designated as administrative because it does not result in technical changes to the CTS. A.10 CTS 4.4.5.5.c requires a prompt notification to the NRC pursuant to CTS 6.9.1 prior to resumption of plant operation and a followup written report if the results of the steam generator tube inspection fall into the Category C-3. ITS 5.6.7.c requires Category C-3 results to be reported to the NRC in accordance with 10 CFR 50.72 and a Licensee Event Report to be submitted in accordance with 10 CFR 50.73. This changes the CTS by explicitly referencing the applicable Regulations that require the report. The purpose of CTS 4.4.5.5.c is to ensure NRC prompt notification and followup written reporting if an inspection result falls into Category C-3. 10 CFR 50.72 governs prompt phone notifications and 10 CFR 50.73 governs written reports. These changes are acceptable because they are consistent with the current manner in which the CTS 4.4.5.5.c notification and reporting are performed. This change is designated as administrative because it does not result in technical changes to the CTS. MORE RESTRICTIVE CHANGES M.1 The second paragraph of ITS 5.6.2 includes details required to be included in the Annual Radiological Environmental Operating Report. CTS 6.9.1.6 does not contain this level of detail. This changes the CTS by requiring additional detail to be included in the Annual Radiological Environmental Operating Report. CNP Units 1 and 2 Page 3 of 6 Attachment 1, Volume 16, Rev. 1, Page 208 of 256
Attachment 1, Volume 16, Rev. 1, Page 209 of 256 DISCUSSION OF CHANGES ITS 5.6, REPORTING REQUIREMENTS The purpose of the second paragraph of ITS 5.6.2 is to specify details to be included in the Annual Radiological Environmental Operating Report. This change is acceptable because the content requirements are consistent with the objectives outlined in the Offsite Dose Calculation Manual. This change is designated more restrictive because it adds new reporting requirements to the Technical Specifications. RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES LA.1 (Type 3 - Removing Procedural Details for Meeting TS Requirements or Reporting Requirements) CTS 6.9.1.9.2 specifies the revision numbers and dates of the referenced methodologies used for the development of the COLR. ITS 5.6.5.b does not contain this level of detail. This changes the CTS by moving the specific methodology references for revisions and dates to the COLR. The removal of these details, which are related to meeting Technical Specifications requirements, from the Technical Specifications is acceptable because this type of information is not necessary to be included in Technical Specifications to provide adequate protection of public health and safety. The ITS still retains the references for the COLR and only NRC-approved methodologies may be used. The methodologies used to develop the parameters in the COLR have obtained prior approval by the NRC in accordance with Generic Letter 88-16. Also, this change is acceptable because the removed information will be adequately controlled in the COLR under the requirements provided in ITS 5.6.5, "CORE OPERATING LIMITS REPORT". ITS 5.6.5 ensures that the applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems limits, and nuclear limits such as SDM, transient analysis limits, and accident analysis limits) of the safety analyses are met and that only NRC-approved methodologies are used. This change is designated as a less restrictive removal of detail change because information relating to the methodology used to develop cycle-specific parameter limits is being removed from the Technical Specifications. LESS RESTRICTIVE CHANGES L.1 (Category 8- Deletion of Reporting Requirements) CTS 6.9.1.1, CTS 6.9.1.2, and CTS 6.9.1.3 contain requirements for submitting a report of plant startup and power escalation testing following receipt of an operating license; amendments to the license involving planned increase in power level; installation of fuel that has a different design or has been manufactured by a different fuel supplier; and modifications that may have significantly altered the nuclear, thermal, or hydraulic performance of the unit. The ITS does not contain such reporting requirements. This changes the CTS by deleting the requirements of CTS 6.9.1.1, CTS 6.9.1.2, and CTS 6.9.1.3. CNP Units 1 and 2 Page 4 of 6 Attachment 1, Volume 16, Rev. 1, Page 209 of 256
Attachment 1, Volume 16, Rev. 1, Page 210 of 256 DISCUSSION OF CHANGES ITS 5.6, REPORTING REQUIREMENTS The purpose of CTS 6.9.1.1, CTS 6.9.1.2 and CTS 6.9.1.3, is to provide a summary of plant startup and power escalation testing following the four specified conditions as verification that the unit operated as expected. This change is acceptable because the regulations provide adequate reporting requirements. If there were any unit conditions outside the expected parameters during unit startup, they would be reported to the NRC if they met the reporting requirements in the regulations. Otherwise, the reports would document that the unit operated as expected and already approved by the NRC, as required by regulations. This change is designated as less restrictive because reports that would be submitted under the CTS will not be required under the ITS. L.2 (Category 1- Relaxation of LCO Requirements) CTS 6.9.1.4 requires annual reports described in CTS 6.9.1.5, which include the Occupational Radiation Exposure Report, to be submitted prior to March 1 of each year. CTS 6.9.1.6 requires the Annual Radiological Environmental Operating Report to be submitted before May 1 of each year. ITS 5.6.1 requires the Occupational Radiation Exposure Report to be submitted by April 30 of each year. ITS 5.6.2 requires the Annual Radiological Environmental Operating Report to be submitted by May 15 of each year. This changes the CTS by allowing an additional time to submit these reports each year. The purpose of the due date for submitting the Occupational Radiation Exposure Report and Annual Radiological Environmental Operating Report is to ensure that the reports are provided in a reasonable period of time to the NRC for review. This change is acceptable because the reports are still required to be submitted in a reasonable time frame. Given that the reports are still required to be provided to the NRC on or before April 30 or May 15, respectively, and cover the previous calendar year, report completion and submittal is clearly not necessary to assure operation in a safe manner for the interval between March 1 and April 30, and May 1 and May 15, respectively. Additionally, there is no requirement for the NRC to approve the reports. This change is designated as less restrictive because it allows more time to prepare and submit the reports to the NRC. L.3 (Category 8 - Deletion of Reporting Requirements) CTS 6.9.1 .5.c and 6.9.1.8 require annual and monthly reporting of all challenges to the Reactor Coolant System pressurizer operated relief valves (PORVs) or safety valves. ITS 5.6 does not include these reporting requirements. This changes the CTS by deleting the requirement to include documentation of all challenges to the Reactor Coolant System PORVs or safety valves in the annual and monthly reports. The purpose of the annual and monthly reporting requirements is to ensure the NRC receives appropriate routine reports of operating statistics and shutdown experience. This change is acceptable because the regulations provide adequate details of reporting requirements, and the reporting of these challenges does not affect continued plant operation. The change deletes the requirement to include documentation of all challenges to the Reactor Coolant System PORVs or safety valves in the annual and monthly reports. This change is designated as less restrictive because reports that would be submitted under the CTS will not be required under the ITS. CNP Units 1 and 2 Page 5 of 6 Attachment 1, Volume 16, Rev. 1, Page 210 of 256
Attachment 1, Volume 16, Rev. 1, Page 211 of 256 DISCUSSION OF CHANGES ITS 5.6, REPORTING REQUIREMENTS LA (Category 8 - Deletion of Reporting Requirements) CTS 6.9.1 .5.d requires annual reporting of information regarding any instances when the 1-131 specific activity limit for the primary coolant is exceeded. ITS 5.6 does not contain any requirements for such a report. This changes the CTS by not including the requirements for the annual reporting of instances when the Technical Specification 1-131 specific activity limit for the primary coolant is exceeded. The purpose of CTS 6.9.1 .5.d is to specify the requirements for submitting information regarding any instances when the Technical Specification 1-131 specific activity limit for the primary coolant is exceeded in an annual report. This change is acceptable because the regulations provide adequate details of reporting requirements, and the reporting of exceeding the 1-131 limit does not affect continued plant operation. Operations or conditions prohibited by the plant's Technical Specifications are required to be reported in accordance with 10 CFR 50.73. Subsequent reports would be provided if necessary, without requiring a specific annual report. This change is designated as less restrictive because reports that would be submitted under the CTS will not be required under the ITS. CNP Units I and 2 Page 6 of 6 Attachment 1, Volume 16, Rev. 1, Page 211 of 256
Attachment 1, Volume 16, Rev. 1, Page 212 of 256 Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs) Attachment 1, Volume 16, Rev. 1, Page 212 of 256
Attachment I, Volumie 16Rv ae213 of 256 Reporting Requirements'
*50ADMINISTRATIVE CONTROLS; ~* ' 5.6Reporting Requ6Iements.
