Text

Improved Technical Specification Conversion License Amendment Request, Volume 14, Revision 1, ITS Section 3.9, Refueling Operations
	ML082270671
Person / Time
Site:	Davis Besse
Issue date:	08/07/2008
From:	; FirstEnergy Nuclear Operating Co
To:	; Office of Nuclear Reactor Regulation
References
	L-08-240, TAC MD6398
	Download: ML082270671 (144)
	v • d • e

I DAVIS-BESSE NUCLEAR POWER STATION UNIT 1 IMPROVED TECHNICAL SPECIFICATION CONVERSION LICENSE AMENDMENT REQUEST (I

VOLUME 14 (Rev. 1)

SECTION 3.9 - REFUELING OPERATIONS 77777777777=,

Attachment 1, Volume 14, Rev. 1, Page i ofi Summary of Changes ITSSection 3.9 Change Description Affected Pages The changes described in the Davis-Besse Pages 23, 26, 28, 30, 32, 33, and 36 response to Question 200711271441 have been made. This change deleted the ITS 3.9.2 requirement that the two required source range monitors be on different sides of the core.

The change described in the Davis-Besse response Page 35 to Question 200712261030 (in Section 3.3) has been made. This changes the term "RPS cabinet" to "pre-amplifier," consistent with the ISTS.

The changes described in the Davis-Besse Pages 2, 39, 41, 42, 44, 45, 47, 48, 49, 51, 58, 61, response to Question 200801161532 (in Section 62, 67, 68, 77, 78, 85, 86, 92, 93,114,116,117, 3.3) have been made. This change adds back into' 118,119, 121,129, 130, 131,132,133, 134, 135, the ITS the Decay Time Specification and deletes 136, 137, 138, 139, 140, 141, and 142 the Containment Penetrations Specification.

Added titles for UFSAR Appendix 3D references in Pages 17, 35, 67, and 92 the Bases (editorial change for consistency with the resolution to a question on a different section).

Page 1 of 1 Attachment 1, Volume 14, Rev. 1, Page i ofi

Attachment 1, Volume 14, Rev. 1, Page 1 of 142 ATTACHMENT 1 VOLUME 14 DAVIS-BESSE IMPROVED TECHNICAL SPECIFICATIONS CONVERSION ITS SECTION 3.9 REFUELING OPERATIONS Revision 1 0

Attachment 1, Volume 14, Rev. 1, Page 1 of 142

Attachment 1, Volume 14, Rev. 1, Page 2 of 142

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

8. Improved Standard Technical Specifications (ISTS) Not Adopted in the Davis-Besse ITS 0

Attachment 1, Volume 14, Rev. 1, Page 2 of 142

Attachment 1, Volume 14, Rev. 1, Page 3 of 142 ATTACHMENT 1 ITS 3.9.1, BORON CONCENTRATION 0

Attachment 1, Volume 14, Rev. 1, Page 3 of 142

Attachment 1, Volume 14, Rev. 1, Page 4 of 142

Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) 0 Attachment 1, Volume 14, Rev. 1, Page 4 of 142

Attachment 1, Volume 14, Rev. 1, Page 5 of 142 ITS 0 ITS 3.9.1 314,L REFUELING OPERATIONS BORON CONCENTRATION LIMITING CONDITION FOR OPERATION A02 LCO 3.9.1 System The 3.9.1 and boron canal- of concentration the refueling all befilled shall' portions/of maintained Reactor Coolant Funtormtheand ufficlent to I ensure aK~o 0.95 or less;7 which i~nludes a 1%6k/k cons erv~ulve a11owanceJl-for uncertlafntltesý ACTJT.ON .A MODE Add proposed Applicability Note ACTION A With the.requirements of the above specification not satisfied, immediately L02 suspend all operations involving CORE ALTATIONS or positive reactivity changes and initiate and continue borationfo E1Z gp* of 7875 at bricl [

lacid solution, or Its /eauivalennt until !ý,, s reduced io * '0."96. eL0 Iprovisions of/specitication 3.0,3 are qot applicablel.1 SURVEILLANCE REQUIREMENT$ 0

%.:1. 1.1 Ihe above con Mlon shnal.l be oetermnea pr or To: I

a. Removing or nbolting~the reactor vessel h Ad, and L04

b. Withdrawal o any safety orregulating ro In excess of 3 feet 0 SR3.9.1.1 from itts' Vessel.

ful y inserted position wlthin-t e, reactor pressure 4.9.1.2 The boron concentration of the reactor pressure vessel and the refueling canal shall be determined [by chemicilanalysis] at least once each 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . I LA02 DAVIS-BESSE, UNIT I 3/4 9-1 Amendment. No. -44, 207 Page 1 of 1 Attachment 1, Volume 14, Rev. 1, Page 5 of 142

Attachment 1, Volume 14, Rev. 1, Page 6 of 142 DISCUSSION OF CHANGES ITS 3.9.1, BORON CONCENTRATION ADMINISTRATIVE CHANGES A01 In the conversion of the Davis-Besse 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-1430, Rev. 3.1, "Standard Technical Specifications-Babcock and Wilcox Plants" (ISTS).

These changes are designated as administrative changes and are acceptable because they do not result in technical changes to the CTS.

A02 CTS 3.9.1 provides requirements on the boron concentration of all filled portions of the Reactor Coolant System (RCS) and the refueling canal. ITS 3.,9.1 provides requirements on the boron concentration of the RCS and the refueling canal. This changes the CTS by deleting the term "all filled portions" when referring to the RCS.

This change is acceptable because the technical requirements have not changed. The term RCS, in the context of this Specification, is referring to the water volume. Furthermore, the ITS Bases states that the boron concentration is the soluble boron concentration "in the coolant" in each of these volumes, which further clarifies that the term "RCS" is referring to the water volume. Thus, use of the term "all filled portions" is redundant. This change is designated as administrative because the technical requirements of the specification have not changed.

A03 CTS 3.9.1 Action 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 None RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES LA01 (Type 5 - Removal of Cycle-Specific ParameterLimits from the Technical Specification to the Core OperatingLimits Report) CTS 3.9.1 states that the boron concentration in MODE 6 shall be sufficient to ensure a keff of 0.95 or less, Davis-Besse Page 1 of 4 Attachment 1, Volume 14, Rev. 1, Page 6 of 142

Attachment 1, Volume 14, Rev. 1, Page 7 of 142 DISCUSSION OF CHANGES ITS 3.9.1, BORON CONCENTRATION which includes a 1% Ak/k conservative allowance for uncertainties. 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 Specification," 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 continued 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 limits 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 5.6.3, "CORE OPERATING LIMITS REPORT." ITS 5.6.3 ensures that the applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling System limits, and nuclear limits such as the 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.

LA02 (Type 3 - Removing ProceduralDetails for Meeting TS Requirements or Reporting Requirements) CTS 4.9.1.2 requires that the boron concentration of the reactor pressure vessel and the refueling canal be determined "by chemical analysis" at least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . ITS SR 3.9.1.1 requires a similar verification, but does not specify that the boron concentration be determined "by chemical analysis." This changes the CTS by moving the details of how the boron concentration is determined 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 limits. In addition, 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 asa less restrictive removal of detail change because procedural details for meeting Technical Specification Requirements are being removed from the Technical Specifications.

0 Davis-Besse Page 2 of 4 Attachment 1, Volume 14, Rev. 1, Page 7 of 142

Attachment 1, Volume 14, Rev. 1, Page 8 of 142 DISCUSSION OF CHANGES ITS 3.9.1, BORON CONCENTRATION LESS RESTRICTIVE CHANGES L01 (Category 2 - 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 that states "Only applicable to the refueling canal when connected to the RCS."

This changes the CTS by eliminating the applicability of the boron concentration limits on the refueling canal when those volumes are not connected 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 is'not connected to the RCS (such as when the reactor vessel head is on the reactor vessel), the boron concentration to this volume cannot affect the SHUTDOWN MARGIN. In addition, prior to connecting the refueling canal to the RCS, a boron concentration verification will be performed (as required by SR 3.0.4) 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.

L02 (Category 4 - Relaxation of Required Action) The CTS 3.9.1 Action specifies the compensatory action for when the boron concentration requirement is not met.

One of the compensatory actions is to suspend CORE ALTERATIONS. Under similar conditions, ITS 3.9.1 does not require suspension of CORE ALTERATIONS. This changes the CTS by deleting the requirement to suspend CORE ALTERATIONS when the boron concentration requirement is not met.