With 10 CFR 50A.4. A. The foll1owing repo rts shall be submitted In accordance ~ ~ 5.6.1 *OccuOEationial Radiation Exoosure Re6ort, eM1el 7 (0~4') 2 ln submittal may be'made CofrKa multiple unit staflion. The ubitlsod
'combie secftios cmmron to aflunits atthestatlon{(.~'
(~ nd*ther Atabulation' orn an annual basis 6f the number of satbton, utlty receiving monitoring'Was performed, J.. personnel (Including coniractors). for whomi and the associated collective deep' an annual deep dose equivalent > 100 mrems qivlet(rprted Inperson r erem)aacording to w~ork end jobfunctions doe reactor operations and ure1 *Ieri nspaction. routine
*(e.g.. maintenianceae aste processing_%
wmineac
"-.maintenance, specialr and refueling). This tabulation supplements the requiremenisFof.
'Th~ dose-assignments to'various~duty functions ma be
- a. .
10 CFR 20.2206.
'estimated baiied onpce chamber, thermolu~nnescenc dosimeters 6aialn Small exposures'
-(TLD), electronic dosimeter, or film badge rieasuremreints.
dose need not be accounted for. In totaling < 20 percent bf the Individual total of the total ditep dose equivlent received the aggregate, at least 8O percent function's.' The from external sources should be assigned to specific'majorJwork by April 30 of eac-h report covering the previous calendarershl be submitted o ~e3Froow yea submfteP7?ra beireot
.5.62 AnnualRdooca Env rnetlOperating Reeor The submittal s'hould 1j~A single subrn~ittal may be made Ior a multiple unit statio~n combiniesctons comnio6nto allunits at the statoon!,.
covering the operation The Annual Radiological Environmental Operating Report by May 15 of. of the ufiitduring the prevIous calendir year shall be submitted each year. The report shall include summaries, Interpretations, and analyses of' trends of the results of the Radiological Environmental Monitoring Program' for The material provided shall be consistent with the the reporting period. Manual (ODCM) a-nd in. objectives outlined Inthe Offsite Dose Calculationi 10 CFR 50, Appendix 1,Se~tions IV.B.2, IV.13.3. 'and IV.C'.. The Annual Radioilogical Environim"ental Operating Report shall include the results of analyses of all radiological environmental samrples and of all 5.6-l-'
- Rev. 2. 04130101
- WOG T
'Attachment1, Vlume"16, Rev. 1IPg 213 of25
.:AttachmentI 1Vlme.16 Rev Pae214 a o6f.256:
Reporing Requirements 5.6Reporting Requirements-5.6. AnaRaological Environmrental OoemAnnuaRepodi (continued) ([ Al environ~meintal radaiaion meas'urements'takn during the peniod pursuiant to the
*M~e3 4 locations spe6Jifed Inthe table And figures In the ODCM. as well as summarized and tabulated results of these Analyses and measur'ementsfn the format of the
- . .'table in th6 Ridilorogcal Assessment Branch Technical PositioriRevisi'on I.
fNovember 197V' In the 'event that some Individual resuits 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. 1 *
~..,15.6.3 Ra-a'ieEfluentReasReot .
j . 1ringle subimittal may be niade foe a multiple unit station. The submittal shall
* .cominesetioscmmon. to 'all Units at the statiol however, or its with) * . Th'e Radioactive Efflue~nt:Relea'se Report covering the operation of the unilt In the*.*
previous 11b submitted
* -Terportshal c e sumaryof the quantitieso
*.radioactive liquid and gaseous efuents nd solid waste released from the unit.
- The material provided shall be consistent with the objectives outlined In the'
* .. O0CM and Process Control Programf and in conformance With 10 CFR 50.36a andl10CFR Part 50,AppendixI, Section IV.B.1. . *:., ~,
5.6.4 Monthly Operatln~~rs *. .. **
- eorsofoerating itatistics; n shuitdow experish6 shall b subittd o amonthly'basis no -late'rthan the 15th of each monthfolwnth calendar month overed by the'report.'*.. .. olwn h 5.6.5 CORE OP RTNGUIS REPR CI). ~ .
aCore operating limits shall beesitablished prior to each reload cycle, or prior t~. ' I ' I *to any remaining portion of a reload cycle. and shallI be documented In the COLR for the foil wing: WOG STS . ..... 5.6-2 - . . ev.2 04/30101' "Attachmet1,Vlme6 Rev .1, Page 214 of 256.
Attachment 1, Volume 16, Rev. 1, Page 215 of 256 5.6 CMS INSERT 1 6.9.1.7 within 90 days of January 1 of each year Q INSERT 2 6.9.1.9.1 1. SL 2.1.1, "Reactor Core Safety Limits";
- 2. LCO 3.1.1,"SHUTDOWN MARGIN (SDM)";
- 3. LCO 3.1.3, "Moderator Temperature Coefficient (MTC)";
- 4. LCO 3.1.5, "Shutdown Bank Insertion Limits";
- 5. LCO 3.1.6, "Control Bank Insertion Limits";
- 6. LCO 3.2.1, "Heat Flux Hot Channel Factor (Fa(Z))";
- 7. LCO 3.2.2, "Nuclear Enthalpy Rise Hot Channel Factor (FH)";
- 8. LCO 3.2.3, "AXIAL FLUX DIFFERENCE (AFD)";
- 9. LCO 3.3.1, "Reactor Trip System (RTS) Instrumentation," Functions 6 and 7 (Overtemperature AT and Overpower AT, respectively) Allowable Value parameter values;
- 10. LCO 3.4.1, "RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits"; and
- 11. LCO 3.9.1, "Boron Concentration."
Insert Page 5.6-2 Attachment 1, Volume 16, Rev. 1, Page 215 of 256
Afttachmentl1,Volume 16 Rh 1 age 2'16`0 o256 7 Reporting Requirements 56Reportinig Re-quirmet
.5.65 COE PERATING LIMITSRP T (continued)'
- b. The usb to c trietecn e glmt hl e described in the following documents.
jietf.eTopical b br _ott and title r idetfy safSe
*Evalu on Report f a pat spi fic mehdly by.NR Jterndt.
The LRwillcona the corn ete identifica n for each the TS.
- rfecdtopi reports use to prepare tV COLIR Qereport n'um r,
*e, revlsion, te aind an Vpplements).A
(~ ' , c The core operating limits shall be determined such'that all applicable limits (egfuel thermal mechanicallimilts, core thermal hydraulic limits. Emergency Core Cooling Systems (ECCSJ limits,huclear limits s'uch as'.*
§:-SDM, tra'nsient analysis rimits,'and accident analysis limits) of the safety analysis are -met-'
- 9;.~q 9 d.. The COLR, Including any rfiidcycie revisions or supple'nents, .shall be **..-
provided upon Issuance for each reload cycle to the. NRC.;* 56.6 etorC n eS REN MPR TERL
'a 'RCS pressure and temprt Imts for heat up, 6cldlown,w temperature operation crt ity, and hydrostatic testl~ngLj P rming, and etnsas -OVlfwell s hatup and cooldw rates alb esalse n ocume td In the PTLR for th oio :~
[The indivfdual'spe~cifi tns that address RC pres re and temperaturn. limits 'mustbe refrne here.] Ii' b 'The analytical hd used to determine the CS 'pressure and .:~.* temperatureI t hl be thosepviul r ew'ed and approve by te NRC, sp'ec Ily those describedin hol n documents: [Identify aRCsaff approval doue b ae.] eh TRshl eprovided to the N upon issuance for each re or slfuneperiod and for any re sion or supplement thereto 5.6~'3'- -. 04.3016 WOG STS§ * *. *. .. 5.- Rev.2 4100 Atachmen't 1,Vlm 6 ev.1 ae26o 5
Attachment 1, Volume 16, Rev. 1, Page 217 of 256 5.6 CTS (2) INSERT 3 6.9.1.9.2
- 1. WCAP-9272-P-A, "Westinghouse Reload Safety Evaluation Methodology,"
(Westinghouse Proprietary);
- 2. WCAP-8385, "Power Distribution Control and Load Following Procedures - Topical Report," (Westinghouse Proprietary);
- 3. WCAP-10216-P-A, "Relaxation of Constant Axial Offset Control/F 0 Surveillance Technical Specification," (Westinghouse Proprietary);
- 4. WCAP-10266-P-A, "The 1981 Version of Westinghouse Evaluation Mode Using BASH Code," (Westinghouse Proprietary);
- 5. WCAP-12610-P-A, "VANTAGE+ Fuel Assembly Reference Core Report,"
(Westinghouse Proprietary); and
- 6. WCAP-8745-P-A, "Design Bases for the Thermal Overpower AT and Thermal Overtemperature AT Trip Functions," (Westinghouse Proprietary).
Insert Page 5.6-3 Attachment 1, Volume 16, Rev. 1, Page 217 of 256
Attac'hmen 1Voum 16, Rv1 Pe28of256`` cars ~Rep ortling Requ~iremients
.5.6 Reporting Requirements
.6 ~ S ESR N EMYRATURE UMITS.REP conu REVIEWERS E Ei
.The etooo fothcaculato ofteP limits for NRC approval shoul include the flling provisions:,..