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. Thus, when the limit is not met, the CTS 3.9.1 Action suspends CORE ALTERATIONS to preclude an event that could result in not meeting the SHUTDOWN MARGIN limit. CORE ALTERATION is defined in CTS 1.12, in part, as "the movement of any fuel, sources, or reactivity control components, within the reactor vessel with the vessel head removed and fuel in the vessel."

There are two evolutions encompassed under the term CORE ALTERATION that could affect the SHUTDOWN MARGIN: addition of fuel to the reactor vessel and withdrawal of control rods. However, ITS 3.9.1 Required Action A.1 requires immediate suspension of positive reactivity changes. This would include both the addition of fuel to the reactor vessel and the withdrawal of control rods.

Furthermore, another accident considered in MODE 6 that could affect SHUTDOWN MARGIN is a boron dilution event. A boron dilution accident is initiated by a dilution source which results in the boron concentration dropping below that required to maintain the SHUTDOWN MARGIN. A boron dilution accident is mitigated by stopping the dilution. Suspension of CORE ALTERATIONS has no effect on the mitigation of a boron dilution accident.

Therefore, since the only CORE ALTERATIONS that could affect the SHUTDOWN MARGIN are suspended by ITS 3.9.1 Required Action A.1, deletion of the requirement to suspend CORE ALTERATIONS is acceptable. This Davis-Besse Page 3 of 4 Attachment 1, Volume 14, Rev. 1, Page 8 of 142

Attachment 1, Volume 14, Rev. 1, Page 9 of 142 DISCUSSION OF CHANGES ITS 3.9.1, BORON CONCENTRATION change is designated as less restrictive because less stringent Required Actions are being applied in the ITS than were applied in the CTS.

L03 (Category 4 - Relaxation of Required Action) CTS 3.9.1 Action states that when the boron concentration requirement is not met, initiate and continue boration of

> 12 gpm of 7875 ppm boric acid solution or its equivalent until keff is reduced to

< 0.95. ITS 3.9.1 Required Action A.2 requires initiation of action to restore boron concentration to within limit, but does not include the boric acid concentration or flow rate requirements of the borated water being added. This changes the CTS by eliminating the specific requirements for the boric acid solution concentration and flow rate to be used to restore compliance with the LCO:

The purpose of CTS 3.9.1 Action 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 restore to within the required limit. Specifying the boric acid solution concentration and flow rate 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.

L04 (Category5 - 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 safety or regulating rod in excess of 3 feet from its fully inserted position within the reactor pressure vessel.

ITS 3.9.1 does not contain this Surveillance Requirement. This changes the CTS by deleting this specific 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 safety or regulating 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 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, as required by LCO 3.0.4 and SR 3.0.4, in order to avoid immediately entering into the ITS ACTION (which prohibits withdrawal of control rods when the boron concentration requirement is not met). This change is designated as less restrictive because Surveillances required in the CTS will not be required in the ITS.

0 Davis-Besse Page 4 of 4 Attachment 1, Volume 14, Rev. 1, Page 9 of 142

Attachment 1, Volume 14, Rev. 1, Page 10 of 142 Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs)

Attachment 1, Volume 14, Rev. 1, Page 10 of 142'

Attachment 1, Volume 14, Rev. 1, Page 11 of 142 CTS Boron Concentration 3.9.1 3.9 REFUELING OPERATIONS 3.9.1 Boron Concentration 3.9.1 LCO 3.9.1 Boron concentrations of the:Reactor Coolant SysterTMthe refueling canalo ndcavi be maintained mashall within the lirnitspecified in the 00

,COLR.

APPLICABILITY:: MODE 6:

w -- .....--NOTE--------------------------------------------

DOC 1I01 Only applicable to the :refueling cbaiaR connected to the RCS.

When 0

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME Action A. Boron concentration not A.1 A.1 Su pend CORE Su pend COR E /Immediately TST within limit. A TERATIONS.

-47 A4NDJ A Suspend positive reactivity Immediately additions. ©ST AND A. Initiate action to restore Immediately boron concentration to within limit, SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY 4.9.1.2 SR 3.9.1.1 Verify boron concentration iS within the limit 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months specified in the COLR.

BWOG STS 3.9.1-1 Rev. 3.0, 03/31/04 0

Attachment 1, Volume 14, Rev. 1, Page 11 of 142

Attachment 1, Volume 14, Rev. 1, Page 12 of 142 JUSTIFICATION FOR DEVIATIONS ITS 3.9.1, BORON CONCENTRATION

1. Editorial change made for consistency.

2. The term "refueling cavity" is not used at Davis-Besse. This area is considered part of the refueling canal.

0 Davis-Besse Page 1 of 1 Attachment 1, Volume 14, Rev. 1, Page 12 of 142

Attachment 1, Volume 14, Rev. 1, Page 13 of 142 Improved Standard Technical Specifications (ISTS) Bases Markup and Justification for Deviations (JFDs)

Attachment 1, Volume 14, Rev. 1, Page 13 of 142

Attachment 1, Volume 14, Rev. 1, Page 14 of 142 Boron Concentration B 3.9.1.

B 3.9 REFUELING OPERATIONS B13.9:1 Boron Concentration BASES BACKGROUND The limit on the boron concentrations of the. Reactor Coolant System tJ (RCS)kthe refueling canal[ and tIerefuelir rcavi' during refueling ensures that the reactor remains subcritical during MODE 6. Refueling 0

boron concentration is the soluble boron concentration in the coolant in each of the@ volumes having direct access to the reactor core during 0 refueling.

The soluble boron concentration offsets the core reactivity and is measured by chemical analysis of a representative sample of the coolant in each of the volumes. The refueling boron concentration limit is specified in the COLR. Unit procedures ensure the specified boron concentration in order to maintain an overall core reactivity of kf :- 0.95 during fuel handling, with control rods and fuel assemblies assumed to be in the most adverse configuration (least negative reactivity) allowed by unit procedures. -I UFSAR,.Appendix 3D 1.22 [makeup Mal.- . and. ....

caiojn nd.P.ri...at..n IGDC 26 " CFR 50A1endix requires that two independent reactivity control systems of different design principles be provided 0 (Ref. 1). One of these systems must be capable of holding the reactor core subcritical under cold conditions. ThelChemicat-Ad ition System 0 serves as 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 unboltedt removed to form the ref eng cavity. Th efueling canal refu then flooded with borated water from the borated water storage tank into the open reactor vessel by gravity feeding or by the use of Decay Heat Removal (DHR) System pumpM.

The pumping action of the DHR System in the RCS, and the natural circulation due to thermal driving heads in the reactor vessel Irefueli i, mix the added concentrated boric acid with the water in 0

the refueling canal. The DHR System is in operation during refueling (see LCO 3.9.4, "DHR and Coolant Circulation - High Water Level," and LCO 3.9.5, "DHR and Coolant Circulation - Low Water Level") to provide forced circulation in the RCS and assist in maintaining the boron concentrations in the RCS he refueling cana an above the COLR limit.

e re ing cavityl 0 a.nd BWOG STS B 3.9.1-1 Rev. 3.0, 03/31/04 0

Attachment 1, Volume 14, Rev. 1, Page 14 of 142

Attachment 1, Volume 14, Rev. 1, Page 15 of 142 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 accldentanalysis 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 includes an uncertainty allowance.

The required boron concentration and the unit refueling procedures that demonstrate the correct fuel loading plan (including full core mapping) ensure the keff of the core will remain !50.95 during the refueling operation. Hence, at least a 5% Aktk margin of safety is established during refueling.

During refueling, the water volume in the spent fuel pool, the transfer canal, the refueling canal, Ithe refue g cavity,and the reactor vessel form a single mass. As. a result, the soluble boron concentration is relatively 0 the same in each of these volumes.

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 and* the RCSthe refueling cana(,and the re 0ling cay while in MODE 6.

The boron concentration limit specified in the COLR ensures a core kff of 0

< 0.95 is maintained during fuel handling operations.

Violation.of the LCO could lead toan 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 kef- 0.95. Above MODE 6, LCO 3.1.1, "SHUTDOWN MARGIN (SOM)," and LCO 3.1.2, "Reactivity Balance," ensure that an adequate amount of negative reactivity is available to shut down the reactor and to maintain it subcritical-The Applicability is modified by a Note. The Note states that the limits on this boron concentration are only applicable to the refueling canal[a e s retueli cay when se volumeu[aLelconnected to the RCS. When the refueling canal and the re elin cavi ai isolated from the RCS, no potential path for boron dilution exists.