I The thodolog shall dlescribe h the neutron flueneicacad - . .7 : '~(ref none new Regulatory Guld hnIssued). ..- '.. .-
.2 eReactor Vessel atrl rvianePoram shall corn wt Apendix H to 1'CR 50 reco eslmateriaq ra to re surveillance spec~men re Iaschedule shall be pro vide aogwt how the specimen examinat sshall be used to update therLR cuvs
- 3. ow Temperature Ov rressure Protection (LTOP)-. stem lift setting limits for the Power Ope eReIeVaves (PORM). de/eloped using NRC bpoed metho oge a eIncluded in te LR.
4 'The adjutdr erneeprture (ART) to each reactor beltilne material shal beci td1acutn o radiation mbrittlement. In accordance4
*with Regualor Gulde 1.'99, Revision 2.
.5 h I ng ART shall be In'corpoat to the calculation of the'pes and t prature limit curves In ccrace with NUREG-0806 Stad d .
Re w Plan' 5.3.2, Pre~sure~-Ternpyfature Limits. .... *..- 6.'6 r..r lipmie'tmeth~odology. . The minimrum temperature uremet oAppenix G to I CFR Part 50 shall be Incorporated Into pressure and tempieraure UI curves.. 8 Llcenseeswho have re ved'twib or more6 easus ud compare for.' each surveillance mnat althe measured Increase In trence temipera'ture (RTNODT) tthprd dinea nRTNO wu eret predicted increase in RTNm Is based on e mean shift In.RTc~ plus th astandard deviation
*... .*.value (2 caA) sp~pci dIn Regulatory Guide I 'g9fels~lon 2. 'If the measure value !exceeds prdicted vialue (increase ,. + c~.the ricensee shoud prvfd upplment to the PTLR to emon~strate howy the resul affect the apoe mehodolobgy.
WOT .. *... ... 6; f~ 2, 04/30/01 "Rev.
-Attachment i, :16,Re.A oume "V1,Page 218 of '256-6.-'.-.'"..
Attachment 1, Volume 16, Rev. 1, Page 219 of 256 Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.0O Post Accident Monitoring Report Acka.. ~2.-A.-Z When a report Is required by Condition B or of LCO 3.32*'Post Accident Monitoring (PAM) 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.
.6.8 o~no uvilneeo t
/ Ay abnormal de adation of the conta nme ~tructure detected during~h test
/required by the re-stressed Concrete Cor anment Tendon Sdrveillane Poram sh~b reported to the NRC wJ n 30 days. The report s~l Include a P
/ ~~~depriptoc f the tendon condition, th ecndition of the conrt sellya
- tendon aihrages), the Inspection ocdures, the tolerane caing, and the cgrcive action taken.} -.
Iq( 6,q~f q 5.6 3 Steam Generator Tube Inspection ReDgA~ (C Z 6 .4¶.(- X7$
- RpEWER'S NOTES -
- 1. R rts required by the censee's current lice sing basis regarding st m nerator tube surveil ce requirements sh be Included here. An appropriate adminis tive controls format ould be used.
- 2. These reports y be required cove rinspection, test, and intenance activities. Th e reports are determied on an individual bas for each unit land their praration and submita re designated In the T hnical Specificatins.. /
WOG STS 5.6-5
- Rev. 2, 04/30/01 Attachment 1, Volume 16, Rev. 1, Page 219 of 256
Attachment 1, Volume 16, Rev. 1, Page 220 of 256 5.6 CTS Q INSERT 4 6.9.1 .5.b, 4.4.5.5
- a. Within 15 days following the completion of each inservice inspection of steam generator tubes, the number of tubes plugged in each steam generator shall be reported to the NRC.
- b. The complete results of the steam generator tube inservice inspection shall be submitted to the NRC prior to March 1 for the inspection that was completed in the previous calendar year. This report 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.
- c. Results of steam generator tube inspections which fall into Category C-3 shall be reported to the NRC in accordance with 10 CFR 50.72. A Licensee Event Report shall be submitted in accordance with 10 CFR 50.73 and shall provide a description of investigations conducted to determine the cause of the tube degradation and corrective measures taken to prevent recurrence.
Insert Page 5.6-5 Attachment 1, Volume 16, Rev. 1, Page 220 of 256
Attachment 1, Volume 16, Rev. 1, Page 221 of 256 JUSTIFICATION FOR DEVIATIONS ITS 5.6, REPORTING REQUIREMENTS
- 1. Grammatical/typographical error corrected.
- 2. The brackets are removed and the proper plant specific information/value is provided.
- 3. ISTS 5.6.3 requires submittal of the Radioactive Effluent Release Report prior to May 1 of each year in accordance with 10 CFR 50.36a. The phrase "in accordance with 10 CFR 50.36a" is duplicative of the requirements in 10 CFR 50.36a, and is therefore not required to be in the Technical Specifications. 10 CFR 50.36a states that the report must be submitted within one year of the previous report. The existing CNP CTS submittal date for this report is not May 1 of each year. Since Technical Specifications cannot supersede the requirements of 10 CFR 50, implementation of this change would require NRC approval of an exemption request in accordance with 10 CFR 50.12. This is considered to be outside the scope of the ITS conversion.
Therefore, the submittal date for this report is revised in ITS 5.6.3 to reflect the CNP CTS (i.e., within 90 days of January 1 of each year).
- 4. ISTS 5.6.6, "Reactor Coolant System (RCS) PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)," is not adopted in the ITS. CTS Figures 3.4-2 and 3.4-3, which provide Reactor Coolant System heatup and cooldown limitations, respectively, were adopted in ITS 3.4.3, "RCS Pressure and Temperature (P/T)
Limits." Subsequent Specifications are renumbered accordingly. In addition, since the PTLR is not included in the ITS, approved TSTF-419, which modifies ISTS 5.6.6, is not incorporated.
- 5. The ISTS Reviewer's Notes have been deleted since they were not intended to be included in the ITS. The requirements for the Steam Generator Tube Inspection Report have been included consistent with these ISTS Reviewer's Notes and the CNP CTS requirements.
- 6. Changes made to reflect those changes made to ITS 3.3.3, "Post Accident Monitoring (PAM) Instrumentation."
CNP Units 1 and 2 Page 1 of 1 Attachment 1, Volume 16, Rev. 1, Page 221 of 256
Attachment 1, Volume 16, Rev. 1, Page 222 of 256 Specific No Significant Hazards Considerations (NSHCs) Attachment 1, Volume 16, Rev. 1, Page 222 of 256
Attachment 1, Volume 16, Rev. 1, Page 223 of 256 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 5.6, REPORTING REQUIREMENTS There are no specific NSHC discussions for this Specification. CNP Units 1 and 2 Page 1 of 1 Attachment 1, Volume 16, Rev. 1, Page 223 of 256
, Volume 16, Rev. 1, Page 224 of 256 ATTACHMENT 7 ITS 5.7, High Radiation Area , Volume 16, Rev. 1, Page 224 of 256 , Volume 16, Rev. 1, Page 225 of 256 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) , Volume 16, Rev. 1, Page 225 of 256
Attachment 1, Volume 16, Rev. 1, Page 226 of 256 ITS 5.7 ITS 6.0 ADMINISTRATIVE CONTROLS 6.11 RADIATION PROTECTION PROGRAM See CTS Procedures for personnel radiation protection shall be prepared consistent with the requirements of 10 CFR Part 20 6.0 and shall be approved, maintained and adhered to for all operations involving personnel radiation exposure. 6.12 HIGH RADIATION AREA 5.7 6.12.1 Pursuant to 10 CFR 20.1601(c), in lieu of the requirements of 10 CFR 20.1601(a) and (b), each high radiation area in which radiation levels from radiation sources external to the body could result in an individual receiving a dose equivalent in excess of 100 mrem but less than or equal to 5.7.1 1000 mrem in I hour at 30 cm from the radiation source or 30 cm from any surface that the radiation penetrates, shall be barricaded and conspicuously posted as a high radiation area and entrance thereto shall be controlled by requiring issuance of a Radiation Work Permit. Any individual or group of individuals permitted to enter such areas shall be provided with or accompanied by one or more of the following:
- a. A radiation monitoring device which continuously indicates the radiation dose rate in the area.
- b. A radiation monitoring device which continuously integrates the radiation dose rate in the area and alarms when a preset Integrated dose is received. Entry into such areas with this monitoring device may be made after the dose rate level in the area has been established and personnel have been made aware of it.
- c. An individual qualified in radiation protection procedures who is equipped with a radiation dose rate monitoring device. This individual shall be responsible for providing positive control over the activities within the area and shall perform periodic radiation surveillance at the frequency specified by the[iWadiationrotectionKtanager in the.