BWOG STS B 3.91-2 Rev. 3.0, 03/31104 Attachment 1, Volume 14, Rev. 1, Page 15 of 142

Attachment 1, Volume 14, Rev. 1, Page 16 of 142 Boron Concentration B 3.9.1-BASES ACTIONS Al n A(TS Continuation of CORE ALT"ATIONS or positive reactivity additions (including actionsto reduce boron concentration) is contingent upon maintaining the unitin compliance with the LCO. Ifthe boronto concentration of any coolant volume in the RCS the refueling canaLEI herefuWnqcavit is less than its.limit, all operations involving F TSTF

[ALTERNS orlpositive reactivity additions must be suspended immediately.

Suspension of ICORE ALTERATIONS andl positive reactivity additions -471 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 fluctuations),

but when combined with all other operations affecting core reactivity (e.g.,

intentional boration) result in overall net negative reactivity addition, are not precluded by this action.

In addition to immediately suspending ICORE ALTERATIONS and J positive reactivity additions, action to restore the concentration must be initiated immediately.

In determining the required combination of boration flow rate and concentration, there is no unique Design Basis Event that must be satisfied. The only requirement is to restorethe 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 actiorn ha been initiated, must be continued until the boron concentration is restored. The restoration time depends on the amount of boron that must be injected to reach the required concentration.

BWOG STS B 3.9.1-3 Rev. 3.0, 03/31/04 Attachment 1, Volume 14, Rev. 1, Page 16 of 142

Attachment 1, Volume 14, Rev. 1, Page 17 of. 142 Boron Concentration B 3.9.1 BASES SURVEILLANCE SR 3.9.1.1 REQUIREMENTS This SR ensures the coolant boron ýconcentrationrin the RC'8 and connected portions of the refueling canal land the ref ini cav is.within the COLR limits. The boron concentration of the coolant in each required volume is determined periodically by .chemical analysis. Prior to re-connecting portions of the refueling canalIor the'ref elin- cavi to the; 0 RCS, this SR must bermet per SR 3.0.4. Ifany dilution activity has -

occurred while the cavy o canai -disconnected from.the RCS, this.

SR ensures the correct boron concentration prior to communication with the.RCS.

The SR 3.9.1.1 A u Frequency of once every 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months i th oa reason'able amount.of time to verify the. boron concentration of representative..

samples. The Frequency is based on operating experience, which has shown 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months to be adequate.

0 REFERENCES 1., 110CF&W-A2pg:ngi:x ýA, G D C UFSAR, Appendix 313.1.22, Critenort 26 - ReacthAty Control tern Redundancy and Capabill jility 0

BWOG STS B 3.9.1-4 Rev. 3.0, 03/31/04 Attachment 1, Volume 14, Rev. 1, Page 17 of 142

Attachment 1, Volume 14, Rev. 1, Page 18 of 142 JUSTIFICATION FOR DEVIATIONS ITS 3.9.1 BASES, BORON CONCENTRATION

1. 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.

2. Changes made to be consistent with the Specification.

3. Changes made to be consistent with changes made to the Specification.

4. Editorial change for clarity.

0 Davis-Besse Page 1 of 1 Attachment 1, Volume 14, Rev. 1, Page 18 of 142.

Attachment 1, Volume 14, Rev. 1, Page 19 of 142 Specific No Significant Hazards Considerations (NSHCs)

Attachment 1, Volume 14, Rev. 1, Page 19 of 142

Attachment 1, Volume 14, Rev. 1, Page 20 of 142 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.9.1, BORON CONCENTRATION There are no specific NSHC discussions for this Specification.

Davis-Besse Page 1 of 1 Attachment 1, Volume 14, Rev. 1, Page 20 of 142

Attachment 1, Volume 14, Rev. 1, Page 21 of 142 pATTACHMENT 2 ITS 3.9.2, NUCLEAR INSTRUMENTATION

(

S.

Attachment 1, Volume 14, Rev. 1, Page 21 of 142

, Volume 14, Rev. 1, Page 22 of 142 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) , Volume 14, Rev. 1, Page 22 of 142

Attachment 1, Volume 14, Rev. 1, Page 23 of 142 ITS 3.9.2 ITS REFUELING OPERATIONS INSTRUMNATO I I ITI G ,CN D TIO N FOR OPERA.T ION ';:. . ..

LCO 3.9.2 3.9'2 Two source range neutron flux monito one from each si the reator core, shallbe OPERABLE.

APPLICABILITY: MODE 6:

ACTION:

ACTION A a. With only one of ,the required OPERABLE source rarge~neutron flux monitors,

-4 L01 I. Immediately suspend CORE ALTERATIONS and

2. Immnediatelysuspend operations that Would cause introduction of coolaritinto the RCS with boron concentration`less than the RCS boron conceniration requirement of LCO 3.9.1.

ACTION A, b. With noOPiERABLE source range.neutron flux monitor, ACTION B

1. :Perform ON a",and A0S

2. Immediately initiate aition to restoreone source.range nefitron flux monitor to OPERABLE status, and 0 3. Once per 12 hobrs ver'fythat:the RCS boron concentration mees the requirementof Ire-LCO actor3.9.1 usinghiTal pressurev ves iand analysisitb determine the~refu eling eanai.I theSb~rhnconentration of the A03 4.9.2 As a minimurm, two, source range neutron flux monitors, one from each side of the reactor core, shall be defninitratid OPERABLE by performance of:

b. elted SR 3.9.2.1 c. A CHANNEL CHECK at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , and.

SR 3.9.2.2 d.rA CHANNEL CALIBRATION nrior. to entry into MOD not erformed within theA04 I 18tISmonths. Neutron detectors are excluded from CHANNEL CALIBRATION.

DAVIS-BESSE, UNIT I 3/4 9-2 Amendment No. 172, 269 0 Page 1of 1 Attachment 1, Volume 14, Rev. 1, Page 23 of 142

Attachment 1, Volume 14, Rev. 1, Page 24 of 142 DISCUSSION OF CHANGES ITS 3.9.2, NUCLEAR INSTRUMENTATION ADMINISTRATIVE CHANGES A01 In the conversion of the Davis-Besse 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-1430, Rev. 3.1, "Standard Technical Specifications-Babcock and Wilcox Plants" (ISTS).

These changes are designated as administrative changes and are acceptable because they do not result in technical changes to the CTS.

A02 CTS 3.9.2 Action b.1 states that when there is no OPERABLE source range neutron flux monitor to perform Action a. ITS 3.9.2 does not contain this specific requirement. this changes the CTS by deleting the specific statement to "perform Action a."

The purpose of CTS 3.9.2 Action b.1 is to clarify that when there is no OPERABLE source range neutron flux monitor, Action a, which provides the actions when there is only one OPERABLE source range neutron flux monitor, must be performed. This statement is not needed in ITS because ITS 3.9.2 ACTION A, applies when there is one inoperable source range neutron flux monitor. Thus, whenever two required source range neutron flux monitors are inoperable, both ITS 3.9.2 ACTION B, which provides the actions for two inoperable source range neutron flux monitors, and ITS 3.9.2 ACTION A must be 0designated entered. Therefore, there is no need for the specific statement. This change is the CTS.

as administrative because it does not result in technical changes to A03 CTS 3.9.2 Action b.3 states that, when both source range neutron flux monitors are inoperable, to verify that the RCS Boron meets the requirement of LCO 3.9.1, using chemical analysis to determine the boron concentration of the reactor pressure vessel and the refueling canal once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . Under similar conditions, ITS 3.9.2 Required Action B.2 requires performance of SR 3.9.1.1 once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . This changes the CTS by replacing the prescriptive requirement for verification of boron concentration with a more general requirement.

This change is acceptable because the CTS requirements have not changed.

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.

A04 CTS 4.9.2.d requires performance of a CHANNEL CALIBRATION on the source range neutron flux monitors "prior to entry into MODE 6 if not performed within the last" 18 months. ITS 3.9.2.2 only requires performance of the CHANNEL CALIBRATION every 18 months. This changes the CTS by deleting the statement "prior to entry into MODE 6 if not performed within the last."

This change is acceptable because the CTS requirement has not changed.