Radiation Work Permit. 5.7.2 6.12.2 The requirements of 6.12.1 shall also apply to each high radiation area in which the radiation level at 30 cm from the radiation source or 30 cm from any surface that the radiation penetrates is greater than 1000 mrem in 1 hour. When possible, locked doors shall be provided to prevent unauthorized entry into such areas, and the kes hall be mainjained under the administrat ivi* control of theJift ?/anager on duty~igor the0l adiationrotectonJ tanager. Doors shalJ I LA.1) remain locked except during periods of access by personnel under an approved RWP which shall - specify the dose rate levels in the immediate work areas. In the event that it is not possible or practicable to provide locked doors due to area size or configuration, the area shall be roped off, conspicuously posted and a flashing light shall be activated as a warning device. 5.7.1 IHealth Physics (Radiation Protection) personnel shall be exempt from the RWP issuance requirement during the performance of their assigned radiation protection duties, provided they comply with approved radiation protection procedures for entry into high radiation areas. COOK NUCLEAR PLANT-UNIT I Page 6-14 AMENDMENT 94,460,454,439, 3;, 346, 279 Page 1 of 2 Attachment 1, Volume 16, Rev. 1, Page 226 of 256
Attachment 1, Volume 16, Rev. 1, Page 227 of 256 ITS 5.7 ITS 6.0 ADMINISTRATIVE CONTROLS
-1 6.11 RADIATION PROTECTION PROGRAM See CTS -
Procedures for personnel radiation protection shall be prepared consistent with the requirements of 10 CFR Part 20 6.0 J and shall be approved, maintained and adhered to for all operations involving personnel radiation exposure. 6.12 HIGjH RAMiATION AREA 6.12.1 l FFrsuant to 10 CFR 20.1601(c). in lieu of the requirements of I0CFR20.1601(a) and (b). each 5.7 - high radiation area in which radiation levels from radiation sources external to the body could result In an individual receiving a dose equivalent in excess of 100 mnrem but less than or equal to 5.7.1 1000 mirem in I hour at 30 cm from the radiation source or 30 cm from any surface that the radiation penetrates, shall be barricaded and conspicuously posted as a high radiation area and entrance thereto shall be controlled by requiring Issuance of a Radiation Work Permit. Any individual or group of individuals permitted to enter such areas shall be provided with or accompanied by one or more of the following:
- a. A radiation monitoring device which continuously indicates the radiation dose rate in the area.
- b. A radiation monitoring device which continuously Integrates the radiation dose rate in the area and alarms when a preset Integrated dose is received. Entry into such areas with this monitoring device may be made after the dose rate level in the area has been established and personnel have been made aware of It.
C. An individual qualified In radiation protection procedures who is equipped with a radiation dose rate monitoring device. This individual shall be responsible for providing positive control over the activities within the area and shall perform periodic radiation surveillance at the frequency specified by thebfwadiation wrotection Manager in the f. y Radiation Work Permit. o 5.7.2 6.12.2 The requirements of 6.12.1 shall also apply to each high radiation area in which the radiation level at 30 cm from the radiation source or 30 cm from any surface that the radiation penetrates is greater than 1000 mrem in 1 hour. When possible. locked doors shall be provided to prevent unauthorized en into such areas. and the keys shall be maintained under the administrative control of the hifanager on duty[~or theIltadiation jrotection $anager. Doors shal LA1) remain locked except during periods of access by personnel under an approved RWP which shall specify the dose rate levels in the immediate work areas. In the event that it is not possible or practicable to provide locked doors due to area size or configuration, the area shall be roped off. conspicuously posted and a flashing light shall be activated as a warning device. 5.7.1 Health Physics (Radiation Protection) personnel shall be exempt from the RWP issuance requirement during the performance of their assigned radiation protection dudes, provided they comply with approved radiation protection procedures for entry into high radiation areas. COOK NUCLEAR PLANT-UNIT2 Page 6.14 . AMENDMENT 80,436,438,4A5,310,22, 261 Page 2 of 2 Attachment 1, Volume 16, Rev. 1, Page 227 of 256
Attachment 1, Volume 16, Rev. 1, Page 228 of 256 DISCUSSION OF CHANGES ITS 5.7, HIGH RADIATION AREA ADMINISTRATIVE CHANGES A.1 In the conversion of the CNP Current Technical Specifications (CTS) to the plant specific Improved Technical Specifications (ITS), certain changes (wording preferences, editorial changes, reformatting, revised numbering, etc.) are made to obtain consistency with NUREG-1431, Rev. 2, "Standard Technical Specifications-Westinghouse Plants" (ISTS). These changes are designated as administrative changes and are acceptable because they do not result in technical changes to the CTS. MORE RESTRICTIVE CHANGES None RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES LA. 1 (Type 3 - Removing Procedural Details for Meeting TS Requirements or Reporting Requirements) CTS 6.12.1.c uses the title "Plant Radiation Protection Manager" and CTS 6.12.2 uses the titles "Shift Manager" and "Plant Radiation Protection Manager." ITS 5.7.1.c uses the generic title "radiation protection manager" and ITS 5.7.2 uses the generic titles "shift manager" and "radiation protection manager." This changes the CTS by moving the specific CNP organizational titles to the UFSAR and replacing them with generic titles. The removal of these details, which are related to meeting Technical Specification requirements, from the Technical Specifications is acceptable because this type of information is not necessary to be included in Technical Specifications to provide adequate protection of public health and safety. The allowance to relocate the specific CNP organizational titles out of the Technical Specifications is consistent with the NRC letter from C. Grimes to the Owners Groups Technical Specification Committee Chairmen, dated November 10, 1994. The various requirements of the radiation protection manager and the shift manager are still retained in the ITS. Also, this change is acceptable because the removed information will be adequately controlled in the UFSAR. Any changes to the UFSAR are made under 10 CFR 50.59 or 10 CFR 50.71(e), which ensures changes are properly evaluated. This change is designated as a less restrictive removal of detail change because information related to meeting Technical Specification requirements are being removed from the Technical Specifications. CNP Units 1 and 2 Page 1 of 2 Attachment 1, Volume 16, Rev. 1, Page 228 of 256
Attachment 1, Volume 16, Rev. 1, Page 229 of 256 DISCUSSION OF CHANGES ITS 5.7, HIGH RADIATION AREA LESS RESTRICTIVE CHANGES None CNP Units I and 2 Page 2 of 2 Attachment 1, Volume 16, Rev. 1, Page 229 of 256
Attachment 1, Volume 16, Rev. 1, Page 230 of 256 Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs) Attachment 1, Volume 16, Rev. 1, Page 230 of 256
Attachment 1,,Volume. 1 Rev IPg 31o 5 High Radiation Area 5.7.
- 5.0 -ADMINISTRATIVE CONTROLS:..I-.
* . ~~.7.-High Radiation Area?
6.a. As provided in'para'graph 20.160'1(c) of 10 CFR Pat20heflloig controlssh6all be applied
-to high radiation areas in place of the controls required b'yparagraph20.1601(a) and (b) of 10 CFR Pakt2O:. :.*., .
7HihRadiation Wokoerise)o Exciiaenttat includes peicato r~~~h proeoueqipenad adiation oesres po.e asahg radiation moitring deviea otnoslydisplays rdatindse. ratesinxthparea or e
.: ..Accs rdain cll~e~I oitor~ dvce tuhpr'hatlbcotnu slinteralledbtherani adiose Wrate ineth ad aarmlwent ea thet devices dspecala 3.Ado rdis at mntoringedeicetate continuously tansm doserraean
-rdation protectionpersonndel sresosibefr iin esne hieiIh ndividual ae In radiation qualified protectoonrcduesane ent Volum reqAuach foran 16, Reqva wil Paer 231 ofr25
Attachment 1, Volume 16, Rev. 1, Page 232 of 256 CTS 5-7 Q INSERT 1 6.12.1 5.7.1 Each high radiation area in which radiation levels from radiation sources external Fotdote
- to the body could result in an individual receiving a dose equivalent in excess of 100 mrem but less than or equal to 1000 mrem in 1 hour at 30 cm from the radiation source or 30 cm from any surface that the radiation penetrates, shall be barricaded and conspicuously posted as a high radiation area and entrance thereto shall be controlled by requiring issuance of a Radiological Work Permit (RWP). Radiation protection personnel shall be exempt from the RWP issuance requirement during the performance of their assigned radiation protection duties, provided they comply with approved radiation protection procedures for entry into high radiation areas. Any individual or group of individuals permitted to enter such areas shall be provided with or accompanied by at least one of the following:
- a. A radiation monitoring device that continuously indicates the radiation dose rate in the area;
- b. A radiation monitoring device that continuously integrates the radiation dose rate in the area and alarms when a preset integrated dose is received. Entry into such areas with this monitoring device may be made after the dose rate level in the area has been established and personnel have been made knowledgeable of it; or
- c. An individual qualified in radiation protection procedures who is equipped with a radiation dose rate monitoring device. This individual shall be responsible for providing positive control over the activities within the area and shall perform periodic radiation surveillance at the frequency specified by a radiation protection manager in the RWP.
6.12.2 5.7.2 In addition to the requirements of Specification 5.7.1 above, for each high radiation area in which the radiation level at 30 cm from the radiation source or 30 cm from any surface that the radiation penetrates is greater than 1000 mrem in 1 hour, locked doors shall be provided, when possible, to prevent unauthorized entry into such areas and the keys shall be maintained under administrative control of the shift manager on duty or a radiation protection manager. Doors shall remain locked except during periods of access by personnel under an approved RWP that shall specify the dose rate levels in the immediate work areas. In the event that it is not possible or practicable to provide locked doors due to area size or configuration, the area shall be roped off, conspicuously posted and a flashing light shall be activated as a warning device. Insert Page 5.7-1 Attachment 1, Volume 16, Rev. 1, Page 232 of 256
Attachmeint 1,Volume' 16, Rev.1, Page 233 of 256 K High Radlation Ara 5.7 57High Radiation Area, Ig Ka 12ion Areas with Dose es a [Ua en imeIers fro th adiation Source or froman Surfa Penetrated by the RadiationI (co ued)
..contrinuouslydispla raiation dose rtes intileareawoi
.responsible for oiing personnel exposure within the'area, or (Ii)B ude th ~rvllance as specified In the RWP oreqvan while In the*e by means of closed circuit televisiono personnel uIiied inradi'ation protection proceduresrsnie for con lig persornnelradiation exposure in the airea dwt
.:the m nslo communicate with Individuals Inthe are oar cov by such surveillance.
- e. Except for dvduals qiualified In radiation protection pro urso person0 cntinuously escorted by such individuals, e intosc areas_
shall made only after dose rates in the area have en determined and ent eronnel are, knowledgeable of them. Thes ntinuously escorted p nnel will receive a pre Jo nfng pnrit yIt such areas. This., se rate determination. knowledge, and prJbefn does not require' dcumnentation prior to Initial entry. 5.72 kr§wt o a'10rmhu t3 bt essthn 00ra/ou ratfMete from te Radlatio Source or Radaton radiEatio6n*~ ry t o such an are shl tie conspicuously posted as a high rdao area and shall be ovdd with'a locked or continuously gu ard door or gate that pentnauthorized entry, and, In addition: 17 All such door a gate keys 'shall be maintained underth administrative ntrol of the shift supervisor, radiation po to manager, or isor her designees and
* . or gteihall remain locked except during' adof persro elo qipment entry or exilt b.Ac ,and actlivities in, each such area'shall b cnrleby means o an Por equivalent tha inlue peicai frdation dose rates In th mediate work area(s) and other erappo eradiation protection, uipment-and measures.:
inivd alsqaified in'radiation protect i procedures may. beexmp froth reuiemet aran RWP or uivalent while perfrIng'r to WOG STS -:5.7 -2. . :Rev; 2,043/1 Attaichmient 1, Volume 16, Rev.1, IPagei 233 of 256
Attacmet1, "Volumre 16 Rev 1,-age 234:of 256-- 5.7... RdighRaditionAre High .*5.7
.2 Hih R tion Areas with Dose Rates Greater W n rem/hour at 30 Centimeter fro aRdaonSource or from any Surfac Penetrated by the Radiation, but less tha0 rds/ourat 1 Meter from the'Radplion Source or from.any'Surface Ptaed by the Radiation (continued surveys In suc areas rv that they are otherwise6 followin~g p
- .jradiation protection ri c rs for entry to, exit from, and work.Isc areas.: .
-d. Each idvual grq entering such an area shiall posses
...A radiao moitrngd~evice tha~t cointinuously I rte[h radiation
*.rates I erandalarms when the devie oearmetpoint Is radwth an appropri~te'alarm setpoin 2 raiatin moitoring device thatcotin usytasmits dose rate and cmltvdose Inomtint rem rcebiver monitored by radiation protectfon er-scnnel rsoible for controlling personnel radiation exposure within theareap ith temen tocmuiaewt and control every individual IntV4area, or 3.. A self-reading doslmeter (e. okt oiainchamber or elctronic dosimeter) and.
-. (I) Be undersurve' as speciied In the.RWP or equivale. 3/4 prcdu equipped wiAth a radiation monitorndel a cponrt inuuo ly displays radiation dose rates In te'area- oIs 7 respo lIbe'for controlling personnel exposure ~vth th area, or (li) B drsrveillanc'e as specified In the RWP equivalent,
- I ~e i thearea, bj means of closed circuit evsoor pesne ualified Inradiation protection ceueresoonsible, forcotrllngpersonnel radijatio expos e nthe area, and with man tcommu lae with and th ti~ol every Individual Inthe area. .. ..
4 In thos6 cases'wtiere optIons (2) an'3,aoeaeIpatclo determined to be InconsIstent wl sLwA sRaoal Achilevable' principle, a radiatlo oitoring device thtcn inuul
'displaces radiation dos rate the erea.
eExcet forIndividtials.qualifle n radiation protection procedures personnel contindiously e d by such Individuals, entry Into ra shall be made one 'after' se rates in the area have been de ean WoG STS . ...- 5 3 Rev. 2, 04310 Atahmet 1, Volume 1,Rv1,Pg234 of 256'
Attachment , Volume 16, Rev. 1, Page 5 256 : ;2.-
..High Radiation Area, 5.7 5.7 High Radiation Area gh ttlo Aras with Dose Rates eater than 1.0 rem/hour at 3Centitr frop e Radiation Source or from aeSurface Penetrated by the Radiati but less n 500 rads/hour at 1 Meter frKthe Radiation Source or from any ace Penetrated by the Radation onUnued):. -; . .. .'
entry person arelknowledgeablefthen. Theseg ntnuoslyescorted
-;personn i receive a pre-job briefing prior to a into such areas. This 9e determination, knowledge, and pre briefing does not require
'dose
*d mentation prior tolnitial entry.
* 'Such individual areas thatare within arger area where'no enlosure /
exists for the purpose of locking a where no enclosure can reasonab e constructed around the Individ area need riot be controlled by a e' door or gate, nor continuousd guarded, but shall be barricaded,. conspicuously posted, an a clearly visible flashing light shall activated the area as a wa device.. WOG STS' NV. 2,04130101 Attachmenl Voldme .16, Rev. Pagie 235 256
Attachment 1, Volume 16, Rev. 1, Page 236 of 256 JUSTIFICATION FOR DEVIATIONS ITS 5.7, HIGH RADIATION AREA
- 1. ISTS 5.7 provides requirements for High Radiation Areas. The brackets are removed and the proper plant specific information/value is provided. ITS 5.5.7 is revised to reflect the CNP current licensing basis and high radiation area controls The change is consistent with the requirements in CTS 6.12.
CNP Units I and 2 Page 1 of 1 Attachment 1, Volume 16, Rev. 1, Page 236 of 256
Attachment 1, Volume 16, Rev. 1, Page 237 of 256 Specific No Significant Hazards Considerations (NSHCs) Attachment 1, Volume 16, Rev. 1, Page 237 of 256
Attachment 1, Volume 16, Rev. 1, Page 238 of 256 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 5.7, HIGH RADIATION AREA There are no specific NSHC discussions for this Specification. CNP Units 1 and 2 Page 1 of 1 Attachment 1, Volume 16, Rev. 1, Page 238 of 256
Attachment 1, Volume 16, Rev. 1, Page 239 of 256 ATTACHMENT 8 Relocated/Deleted Current Technical Specifications (CTS) Attachment 1, Volume 16, Rev. 1, Page 239 of 256
, Volume 16, Rev. 1, Page 240 of 256 CTS 6.0, Administrative Controls , Volume 16, Rev. 1, Page 240 of 256 , Volume 16, Rev. 1, Page 241 of 256 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) , Volume 16, Rev. 1, Page 241 of 256
Attachment 1, Volume 16, Rev. 1, Page 242 of 256 CTS 6.0 6.0 ADMINISTRATIVE CONTROLS 6.3 FACILITY STAFF OUAL!FICATIONS 6.3.1 Each member of the facility staff shall meet or exceed the minimum qualifications of ANSI N18.1-1971 for comparable positions, except for (1) the Plant Radiation Protection Manager, who shall meet or exceed jee ITS 5 2) qualifications of Regulatory Guide 1.8, September 1975, (2) the Shift Technical Advisor, who shall have a land ITS 5.3 J bachelor's degree or equivalent in a scientific or engineering discipline with specific training in plant design, and response and analysis of the plant for transients and accidents and. (3) the Operations Director, who must be qualified as specified in Section 6.2.2.g. 6.4 TRAINING 6A.1 A re ning and replacement training of Training Manager and shall lN18.1-1971 and 10CFRPart for the facility staff shall be ntained under the direction or exceed the requirements and re mmendations of Section 5.5 of -o 6.5 DELETED COOK NUCLEAR PLANT-UNIT 1 Page 64 AMENDMENT49,63,,454, 116, 492,226, 43, 279 Page 1 of 10 Attachment 1, Volume 16, Rev. 1, Page 242 of 256
Attachment 1, Volume 16, Rev. 1, Page 243 of 256 CTS 6.0 6.0 ADMINISTRATIVE CONTROLS
- b. EcP RTAL EVNT ;hall Aleviwdby the PORC, adttrsults Of this review shallI P !bmitjted lo heNS n t~te Vice Prsdent. / p)
§2Ji SAFETY LIMT VlOLATION 6.7.1 The following actions shall be taken in the event a safety limit is violated:
- a. The NRC Operations Center shall be notified by telephone as soon as possible and in all cases within I hour. The Chairman of the NSRB shall be notified within 24 hours.