CTS 4.0.4 states that "entry into an OPERATIONAL MODE or other specified applicability shall not be made unless the Surveillance Requirement(s)

associated with the Limiting Condition for Operation have been performed within Davis-Besse Page 1 of 3 Attachment 1, Volume 14, Rev. 1, Page 24 of 142

Attachment 1, Volume 14, Rev. 1, Page 25 of 142 DISCUSSION OF CHANGES ITS 3.9.2, NUCLEAR INSTRUMENTATION the stated surveillance Interval or otherwise specified." This requirement has been maintained in ITS 3.0.4. Therefore, there is no need to restate CTS 4.0.4 (ITS SR 3.0.4). This change is designated as administrative because it does not result in a technical change to the CTS.

MORE RESTRICTIVE CHANGES None RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES

,None LESS RESTRICTIVE CHANGES L01 (Category 4 - Relaxation of Required Action) CTS 3.9.2 Action a specifies the compensatory action for when a source range neutron flux monitor is inoperable.

One of the compensatory actions (CTS 3.9.2 Action a.1) is to suspend CORE ALTERATIONS. Under similar conditions, ITS 3.9.2 Required Action A.1 requires suspension of positive reactivity additions, except the introduction of coolant into the RCS, instead of suspension of CORE ALTERATIONS. This changes the CTS by changing the requirement to suspend CORE ALTERATIONS to only require suspension of positive reactivity additions, not covered by CTS 3.9.2 Action a.2, when a source range neutron flux monitor is inoperable.

The purpose of source range neutron flux monitors is to monitor core reactivity during refueling operations and provide a signal to the operators if an unexpected reactivity change occurs. Thus, when a source range neutron flux monitor is inoperable, CORE ALTERATIONS are suspended to preclude an unmonitored reactivity change. CORE ALTERATIONS is defined in CTS 1.12, in part, as "the movement of any fuel, sources, or reactivity control components, within the reactor vessel with the vessel head removed and fuel in the vessel." CORE ALTERATIONS only occur when the reactor vessel head is removed - it only applies to MODE 6. There are two evolutions encompassed under the term CORE ALTERATION that could affect the reactivity of the core: addition of fuel to the reactor vessel and withdrawal of control rods. However, ITS 3.,9.2 Required Action A.1 requires immediate suspension of positive reactivity changes, except the introduction of coolant into the RCS. This would include both the addition of fuel to the reactor vessel and the withdrawal of control rods. In addition, movement of fuel or control rods that does not add positive reactivity (e.g.,

removal of a fuel assembly from the core) is not required to be suspended since this evolution does not increase core reactivity, thus it is not a safety concern Davis-Besse Page 2 of 3 Attachment 1, Volume 14, Rev. 1, Page 25 of 142

Attachment 1, Volume 14, Rev. 1, Page 26 of 142 DISCUSSION OF CHANGES ITS 3.9.2, NUCLEAR INSTRUMENTATION (i.e., it cannot result in an unexpected criticality event). Therefore, since the CORE ALTERATIONS of concern are only those that could affect positive reactivity in the core, and these are suspended by ITS 3.9.2 Required Action A.1, changing the requirement from suspending to "CORE ALTERATIONS" to suspending "positive reactivity additions, except the introduction of coolant into the RCS" is acceptable. This change is designated as less restrictive because less stringent Required Actions are being applied in the ITS than were applied in the CTS.

L02 (Category 1 - Relaxation of LCO Requirement) CTS LCO 3.9.2 requires that two source range neutron flux monitors, one from each side of the reactor core, be OPERABLE. ITS LCO 3.9.2 requires two source range neutron flux monitors to be OPERABLE. This changes the CTS by eliminating the requirement that the neutron monitors are on each side of the reactor core.

The purpose of the source range neutron flux monitor is to monitor core reactivity during refueling operations and provide a signal to the operators if an unexpected reactivity change occurs. This change is acceptable because the source range monitors have no safety function and are not assumed to function during any UFSAR design basis accident or transient. However, the source range neutron channels provide on scale monitoring of neutron flux levels during refueling conditions. This on scale monitoring function can be performed by any of the two required monitors, even if they are on the same side of the core. Therefore, the wording "one from each side of the reactor core" is not necessary. This change is less restrictive because the requirement that the two source range neutron flux monitors are on each side of the reactor core is not retained.

0 Davis-Besse Page 3 of 3 Attachment 1, Volume 14, Rev. 1, Page 26 of 142

Attachment 1, Volume 14, Rev. 1, Page 27 of 142 UImproved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs)

S Attachment 1, Volume 14, Rev. 1, Page 27 of 142

Attachment 1, Volume 14, Rev. 1, Page 28 of 142 CTS Nuclear. Instrumentation 319.2 3.9 REFUELING OPERATIONS 3.9.2 Nuclear Instrumentation 3.9.2 LCO 3.9,2 Two source range neutron flux monitors shall be.OPERABLE.

APPLICABILITY: MODE 6.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME Action a, Action b A. One Irequiredl source range neutron flux At.1 S us end C RE ALT, RATI N Immediately

,except the introduction of o2(TSTF)

.monitor inoperable, AND [positive reactivity additions*

coolant into the RCS 0

A,2 Suspend operations that Immediately would cause introduction of coolant into the RCS with boron concentration less than required to meet the boron concentration of LCO3.9.,1.

Action b B. Tw6orequirodl source range neutron flux B.1 Initiate action to restore one source range neutron flux Immediately 0

mornitors inoperable, monitor to OPERABLE status.

.AND B.2 Perform SR 3.9.1.1. Once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY 4.9.2.c SR 3.9.2.1 Perform CHANNEL CHECK 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months BWOG STS 3:9.2-1 Rev> 3.0, 03/31/04 Attachment 1, Volume 14, Rev. 1, Page 28 of 142

Attachment 1, Volume 14, Rev. 1, Page 29 of 142 CTS Nuclear Instrumentation 3.92 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY 4.9.2.d SR 3.9.2.2 NOTE--- -

Neutron detectors are. excluded from CHANNEL CALIBRATION.

Perform CHANNEL CALIBRATION. 1181 months 0

BV\OG STS 3.9.2-2 Rev. 3.0, 03/31/04 Attachment 1, Volume 14, Rev. 1, Page 29 of 142

Attachment 1, Volume 14, Rev. 1, Page 30 of 142 JUSTIFICATION FOR DEVIATIONS ITS 3.9.2, NUCLEAR INSTRUMENTATION

1. Not used.

2. The brackets are removed and the proper plant specific information/value is provided.

3. TSTF-471T changed ISTS 3.9.2 Required Action A.1 from requiring suspension of "CORE ALTERATIONS" to requiring suspension of "positive reactivity additions."

However, positive reactivity additions could encompass adding coolant into the Reactor Coolant System (RCS). ISTS 3.9.2 Required Action A.2 allows introduction of coolant into the RCS provided the boron concentration of the added coolant meets the LCO 3.9.1 requirement. Thus, the TSTF essentially precluded all operations of coolant introduction into the RCS unless the boron concentration of the added coolant is greater than or equal to the concentration in the RCS. This was not the intent of the TSTF. Therefore, the amplifying information, "except the introduction of coolant into the RCS" has been included in ITS 3.9.2 Required Action A.1 to allow ISTS 3.9.2 Required Action A.2 control this operation.

Davis-Besse Page 1 of I Attachment 1, Volume 14, Rev. 1, Page 30 of 142

Attachment 1, Volume 14, Rev. 1, Page 31 of 142 Improved Standard Technical Specifications (ISTS) Bases Markup and Justification for Deviations (JFDs)

Attachment 1, Volume 14, Rev. 1, Page 31 of 142

Attachment 1, Volume 14, Rev. 1, Page 32 of 142 Nuclear Instrumentation B3.9.2:

B3.9 REFUELING OPERATIONS B'3.9.2 Nuclear Instrumentation I Reactor Protection System (RPS) and Post Accident Monitoring (PAM) Instrumentation I BACKGROUND- The source range neutron flux monitorsare used during refueling operations to monitor the core reactivity condition. The~installed source range neutron flux monitors are part of the[Nuclear In mentation ISystepi (NIS. These detectors are located external to the reactor vessel and detect neutrons leaking from the core. The use of portable detectors is permitted, provided the LCO.requirements are met. INSERT 1 The installedsource range. neutron flux monitors, areBF3 det ctors oeperating in the roportional region o the.gas filled detector haracteristic c The detectors monitor thez neutron flux in counts per second. The instrument range covers-x*decades of neutron flux 1 E+6 cps)f 1[51% instrur)nt accuracy. The detectors also provide continuous visual indication in the. control room n e a arto alert operators to a possible dilution accident. The [Ns designed in accordance with the criteria presented in Reference .. +Ifused, portable detectors should be functional!y equivalent to the installed M source range monitors.

APPLICABLE Two OPERABLE source range neutron'flux monitors are required to SAFETY

provide a signal to'alert the operator to unexpected changes in core ANALYSES reactivity, such as-by a boron dilution accident or an imprern loaded Ifuel asemblIV The safetv .analvsis of the luncorsfrolledl boron dilution have no safetyfunction are not assumed to accident is described in Reference 2. The analysis of thein accident or transient analysis. However, the boron dilution accident shows thatthe normally available SDI ii would not source range neutron channels provide on be lost;,and there is sufficient time for the operator toltake cor rective.

scale monitoring of neutron flux levels duringI action.

refueling conditions. Therefore, they are . I retained in Technical Specificationsj- The source

,being range neutron ,fluxmonitorsl satisfy COdteiion I P-110 CFR 59-36(e)2) i).

f 0

to) to be LCO This LCO requires two source range neutron flux monitorsOPERABLE ensure that redundant monitoring capability is available to detect changes in core reactivity.

0 APPLICABILITY In MODE 6,the source range neutron flux monitors must be OPERABLE to determine changes in core reactivity. There is no other direct means available to check core reactivity levels. In MODES 2, 3, 4, and 5, these 0 Dsame installedosource range detectors and circuitry are also required to be OPERABLE by LCO 3.3,9, "Source Range Neutron Flux,"

In MODES 1, 2, and 3, these same installed PAM source range detectors and circuitry are also required to be OPERABLE by LCO 3.3.17, "Post Accident Monitoringt (PAM) Instrumentation."7 BVVOG STS B 319.2-I Rev. 3.0, 03/31/04 Attachment 1, Volume 14, Rev. 1, Page 32 of 142

Attachment 1, Volume 14, Rev. 1, Page 33 of 142 B 3.9.2 O INSERT 1 two high sensitivity proportional counters (BF3 chambers).

O INSERT 2 The installed PAM monitors are two safety grade electrically and physically independent fission chamber strings. The channel 1 PAM detector (N15874A) is located near the corresponding channel 1 RPS detector (NI-2) and the channel 2 PAM detector (N15875A) is located adjacent to the corresponding channel 2 RPS detector (NI-1). The detectors monitor the neutron flux in counts per second. The PAM instrument range covers six decades of neutron flux (1 E-1 cps to 1 E+5 cps). The detectors also provide continuous visual indication in the control room and an audible indication to alert operators.

O INSERT4 To be OPERABLE, each monitor must provide continuous visual indication in the control room, and one monitor must provide audible indication in the containment and the control room.

0 Insert Page B 3.9.2-1 Attachment 1, Volume 14, Rev. 1, Page 33 of 142

Attachment 1, Volume 14, Rev. 1, Page 34 of 142 Nuclear Instrumentation B 3.9.2 BASES ACTIONS A.1 and.A.2 With only one trequired~source range neutron flux monitor OPERABLE,-

redundancy has been lost,. Since these instruments'are the only direct mean's of monitoring core,.reactivityconditions, COREAL RATIONS TSTF and introduction of coolant.into the RCS with boron concentration less positive reactivity -471

'than'required to meet the minimnum boron concentration of LCO 3.9.1 additions must be':suspended immediately. Suspending positive reactivityadditions 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 from-sources that have a boron concentration greater thanW 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. Perforrmance.of Required Action A.1 shall not preclude completion of movement of a component to asafe position.

B,1 VVith" norrequired]source range neutron flux monitor OPERABLE; action il to rest6re a monitor'to OPERABLE statuslshall'be =initiated immediately.

Once initiated, action shall becontinued until a source range neutron flux monitor is restored. to OPERABLE status, B.2 Yith no required*]source range neutron flux monitor OPERABLE, there is (I no direct means of detecting changes in core, reactivity. However, since 1CORE.ALT WIONS~andjpsitive. reactivity additions are not to be made, the core reactivity condition is stabilized until the -source range.

neutron flux monitorsare 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 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is sufficient to obtain and analyze:a reactor coolant sample for boron concentration and ensures that unplanned changesin boron concentration wouldbe identified, The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Frequency is reasonable, considering the low probability of a change in core reactiVitY during this time period.

BWOG'STS B.3.9.2-2 Rev. 30, 03/31/04 Attachment 1, Volume 14, Rev. 1, Page 34 of 142

Attachment 1, Volume 14, Rev. 1, Page 35 of 142 Nuclear Instrumentation B 3:9.2, BASES SURVEILLANCE' SR ,3.9.2.1 REQUIREMENTS SR 3.9.2.1 is the performance of a CHANNEL CHECK, which is a comparison of the parameter indicated on one channel to a similar parameteron, other channels. It is based on the assumption that the two indication chaninels should be' consistent with core conditions. Changes in fuel loading and core geometry caniresultijnsignificant differernces between source range channels, but eachchannel Should be consistent with its-local conditions.

The Frequency of 12j hours is consistent withjthe CHANNEL CHECK Frequency specified similarlyfor the same instruments in LCO 3.3.9.

SR 3.9.2.2 SR'3,9.2.2 is the perf6rmance of a CHANNEL CALIBRATION every 118M months. 'This SR ist modified by: a Note stating that neutron detectors are excluded fromtheCHANNEL CALIBRATION. 'The CHANNEL chne S - CALIBRATION for the source rangeFnu ea isa complete check and re-adjustmnentof the channels,'fromthe pre-arplifier input to the indicator The 18 month r uency is, based on the.need to perform this Surveillance durig the conditions that'applydurk a plant outage..

Operating experience has shown these components usually pass the, Surveillance when performed at the V18i))month Frequency.

REFERENCES 1. I10CFR50, Alndix A GDC 13; GDC 26 GD andGDC 228..

2 FSAR, Section~LV 15.4an Apedi 4 and for the PAM source range channels is a complete check of the instrument channel UFSAR. Appendices 3D.1.9, Criterion 13 - Instrumentation and Control; 3D.1.16, Criterion 20- Protection System Functions; 3D. 1.17, Criterion 21 - Protection System Reliability and Testability; 3D.1.18, Criterion 22 - Protection System Independence; 3D.1.19, Criterion 23 - Protection System Failure Modes; 30.1.20, Criterion 24 -

Separation of Protection and Control Systems; and 3D.1.25, Criterion 29 - Protection Against Anticipated Operational Occurrences BVWOGSTS B 3192-3 Rev. 3.0, 03/31/041 Attachment 1, Volume 14, Rev. 1, Page 35 of 142

Attachment 1, Volume 14, Rev. 1, Page 36 of 142 JUSTIFICATION FOR DEVIATIONS ITS 3.9.2 BASES, NUCLEAR INSTRUMENTATION

1. 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.

2. Not used.

3. The brackets have been removed and the proper plant specific information/value has been provided.

4. Typographical error corrected.

Davis-Besse Page 1 of 1 Attachment 1, Volume 14, Rev. 1, Page 36 of 142

Attachment 1, Volume 14, Rev. 1, Page 37 of 142 Specific No Significant Hazards Considerations (NSHCs)

Attachment 1, Volume 14, Rev..1, Page 37-of 142

Attachment 1, Volume 14, Rev. 1, Page 38 of 142 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.9.2, NUCLEAR INSTRUMENTATION There are no specific NSHC discussions for this Specification.

Davis-Besse Page 1 of 1 Attachment 1, Volume 14, Rev. 1, Page 38 of 142

Attachment 1, Volume 14, Rev. 1, Page 39 of 142

ATTACHMENT 3 ITS 3.9.3, DECAY TIME 0

Attachment 1, Volume 14, Rev. 1, Page 39 of 142

, Volume 14, Rev. 1, Page 40 of 142 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) , Volume 14, Rev. 1, Page 40 of 142

Attachment 1, Volume 14, Rev. 1, Page 41 of 142 ITS 3.9.3 ITS I FUELING OPERATIONS DECAY TIME LIMITING CONDITION FOR OPERATION LCO 3.9.3 3.9.3 The reactor shall be subcritical for at least 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

APPLICABILITY: During movement of irradiated fuel in the reactor pressure vessel.