See UTS
- b. A Safety Limit Violation Report shall be prepared. This report shall be reviewed by the PORC. Chapter 2.0)
This report shall describe (1) applicable circumnstances preceding the violation; (2) effects of the violation upon facility components, systems or structures; and (3) corrective action taken to prevent recurrence.
- c. The Safety liumit Violation Report shall be submitted to the Conmnission, the Chairman of the NSRB and the Senior Vice President - Nuclear Operations within 14 days of the violation.
- d. Operation of the unit shall not be resumed until authorized by the Commission.
COOK NUCLEAR PLANT-UNIT I Page 6-5 AMENDMENT $7,154,89,49A,236, 279 Page 2 of 10 Attachment 1, Volume 16, Rev. 1, Page 243 of 256
Attachment 1, Volume 16, Rev. 1, Page 244 of 256 CTS 6.0 6.0 ADMINISTRATIVE CONTROLS 6.1 -- Q~llPROTE-ClIONPRGA Prcd- r 1peonnel radiation poetn 1 bpraedconsistentwihh/qre nso10CRPt20) an hal/b aprved, maintained andahrgtofrilpeton involvin zne I radiation exposure. 6.12 HIGH RADIATION AREA 6.12.1 Pursuant to 10 CFR 20.1601(c). in lieu of the requirements of 10 CFR 20.1601(a) and (b). each high radiation area in which radiation levels from radiation sources external to the body could result in an individual receiving a dose equivalent in excess of 100 mrem but less than or equal to 1000 mrem in I hour at 30 cm from the radiation source or 30 cm from any surface that the radiation penetrates, shall be barricaded and conspicuously posted as a high radiation area and See 57S entrance thereto shall be controlled by requiring issuance of a Radiation Work Permit . Any individual or group of individuals permitted to enter such areas shall be provided with or accompanied by one or more of the following:
- a. A radiation monitoring device which continuously indicates the radiation dose rate in the area.
- b. A radiation monitoring device which continuously integrates the radiation dose rate in the area and alarms when a preset integrated dose is received. Entry into such areas with this monitoring device may be made after the dose rate level in the area has been established and personnel have been made aware of it.
- c. An individual qualified in radiation protection procedures who is equipped with a radiation dose rate monitoring device. This individual shall be responsible for providing positive control over the activities within the area and shall perform periodic radiation surveillance at the frequency specified by the Plant Radiation Protection Manager in the Radiation Work Permit.
6.12.2 The requirements of 6.12.1 shall also apply to each high radiation area in which the radiation level at 30 cm from the radiation source or 30 cm from any surface that the radiation penetrates is greater than 1000 mrem in I hour. When possible, locked doors shall be provided to prevent unauthorized entry into such areas, and the keys shall be maintained under the administrative control of the Shift Manager on duty and/or the Plant Radiation Protection Manager. Doors shall remain locked except during periods of access by personnel under an approved RWP which shall specify the dose rate levels in the immediate work areas. In the event that it is not possible or practicable to provide locked doors due to area size or configuration, the area shall be roped off, conspicuously posted and a flashing light shall be activated as a warning device. Health Physics (Radiation Protection) personnel shall be exempt from the RWP issuance requirement during the performance of their assigned radiation protection dudes, provided they comply with approved radiation protection procedures for entry into high radiation arems. COOK NUCLEAR PLANT-UNIT I Page 6-14 AMENDMENT 94,15,144,139, Z6, 24, 279 Page 3 of 10 Attachment 1, Volume 16, Rev. 1, Page 244 of 256
Attachment 1, Volume 16, Rev. 1, Page 245 of 256 CTS 6.0 6.0 ADMINISTRATIVE CONTROLS 6.13 PROCESS ^l PRGA aPC) 6.13.1 Changesto ePCP:
- a. hail be documented and records (reviews performed shall be retai d as required by the Quality Assurance Program Description. ppendix C, Section 6.10.2.n. This ocumentation shall contain:
- 1. Sufficient information support the change together wi the appropriate analyses or 2.
evaluations justifying e change(s) and A determination that e change will maintain the ovecI conformance of the solidified waste product to exilg requirements of Federal, State, r other applicable regulations.
-0
- b. Shall become effective afte review and acceptance by the P C and the approval of the Plant Manager.
6.14 OFFSIME DOSE CALCULATION MANUAL (ODCM' 6.14.1 Changes to the ODCM:
- a. Shall be documented and records of reviews performed shall be retained as required by the Quality Assurance Program Description, Appendix C, Section 6.10.2.n. This documentation shall contain:
See ITS 5.5) I. Sufficient information to support the change together with the appropriate analyses or evaluations justifying the change(s) and
- 2. A determination that the change will maintain the level of radioactive effluent control pursuant to 10 CFR 20.1302, 40 CFR Part 190, 10 CFR 50.36a. and Appendix I to 10 CFR Part 50 and not adversely impact the accuracy or reliability of effluent, dose, or setpoint calculations.
- b. Shall become effective after review and acceptance by the PORC and the approval of the Plant Manager.
C. Shall be submitted to the Commission in the form of a complete, legible copy of the entire ODCM as a part of or concurrent with the Annual Radioactive Effluent Release Report for the period of the report in which any change to the ODCM was made. Each change shall be identified by markings in the margin of the affected pages, clearly indicating the area of the page that was changed, and shall indicate the date (e.g.. monthlyear) the change was implemented. COOK NUCLEAR PLANT-UNIT 1 Page 6-15 AMENDMENT 37,454,4$9,324,2U4, 279 Page 4 of 10 Attachment 1, Volume 16, Rev. 1, Page 245 of 256
Attachment 1, Volume 16, Rev. 1, Page 246 of 256 CTS 6.0 P!ROC:IRE LPRROL (PCPs _____Z 1.25 Th RCES COpTMOL (CP) shall entalin current formulas, .Mlph , alyse, texts, and terminations to be to ansure that processng angd pclc ing of so ld radioactive its ed en demonstrated prac., ug of acual u or 5 wet solid vas accomplished in uch a to assure coup ma with 10 G& parts 0, 61, and 71, state regu dons burial ground rats and other te governing the fsposal of solid radi Le waste . t,3 1.29 Deletad. omITZ moSJ CAIctAT1o KANATAL (OM) 1.30 The OMsIT DOSE CALCMIATION K*UAL (ODCH) shall contain the methodology and parameters used in the calculation of offalte dozes resulting from radioactive gaseous and liquid effluents, in the calculation of gJoow and liquid effluent monitoring alarx/trIp stpoints. and In the .{SeeITS 5.5) conduct of the Environmental Radiological Montorlng Program.' The ODCH rhall contain (1) the Radioactive Effluent Controls and adiological Environmental Monitoring ?rograms required by Section 6.8.4 and (2) descriptlons of the information that should be included in the Anmal Radiological Eaviromental Operating and Aimual Radioactive Effluent Release Roports required by SpMecfications 6.9.1.4 and 6.9.1.7. GAS1UM RADVASTK ThEATRMDI Rve 1.31 A CASEOCS RADWUSE UREAMIKT SYSTU is ayq sste designed and Installed to reduce radi active gaseous fflunts by cllecting primary coolant system off-gasso from the primary systam end providing for delay or holdup for the purpose of reducing the total radioactivity prior to release to the enviromment See ITS Chapter I.0) V2TIATION EXMUST nRmATKMr MSvM 1.32 A VDnnlATIOB UEDUSM TRZJA1KUT SYSTX is ay sstem designed and installed to reduce gaseous radloiodine or radioactive material In particulat form in effluento b passing ventilation or vent exhaust gases through charcoal absorbers sdUor W A filters for the purpose of removing loWines or particulatao fro, the gaseous exhaust strea prior to the release to the envirooment. Such k system Ls not considared to have arn effect on noble gs effluents. Engineered Safety Feature (EST) atmospheric claonup systems are not conaIderad to be VXUITUATOW-ZMT = S1AZT SYSt components. 1.31 TUROE or IumLO is the controlled process of dischaqring air or gas from a confinement to maintain temperature pressure, humidity. concentration or other operating condition, In such a menoor that replacemeat air or gas Is required to purify the confinement. 1HM 1.34 VNTINC Is the controlled process of discharng air or gas from a confinement to maintain temperature, pressur htumidity, concentration or other operating condition, In such mnner that replacement air or gas is not provided or required during VUlNTIO. Vent, used in sstem nams, does not imply a VUTINO process. COOK NULEAR PIAW