ACTION:

ACTION A Vi1th the reactor subcritical for less than 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , suspend all operations involving movaent of irradiated fuel in the reactor pressure vessel. of Specification

'eprovisjns 3.0.3 are/not applicable._-

SURVEILLAKCE REGUIREMENTS SR 3.9.3.1 4.9.3 The reactor shall be determined to have been subcrititcal for at least 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months by verification of the date and time of :subcritlcality, prior to movement of irradiated fuel in the reactor pressure vessel.

W~IS,.4ESSE. UNIT -I 3/4 9-3.

Page 1 of I Attachment 1, Volume 14, Rev. 1, Page 41 of 142

Attachment 1, Volume 14, Rev. 1, Page 42 of 142 DISCUSSION OF CHANGES ITS 3.9.3, DECAY TIME ADMINISTRATIVE CHANGES A01 In the conversion of the Davis-Besse 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-1430, Rev. 3.1, "Standard Technical Specifications-Babcock and Wilcox Plants" (ISTS).

These changes are designated as administrative changes and are acceptable because they do not result in technical changes to the CTS.

A02 CTS 3.9.3 Action contains the statement, "The provisions of Specification 3.0.3 are not applicable." ITS 3.9.3 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, which is essentially the only time movement of irradiated fuel assemblies within the reactor pressure vessel would be applicable. 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 RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES None LESS RESTRICTIVE CHANGES None Davis-Besse Page 1 of 1 Attachment 1, Volume 14, Rev. 1, Page 42 of 142

Attachment 1, Volume 14, Rev. 1, Page 43 of 142 WImproved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs) 0 Attachment 1, Volume 14, Rev. 1, Page 43 of 142

Attachment 1, Volume 14, Rev. 1, Page 44 of 142 0 Decay Time 3.9.3

. CTS 3.9 REFUELING OPERATIONS 3.9.3 Decay Time 3.9.3 LCO 3.9.3 The reactor shall be subcritical for > 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

APPLICABILITY: During movement of irradiated fuel assemblies within the reactor pressure vessel.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME Action A. Reactor not subcritical A.1 Suspend movement of Immediately for > 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . irradiated fuel assemblies within the reactor pressure vessel.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY 4.9.3 SR 3.9.3.1 Verify reactor subcritical for > 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . Prior to movement of irradiated fuel assemblies within the reactor pressure vessel 0

Davis-Besse 3.9.3-1 Attachment 1, Volume 14, Rev. 1, Page 44 of 142

Attachment 1, Volume 14, Rev. 1, Page 45 of 142 JUSTIFICATION FOR DEVIATIONS ITS 3.9.3, DECAY TIME

1. This Specification has been added to ensure the reactor is shutdown at least 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months prior to moving irradiated fuel assemblies. Even though this Specification is' allowed to be relocated to utility control in accordance with the allowances of NUREG-1430, Davis-Besse has decided to maintain this requirement in the Technical Specifications based on the NRC's desires to maintain control of this time period, as documented in RAIs 200801161530 and 200801161532.

0 Davis-Besse Page 1 of 1 Attachment 1, Volume 14, Rev. 1, Page 45 of 142

Attachment 1, Volume 14, Rev. 1, Page 46 of 142 Improved Standard Technical Specifications (ISTS) Bases Markup and Justification for Deviations (JFDs)

, Attachment 1, Volume 14, Rev. 1, Page 46 of 142

Attachment 1, Volume 14, Rev. 1, Page 47 of 142 0 Decay Time B 3.9.3 B 3.9 REFUELING OPERATIONS B 3.9.3 Decay Time BASES BACKGROUND The movement of irradiated fuel assemblies within the reactor pressure vessel requires that the reactor be subcritical for > 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . This ensures sufficient time has elapsed to allow the radioactive decay of the short lived fission products.

APPLICABLE Prior to movement of irradiated fuel assemblies within the reactor SAFETY vessel, the reactor must be subcritical for > 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . This time period is ANALYSES an initial assumption of the fuel handling accident in containment (Ref. 1) postulated by Regulatory Guide 1.25 (Ref. 2). The minimum time period of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months ensure sufficient time has elapsed to allow the radioactive decay of the short lived fission products, which helps ensure that the offsite doses during a fuel handling accident will be within the 10 CFR 100 limits, as provided by the guidance of Reference 3.

Decay Time satisfies Criterion 2 of 10 CFR 50.36(d)(2)(ii).

LCO The reactor is required to be subcritical for > 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months to ensure that the radiological consequences of a postulated fuel handling accident inside containment are within acceptable limits as provided by 10 CFR 100.

APPLICABILITY LCO 3.9.3 is applicable when moving irradiated fuel assemblies within the reactor pressure vessel. The LCO ensures that the possibility of a fuel handling accident in containment that is beyond the assumptions of the safety analysis cannot occur.

ACTIONS A.1 With the reactor not subcritical for > 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , all operations involving movement of irradiated fuel assemblies within the reactor pressure vessel shall be suspended immediately to ensure that a fuel handling accident in containment cannot occur.

The suspension of fuel movement shall not preclude completion of movement of a component to a safe position.

SURVEILLANCE SR 3.9.3.1 REQUIREMENTS Verification that the reactor has been subcritical for > 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months ensures that the design basis decay time assumption for the postulated fuel handling accident analysis in containment is met.

0 Davis-Besse B 3.9.3-1 Attachment 1, Volume 14, Rev. 1, Page 47 of 142

Attachment 1, Volume 14, Rev. 1, Page 48 of 142 0 Decay Time B 3.9.3 BASES SURVEILLANCE REQUIREMENTS (continued)

The Frequency of prior to movement of irradiated fuel assemblies within the reactor pressure vessel ensures that the 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months limit is met prior to commencement of irradiated fuel movement within the reactor pressure vessel.

REFERENCES 1. UFSAR, Section 15.4.7.3.

2. Regulatory Guide 1.25, March 23, 1972.

3. 10 CFR 100.10.

Davis-Besse B 3.9.3-2 Attachment 1, Volume 14, Rev. 1, Page 48 of 142

Attachment 1, Volume 14, Rev. 1, Page 49 of 142 JUSTIFICATION FOR DEVIATIONS ITS 3.9.3 BASES, DECAY TIME

1. This Specification has been added to ensure the reactor is shutdown at least 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months prior to moving irradiated fuel assemblies. Even though this Specification is allowed to be relocated to utility control in accordance with the allowances of NUREG-1430, Davis-Besse has decided to maintain this requirement in the Technical Specifications based on the NRC's desires to maintain control of this time period, as documented in RAIs 200801161530 and 200801161532.

0 0

Davis-Besse Page 1 of 1 Attachment 1, Volume 14, Rev. 1, Page 49 of 142

Attachment 1, Volume 14, Rev. 1, Page 50 of 142 Specific No Significant Hazards Considerations (NSHCs) 0 0

Attachment 1, Volume 14, Rev. 1, Page 50 of 142

Attachment 1, Volume 14, Rev. 1, Page 51 of 142 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.9.3, DECAY TIME There are no specific NSHC discussions for this Specification.

Davis-Besse Page 1 of 1 Attachment 1, Volume 14, Rev. 1, Page 51 of 142

Attachment 1, Volume 14, Rev. 1, Page 52 of 142 ATTACHMENT 4 ITS 3.9.4, DHR AND COOLANT CIRCULATION - HIGH WATER LEVEL Attachment 1, Volume 14, Rev. 1, Page 52 of 142 *

, Volume 14, Rev. 1, Page 53 of 142 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) , Volume 14, Rev. 1, Page 53 of 142

Attachment 1, Volume 14, Rev. 1, Page 54 of 142

A01 ITS 3.9.4 ITS REFUELING OPERATIONS 3/4.9.8 DECAY HEAT REMOVAL AND COOLANT CIRCULATION ALL WATER LEVELS LIMITING CONDITION FOR OPERATION 3.9.4 3.9.8.1 At least one decay heat removal loop shall be n " aO APPLICABILITY: MODE 6 when the water level above the top of the ite

...... ue asmhA-*fsatdwithin tereactor pressure vesli"0

>23 feet. fag

~ACTION: r except as *

a. With less than one decay heat removal loop in operation, ACTION A provided in b below, suspend all operations involving ran increase in e re~act~or decay heat load or/,areduction in boron concentration of the

.eactor Coolant System./ ose alT ontainment penetrations providing jd i r e c t a c c e s s f ost h n ena t tm hoes p oh e tr eiddto eti s heý - 0 1within. 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . Add proposed Required Action A.3 'e

-LCO3.9.4 *b.. The decay heat removal loop, may be remov~e from ýoperation for up to o~ne M03 Note hour per' 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months period'during the performance of.CORE ALTERATIONS in the Nvicinity of the reactor pressure vessel (hot) legs. M04 1c. The pr,94!is ions of Specif.*tion 3.0.3 are/ot: appli.cabl*. _*

SURVEILLANCE REQUIREMENTS_

4.9.8.1 Surveillance at least once per,12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months 'shall verify at least one SIR3.9.4.1 decay heat removal loop to be.in operation and circulating reactor coolant through the reactor core:

a. At a flow rate, of > 2800 gpm whenever a reductio /in Reactor

oo an ys em oron oncen ration is being made

b. At a flow rate sU that the core outlet tempe ature is maintained

S 140°F,,provide no reduction. in Reactor Coo ant System boron concentration being made.*

Water of a lower boron concentration than the existing RCS M05 Required concentration may be added to the.RCS, wi owrate o Action A.1 Ireactor coo an *u the RCS less than 00Ogpm, provi ed that boo 2he tcnenra-tio 8f the water to be added is" equal to Or greater than the boron concentration corresponding to the more restrictive reactivity condition specified in Specification 3.9.1.

AVIS-BESSE, UNIT 1 3/4 9-8. AmendmenE No. 8A, 188 Page 1 of 1 Attachment 1, Volume 14, Rev. 1, Page 54 of 142

Attachment 1, Volume 14, Rev. 1, Page 55 of 142 DISCUSSION OF CHANGES ITS 3.9.4, DHR AND COOLANT CIRCULATION - HIGH WATER LEVEL ADMINISTRATIVE CHANGES A01 In the conversion of the Davis-Besse 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-1430, Rev. 3.1, "Standard Technical Specifications-Babcock and Wilcox Plants" (ISTS).

These changes are designated as administrative changes and are acceptable because they do not result in technical changes to the CTS.

A02 CTS 3.9.8.1 Action a states, in part, that with less than one DHR loop in operation, suspend all operations involving an increase in the reactor decay heat load of the Reactor Coolant System. Under similar conditions, ITS 3.9.4 Required Action A.2 states to 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.

A03 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. CTS 3.9.8.1 and ITS 3.9.4 are only applicable in MODE 6. 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 M01 CTS 3.9.8.1 requires that at least one decay heat removal loop be in operation.

ITS 3.9.4 requires that one DHR loop shall be OPERABLE and in operation.

This changes the CTS by requiring the DHR loop to also be OPERABLE, instead of just in operation.

The purpose of CTS 3.9.8.1 is to ensure adequate decay heat removal and coolant circulation are available in MODE 6. However, the CTS LCO could be interpreted as allowing a DHR loop to be placed in operation that was not OPERABLE. The ITS eliminates this possible misinterpretation. This change is acceptable because the DHR loop must be OPERABLE (i.e., capable of performing its decay heat removal and coolant circulation function) instead of just Davis-Besse Page 1 of 4 Attachment 1, Volume 14, Rev. 1, Page 55 of 142

Attachment 1, Volume 14, Rev. 1, Page 56 of 142 DISCUSSION OF CHANGES ITS 3.9.4, DHR AND COOLANT CIRCULATION - HIGH WATER LEVEL being in operation. This change is designated as more restrictive because the ITS contains more specific requirements on a component.

M02 CTS 3.9.8.1 requires one DHR loop to be in operation in MODE 6 when the water level above the top of the irradiated fuel assemblies seated within the reactor pressure vessel is > 23 feet. ITS 3.9.4 requires one DHR loop to be OPERABLE and in operation when water level is > 23 feet above the top of the reactor vessel flange. This changes the CTS by changing the point at which either one or two DHR loops are required to be OPERABLE and one in operation. The change requiring the DHR loop to be OPERABLE is discussed in DOC M01.

The purpose of CTS 3.9.8.1 is to ensure adequate DHR is available and in operation for heat removal and coolant circulation. CTS 3.9.8.1 and CTS 3.9.8.2 provide the requirements when water level is > 23 feet and < 23 feet above the top of the irradiated fuel assemblies seated within the reactor pressure vessel, respectively. When water level is > 23 feet above the top of the irradiated fuel assemblies seated within the reactor pressure vessel, only one DHR loop is required to be in operation (and essentially OPERABLE). When water level is

< 23 feet above the top of the irradiated fuel assemblies seated within the reactor pressure vessel, two DHR loops are required to be OPERABLE, and one must be in operation. In ITS 3.9.4 and ITS 3.9.5, the equivalent ITS requirements, the water level reference point is the top of the reactor vessel flange, not the top of the irradiated fuel assemblies seated within the reactor pressure vessel.

Changing this reference point effectively requires a larger complement of DHR loops to be OPERABLE between 23 feet above the top of the irradiated fuel assemblies seated within the reactor pressure vessel and 23 feet above the top of the reactor vessel flange. Therefore, this change is acceptable because more loops will be required to be OPERABLE under certain water level conditions to ensure the decay heat can be removed and the coolant circulated. This change is designated more restrictive because more DHR loops are required OPERABLE in the ITS under certain water level conditions than were required in the CTS.

M03 The CTS 3.9.8.1 Actions do not include an action to immediately initiate action to satisfy the DHR loop requirements in the event the DHR loop requirements are not met. ITS 3.9.4 Required Action A.3 requires that action be immediately initiated to satisfy the DHR loop requirements. This changes the CTS by requiring that action be taken immediately to satisfy the DHR 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 DHR 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.

Davis-Besse Page 2 of 4 Attachment 1, Volume 14, Rev. 1, Page 56 of 142

Attachment 1, Volume 14, Rev. 1, Page 57 of 142 DISCUSSION OF CHANGES ITS 3.9.4, DHR AND COOLANT CIRCULATION - HIGH WATER LEVEL M04 CTS 3.9.8.1 Action b states that the DHR loop may be removed from operation for up to 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months per 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 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 DHR loop may be removed from operation for < 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months per 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months period, provided no operations are permitted that would cause introduction of coolant into the Reactor Coolant System 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 to the CTS. First, the allowance to remove DHR 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 of coolant into the RCS 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 DHR 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. This change is designated as more restrictive because it imposes a new condition to be met when an DHR loop is not in operation.

M05 CTS 4.9.8.1 verifies that the DHR loop is in operation and circulating reactor coolant and provides two flow rate requirements. CTS 4.9.8.1.a requires

> 2800 gpm when a reduction in boron concentration is in progress and CTS 4.9.8.1.b requires a flow rate sufficient to maintain core outlet temperature

< 140OF when a reduction in boron concentration is not in progress. The 2800 gpm flow requirement is also used in CTS 3.9.8.1 footnote *. ITS SR 3.9.4.1 requires the flow rate to be > 2800 gpm under all conditions. This changes the CTS by requiring a higher flow rate when a reduction in boron concentration is not in progress.

The purpose of CTS 4.9.8.1 is to ensure adequate DHR flow necessary to provide sufficient decay heat removal capability and to prevent thermal and boron stratification in the core. This change is acceptable because a higher DHR flow will be required under certain conditions to ensure the above purpose is met.

This change is designated as more restrictive because a higher DHR flow is required under certain conditions in the ITS than in the CTS.

RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES None Davis-Besse Page 3 of 4 Attachment 1, Volume 14, Rev. 1, Page 57 of 142

Attachment 1, Volume 14, Rev. 1, Page 58 of 142 DISCUSSION OF CHANGES ITS 3.9.4, DHR AND COOLANT CIRCULATION - HIGH WATER LEVEL

LESS RESTRICTIVE CHANGES L01 (Category 4- Relaxation of Required Action) CTS 3.9.8.1 Action a states, in part, that with less than one DHR loop in operation, close all containment penetrations providing direct access from the containment atmosphere to the outside atmosphere within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . ITS 3.9.4 Required Actions A.4, A.5, and A.6 state that with the DHR loop requirements not met, within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months 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 and Exhaust System. This changes the CTS Actions by allowing penetrations capable of being closed by an OPERABLE Containment Purge and Exhaust System to remain open when the DHR 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 DHR 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 an accident occurring during the repair period. This change is designated as less restrictive because less stringent Required Actions are being applied in the ITS than were applied in the CTS.