- UNIT 1 1.6 AK T N0. ., 289*
Page 5 of 10 Attachment 1, Volume 16, Rev. 1, Page 246 of 256
Attachment 1, Volume 16, Rev. 1, Page 247 of 256 CTS 6.0 6.0 ADMINISTRATIVE, CONTROLS 633ACILITY STAFF OUALIFICA71ONiS 6.3.1 Each member of the facility staff shall meet or exceed the minimum qualifications of ANSI N18.1-1971 for See ITS 5.2 comparable positions, except for (1) the Plant Radiation Protection Manager. who shall meet or exceed ITS U 5.3 qualifications of Regulatory Guide 1.8. September 1975. (2) the Shift Technical Advisor, who shall have a i bachelor's degree or equivalent in a scientific or engineering discipline with specific training in plant design, and response and analysis of the plant for transients and accidents and, (3) the Operations Director, who l must be qualified as specified In Section 6.2.2.g. 6.4 TRAINI 6.4.1 Aretrajcng and replacement training pro for the facility staff shall be *ntained under the direction ( ) of thekraining Manager and shall meet r exceed the requirements and rec mmendations of Section 5.5 of A]3%1N18.1l1971 and 10CFRPartS 65 DELETED COOK NUCLEAR PLANT-UNIT 2 Pa 6-4 AMENDMENT 4, *444, *8 4 49;,M,24 261 Page 6 of 10 Attachment 1, Volume 16, Rev. 1, Page 247 of 256
Attachment 1, Volume 16, Rev. 1, Page 248 of 256 CTS 6.0 6.0 ADMINISTRATIVE CONTROLS 6.6 REP0RTABW£MNTACTlON 6.6.1 The fo owing actions shall be taken for
- a. /
10 CFR 50.73. RTABLE EVENTS: The Commission shall be tified and a report submitted usuant to the requirements of -0
- b. Ec OTAL EET hll rviwd yth ORan esls fthsreiw hl
- b. EaX ORTABLE EVENT shall Mreviewcd by the PORCan Yesults of this review shall I to the NSRB and t~*ite Vice President. /
6.7 SAFETY LIMIT VIOLATION 6.7.1 The following actions shall be taken in the event a safety limit is violated:
- a. The NRC Operations Center shall be notified by telephone as soon as possible and in all cases within I hour. The Chairman of the NSRB shall be notified within 24 hours. l See UTS
- b. A Safety Limit Violation Report shall be prepared. This report shall be reviewed by the PORC. Chapter 2.0 The report shall describe (I) applicable circumstances preceding the violation; (2) effects of the 0' violation upon facility components, systems or structures- and (3) corrective action taken to prevent recurrence.
- c. The Safety Limit Violation Report shall be submitted to the Commission, the Chairman of the NSRB and the Senior Vice President - Nuclear Operations within 14 days of the violation.
- d. Operation of the unit shall not be resumed until authorized by the Commission.
COOK NUCLEAR PLANT-UNIT 2 Page 6-5 AMENDMENT*3,I8,+75,78,2i, 261 Page 7 of 10 Attachment 1, Volume 16, Rev. 1, Page 248 of 256
Attachment 1, Volume 16, Rev. 1, Page 249 of 256 CTS 6.0 6.0 ADMINISTRATIVE CONTROLS 6.11 RADIATiO PROTECTION PRO RAM Procedures f personnel radiation protection s I be prepared consistent with the equirements of 10 CFR Part 20 and shall /approved, maintained and adher to for all operations involving PeC nncl radiation exposure.
-0 6.12 HIGH RADIATION AREA 6.12.1 Pursuant to I0CFR20.1601(c), in lieu of the requirements of 10CFR2O.1601(a) and (b). each high radiation area in which radiation levels from radiation sources external to the body could result in an Individual receiving a dose equivalent in excess of 100 mrem but less than or equal to 1000 mrem in I hour at 30 cm from the radiation source or 30 cm from any surface that the radiation penetrates, shall be barricaded and conspicuously posted as a high radiation area and (Se ITS 5.]
entrance thereto shall be controlled by requiring issuance of a Radiation Work Permit. Any individual or group of individuals permitted to enter such areas shall be provided with or accompanied by one or more of the following:
- a. A radiation monitoring device which continuously indicates the radiation dose rate in the area.
- b. A radiation monitoring device which continuously integrates the radiation dose rate in the area and alarms when a preset integrated dose is received. Entry into such areas with this monitoring device may be made after the dose rate level In the area has been established and personnel have been made aware of it.
- c. An individual qualified In radiatioin protection procedures who is equipped with a radiation dose rate monitoring device. This individual shall be responsible for providing positive control over the activities within the area and shall perform periodic radiation surveillance at the frequency specified by the Plant Radiation Protection Manager in the Radiation Work Permit.
6.12.2 The requirements of 6.12.1 shall also apply to each high radiation area in which the radiation level at 30 cm from the radiation source or 30 cm from any surface that the radiation penetrates is greater than 1000 mrem in I hour. When possible, locked doors shall be provided to prevent unauthorized entry into such areas. and the keys shall be maintained under the administrative control of the Shift Manager on duty andlor the Plant Radiation Protection Manager. Doors shall remain locked except during periods of access by personnel under an approved RWP which shall specify the dose rate levels in the Immediate work areas. In the event that it is not possible or practicable to provide locked doors due to area size or configuration, the area shall be roped off. conspicuously posted and a flashing light shall be activated as a warning device. Health Physics (Radiation Protection) personnel shall be exempt from the RWP issuance requirement during the performance of their assigned radiation protection duties provided they comply with approved radiation protection procedures for entry into high radiation areas. COOK NUCLEAR PLANT-UNIT 2 Page 6-14. AMENDMENT 80, 46, I, 425,10,32, 261 Page 8 of 10 Attachment 1, Volume 16, Rev. 1, Page 249 of 256
Attachment 1, Volume 16, Rev. 1, Page 250 of 256 CTS 6.0 6.0 ADMINISTRATIVE CONTROLS 6,!13 PROCESS CO1TOL PROGRAM (PCP) // 6.13.1 Changes t the PCP:
- a. hall be documented and records o reviews performed shall b retaineas required by the Quality Assurance Program Description, ndix C, Section 6.10.2.n. This menttion shall contain:
- 1. Sufficient Information t support the change together wi the appropriate analyses or evaluations justifying change(s) and
- 2. A determination that change will maintain the overal conformance of the solidified waste product to exis ng requirements of Federal, State, other applicable regulations.
- b. Shall become effective after view and acceptance by the PO C and the approval of the Plant I Manager.
6.14 OFFSITE DOSE CALCULATION MANUAL (ODCM) 6.14.1 Changes to the ODCM:
- a. Shall be documented and records of reviews performed shall be retained as required by the Quality Assurance Program Description, Appendix C, Section 6.10.2.n. This documentation shall contain:
I. Sufficient information to support the change together with the appropriate analyses or evaluations justifying the change(s) and See ITS 5.5]
- 2. A determination that the change will maintain the level of radioactive effluent control pursuant to 10 CFR 20.1302, 40 CFR Part 190, 10 CFR S0.36a, and Appendix I to 10 CFR Part 50 and not adversely impact the accuracy or reliability of effluent, dose, or setpoint calculations.
- b. Shall become effective after review and acceptance by the PORC and the approval of the Plant I Manager.
C. Shall be submitted to the Commission In the form of a complete, legible copy of the entire ODCM as a part of or concurrent with the Annual Radioactive Effluent Release Report for the period of the report in which any change to the ODCM was made. Each change shall be identified by markings In the margin of the affected pages, clearly Indicating the area of the page that was changed, and shall indicate the date (e.g., monthlyear) the change was Implemented. COOK NUCLEAR PLANT-UNIT 2 Page 6.15 . AMENDMENT 74, M75,N0,2S, 261 Page 9 of 10 Attachment 1, Volume 16, Rev. 1, Page 250 of 256
Attachment 1, Volume 16, Rev. 1, Page 251 of 256 CTS 6.0 1.25 PHYSICS TESTS shall be thos* testa performed to measure the fundamental ruclear characteristics of the reactor core and related instrumentation and 1) described in Chapter 13.0 of the FSAM. 2) authorizsd under the provisions of 10 CT& 50.59, or 3) otherwise approved by the Comiasion. See ITS Chapter 1.0J i - AVERAGE DIS!?tAtTION MhKOf 1.26 I shall be the average (weighted In proportion to the concentration of each radtonuclida In the reactor coolant at the time of sampling) of the mun of the avsrage bae and gSas energies per disintegration (in NOV) for Isotopes, other then Lodines, with half lives greater then 15 minutes, making up at least 95X of the total ronIodine activity In the coolant. ,BOURCE CHI 1.27 A SOURCZ CHZCX shall be the qualitative assessient of Channel response when the Channel mensor is exposed to a radioactive source. .PROCESY O*L PROCRAK (PCP)// 1.21 The 135 CO5ZL (C I) duall contain current formulas, .A. aplinglenalyses, tests, termtonstn to be to ensure that procoss g and packaging of rolld raotive wastes b d an demonstrated process g of actual or sula d vt colid wastes il be accomplished In such a ay MAto a ossUopl ewith 10 CPR Parts , 61, and 71. State re ula ions, burial ground re rementa, and other r remnts governing th d posal of solid ra tiewas t. COO IUCLAI FIAMT - UNIT 2 1.6 hvWCKrN 30. 4, 175 Page 10 of 10 Attachment 1, Volume 16, Rev. 1, Page 251 of 256
Attachment 1, Volume 16, Rev. 1, Page 252 of 256 DISCUSSION OF CHANGES CTS 6.0, ADMINISTRATIVE CONTROLS ADMINISTRATIVE CHANGES A.1 CTS 6.6.1, Reportable Event Action, including CTS 6.6.1.a, specifies, in the case of a Reportable Event, that the Commission be notified and a report be submitted pursuant to the requirements of 10 CFR 50.73. The requirements of CTS 6.6.1 and 6.6.1.a are not included in the ITS. This changes the CTS by removing the requirements for Reportable Event Action. This change is acceptable because the requirements of CTS 6.6.1 and 6.6.1.a are contained in 10 CFR 50.72 and 10 CFR 50.73. Therefore, there is no need to repeat these requirements in the Technical Specifications. Since the CNP Units 1 and 2 Operating Licenses require compliance with 10 CFR 50, the change is designated as administrative because it does not result in technical changes to the CTS. MORE RESTRICTIVE CHANGES None RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES LA.1 (Type 6 - Removal of LCO, SR, or other TS requirements to the TRM, UFSAR, ODCM, QAPD, or lIP) CTS 6.4 states that a retraining and replacement training program for the facility staff shall be maintained under the direction of the Training Manager and shall meet or exceed the requirements and recommendations of Section 5.5 of ANSI N18.1-1971 and 10 CFR Part 55. ITS Chapter 5.0 does not require such a program. This changes the CTS by moving the requirements for the retraining and replacement training program to the UFSAR. The removal of these details from the Technical Specifications is acceptable because this type of information is not necessary to provide adequate protection of public health and safety. These training provisions are adequately addressed by other proposed ITS Chapter 5.0 provisions and by regulations. ITS 5.3, "Unit Staff Qualifications," provides requirements to ensure adequate, competent staff in accordance with ANSI N 18.1-1971 and Regulatory Guide 1.8,1975. ITS 5.2 details organization requirements. ITS 5.2.2.a, 5.2.2.b, and 10 CFR 50.54 state minimum shift crew requirements. Training and requalification of NRC licensed positions is contained in 10 CFR 50.55. Placement of training requirements in the UFSAR will ensure that training programs are properly maintained in accordance with CNP Unit I and 2 commitments and applicable regulations. Also, this change is acceptable because the removed information will be adequately controlled in the UFSAR. Any changes to the UFSAR are made under 10 CFR 50.59 or 10 CFR 50.71(e), which ensures changes are properly CNP Units 1 and 2 Page 1 of 3 Attachment 1, Volume 16, Rev. 1, Page 252 of 256
Attachment 1, Volume 16, Rev. 1, Page 253 of 256 DISCUSSION OF CHANGES CTS 6.0, ADMINISTRATIVE CONTROLS evaluated. This change is designated as a less restrictive removal of detail change because a requirement is being removed from the Technical Specifications. LA.2 (Type 6 - Removal of LCO, SR, or other TS requirements to the TRM, UFSAR, ODCM, QAPD, or lIP) CTS 6.6.1.b states that each reportable event shall be reviewed by the PORC, and the results of this review shall be submitted to the NSRB and the Site Vice President. The ITS does not include this requirement. This changes the CTS by moving these details of Reportable Event Action to the Quality Assurance Program Description (QAPD). The removal of these requirements from the Technical Specifications is acceptable because this type of information is not necessary to be included in the Technical Specifications to provide adequate protection of public health and safety. Given that these reviews and submittal of results are required following the event without a specified completion time, the proposed relocated requirements are not necessary to assure operation of the facility in a safe manner. As such, the relocated requirements are not required to be in the ITS to provide adequate protection of the public health and safety. Also, this change is acceptable because these types of procedural details will be adequately controlled in the QAPD. Any changes to the QAPD are made under 10 CFR 50.54(a), which ensures changes are properly evaluated. This change is designated as a less restrictive removal of detail change because requirements are being removed from the Technical Specifications. LA.3 (Type 6 - Removal of LCO, SR, or other TS requirements to the TRM, UFSAR, ODCM, QAPD, or lIP) CTS 6.11 provides requirements for the Radiation Protection Program. The ITS does not include these requirements. This changes the CTS by moving the requirements for the Radiation Protection Program to the UFSAR. The removal of these requirements from the Technical Specifications is acceptable because this type of information is not necessary to be included in the Technical Specifications to provide adequate protection of public health and safety. The Radiation Protection Program requires procedures to be prepared for personnel radiation protection consistent with 10 CFR 20. These procedures are for nuclear plant personnel and have no impact on nuclear safety or the health and safety of the public. Requirements to have procedures to implement 10 CFR 20 are contained in 10 CFR 20.1101(b). Periodic review of these procedures is addressed in 10 CFR 20.1101 (c). Since the CNP Units 1 and 2 Operating Licenses require compliance with 10 CFR 20, there is no need to repeat the requirements in the ITS. As such, the relocated details are not required to be in the ITS to provide adequate protection of the public health and safety. Also, this change is acceptable because these details will be adequately controlled in the UFSAR. Any changes to the UFSAR are made under 10 CFR 50.59 or 10 CFR 50.71(e), which ensures changes are properly evaluated. This change is designated as a less restrictive removal of detail change because details for meeting Technical Specification and regulatory requirements are being removed from the Technical Specifications. CNP Units 1 and 2 Page 2 of 3 Attachment 1, Volume 16, Rev. 1, Page 253 of 256
Attachment 1, Volume 16, Rev. 1, Page 254 of 256 DISCUSSION OF CHANGES CTS 6.0, ADMINISTRATIVE CONTROLS LA.4 (Type 6 - Removal of LCO, SR, or other TS requirements to the TRM, UFSAR, ODCM, QAPD, or lIP) CTS Definition 1.28 contains the definition for the Process Control Program (PCP). CTS 6.13.1 describe the process for control of changes to the PCP. The ITS does not include these requirements. This changes the CTS by moving the requirements of the PCP to the UFSAR. The removal of these requirements from the Technical Specifications is acceptable because this type of information is not necessary to be included in the Technical Specifications to provide adequate protection of public health and safety. The PCP implements the requirements of 10 CFR 20, 10 CFR 61, and 10 CFR 71. Compliance with these regulations is required by the CNP Units 1 and 2 Operating Licenses, and procedures are the method to ensure compliance with the program. Regulations provide an adequate level of control for the affected requirements and inclusion of this requirement in the Technical Specifications is not necessary. Also, this change is acceptable because these details will be adequately controlled in the UFSAR. Any changes to the UFSAR are made under 10 CFR 50.59 or 10 CFR 50.71(e), which ensures changes are properly evaluated. This change is designated as a less restrictive removal of requirements because details for meeting Technical Specification and regulatory requirements are being removed from the Technical Specifications. LESS RESTRICTIVE CHANGES None CNP Units 1 and 2 Page 3 of 3 Attachment 1, Volume 16, Rev. 1, Page 254 of 256
Attachment 1, Volume 16, Rev. 1, Page 255 of 256 Specific No Significant Hazards Considerations (NSHCs) Attachment 1, Volume 16, Rev. 1, Page 255 of 256
Attachment 1, Volume 16, Rev. 1, Page 256 of 256 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS CTS 6.0, ADMINISTRATIVE CONTROLS There are no specific NSHC discussions for this Specification. p CNP Units 1and 2 Page 1 of 1 Attachment 1, Volume 16, Rev. 1, Page 256 of 256}}