0 Davis-Besse Page-4 of 4 Attachment 1, Volume 14, Rev. 1, Page 58 of 142

Attachment 1, Volume 14, Rev. 1, Page 59 of 142 Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs)

Attachment 1, Volume 14, Rev. 1, Page 59 of 142

Attachment 1, Volume 14, Rev. 1, Page 60 of 142 CTS DHR and Coolant Circulation - High Water Level 3.9.4 3.9 REFUELING OPERATIONS 3.9.4 Decay Heat Removal (DHR) and Coolant Circulation- High'Water Level 3.9.8.1 LCO 3;9.4 One DHR loop shall be OPERABLE and in operation.

....................................... .....M Action b The required DHR loop may be-removed from operation for<* 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . per 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months period, provided no operations are permitted that would cause.

introduction of coolant into the.Reactor Coolant Systen-ltWitbron cs 0 concentration less than that required to meet the minimum required boron concentration of LCO 3.9. 0 APPLICABILITY: MODE 6 with the water level _>23 ft above the top of reactor vessel flange.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME Action a, A. DHR loop requirements A.1 Suspend operations that Immediately Footnote

not met. would cause introduction of coolant into the RCS with boron concentration less than required to meet the boron concentration of LCO 3.9.1.

AND A.2 Suspend loading irradiated Immediately fuel assemblies in the core.

AND A.3 Initiate action to satisfy Immediately DHR loop requirements.

AND BWOG STS 3.9.4-1 Rev. 3.0, 03/31/04 0

Attachment 1, Volume 14, Rev. 1, Page 60 of 142

Attachment 1, Volume 14, Rev. 1, Page 61 of 142 CTS DHRand Coolant Circulation. - High Water Level 3.9.4 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME Action a A.4 Close equipment hatch and secure with ýouJ bolts.

4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months Q

AND A5 Close one door in eachair 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months lock.

AND FVeri fy A6j Cs eachpenetration 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months providing directaccess fromethecontainment 0 is either closed atmosphere to the outside atmosphere with a manual or automatic isolation'valve, blind flange, orequivalen o

0 4o Verif eac enetratio is AB; capa be oMeing cosedby iah OPEKABLEI Containment Purge and Exhaust Isolation System:

SURVEILLANCE REQUIREMENTS ....

SURVEILLANCE FREQUENCY 4.8.9.1 SR 319.4.1 Verify oneDHR loop is in operation and circulating 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months reactor coolant at a flow rate of >42800(gprnm, 0

BWOG STS 3.9.4-2 Rev; 3.0, 03/31/04 0

Attachment 1, Volume 14, Rev. 1, Page 61 of 142

Attachment 1, Volume 14, Rev. 1, Page 62 of 142 JUSTIFICATION FOR DEVIATIONS ITS 3.9.4, DHR AND COOLANT CIRCULATION - HIGH WATER LEVEL

1. Editorial correction to be consistent with the format of the ITS.

2. The brackets are removed and the proper plant specific information/value is provided.

3. ISTS 3.9.4 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 3.9.3, 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 3.9.4 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. Furthermore, since ISTS 3.9.3 has not been adopted, the term OPERABLE has been deleted as requested by the NRC and as documented in RAI 200801161532.

Davis-Besse Page 1 of 1 Attachment 1, Volume 14, Rev. 1, Page 62 of 142

Attachment 1, Volume 14, Rev. 1, Page 63 of 142 Improved Standard Technical Specifications (ISTS) Bases Markup and Justification for Deviations (JFDs)

S 0

Attachment 1, Volume 14, Rev. 1, Page 63 of 142

Attachment 1, Volume 14, Rev. 1, Page 64 of 142 DHR and Coolant Circulation - High Water Level B 3.9.4 B 3.9 REFUELING OPERATIONS B 3.9-4 Decay Heat Removal (DHR) and Coolant Circulation - High'Water Level BASES BACKGROUND The purposes of the DHR System inMODE 6are to remove-decay heat UFSAR, Appendix ard sensible heat from the Reactor Coolant System (RCS), as required 3D.1.30 (Ref. 1) by 2_*, to provide mixing of borated coolant, to provide sufficient coolant circulation to minimize the effects of a boron dilution accident, and to prevent boron stratification (Ref. ). Heat is removed from the RCS by 0 ,

circulating reactor coolant through the DHR heat eyhanger(s where the co..er (1) core

________,x__

flood nozzles heat is transferred to the Component Cooling Water h

co stem angers. The coolant is then returned to the RCS via theR __S

. Operation of the DHR System for normal cooldown or decay 02 heat removal is manually accomplished from the control room. The heat removal rate is adjusted by control of the flow of reactor coolantthrough theDH heatexc n ors andb assin the heat changors. Mixing of the reactor coolant is maintained by this continuous Circulation of reactor coolant through the DHR System.

APPLICABLE If the reactor coolant temperature is not maintained below 200'F, boiling SAFETY of the reactor coolant could result. This could lead to inadequate cooling ANALYSES of the reactor fuel as a result of a loss of coolant in 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, and because of the possible addition of water to the reactor vessel with a lower boron concentration than is required to keep the reactor subcritical. The loss of reactor coolant and the reduction in boron concentration in the reactor coolant would eventually challenge the integrity of the fuel cladding, which is a fission product barrier. One train of the DHR System is required to be operational in MODE 6, with the water level 2 23 ft above the top of the reactor vessel flange, to prevent this challenge. The LCO does permit the DHR pump to be removed from operation for short durations under the condition that the boron concentration is not diluted. This conditional stopping of the DHR pump does not result in a challenge to the fission product barrier.

The DHR System satisfies Criterion 4 of 10 CFR 50.36(c)(2)(ii).

BWOG STS B 3.9.4-1 Rev. 3.0, 03/31/04 0

Attachment 1, Volume 14, Rev. 1, Page 64 of 142

Attachment 1, Volume 14, Rev. 1, Page 65 of 142 DHR and. Coolant Circulation.- High Water Level B 3.9.4 BASES LCO Only one DHR loop is required for decay heat removal in MODE 6, with a water level - 23 ft above the top of the reactor vessel flange. Only one.:

DHR loop is required to be OPERABLE because the.:volume of water above the reactor vessel flange provides backup.decay heat removal capability. At least one DHR loop must be OPERABLE and in operation to provide:-

a. Removal of decay heatl f 00

b. Mixing of borated coolant to minimize the possibility of criticalityfj } 0 Ic. Indipfition of reac~r coolant to perature. 0 An OPERABLE DHR loopincludes.a.DHR pump'aaheat 9change

valves, piping, instruments, and controls to ensure an OPERABLE flow since the DHR System is a manually operated system (i.e., it is not -

pathland to dttermine-tholow end teneratur; The flow ath starts in.

one ofthe RCS hot legs and. is returned to theiRCS c d legs 0

(D 6

automatically actuated), cr lo oz uoc Additionally, each DHR loop is cons ered OPERABLE if it canbe

L(

to manually aligned (remote or 10caljE5the hutdovy cooling .node IK£] decay ha Iremoval o~fdecay hea(* Operation Of one subsystom can maintain the I Ireactor coo ant temp/ rature%as required; ..

The LCO is modified by a Note that allowsthe required DHR loop to be removed from operation for up to 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months in an8 hour period, provided no operations are permitted that would dilute the RCS boron concentration by introduction of coolant into the RCS with boron concentrationi less than " Boron required to meet the minimum boron concentration of LCO 3.9,1 DBoron concentration."

concentration reduction with coolant at boron concentrations less than requiredto assure the RCS boron concentration is maintained is 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 DHR isolation valve testing. During this 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months period, decay heatis removed by natural convection to the large mass~of water.in the refueling BWOG.STS B 3:9;4-2 Rev. 3.0, 03131104 0

Attachment 1, Volume 14, Rev. 1, Page 65 of 142

Attachment 1, Volume 14, Rev. 1, Page 66 of 142 DHR and Coolant Circulation - High Water Level 8 3.9.4 BASES APPLICABILITY One -DHRloop must be OPERABLE and in operation in MODE 6, With the water level ? 23 ft above the top of the reactor vessel flange, to provide decay heat removal, The 23 ft water level was selected because it corresponds to the 23 ft requirement established for fuel movementin LCO 3,9.6, "Refueling Canal Water Level." Requirements for the DHR System in other MODES are covered by LCOs in Section 3.4, Reactor Coolant System (RCS), and SeAion 3.5, Er'rqency C.e Cooling

[ysr(ECCS. DHR loop requirements in MODE 6, with the water 0

level < 23 ft above the top of the reactor vessel flange, are located in LCO3.9.5, "Decay Heat Removal (DHR) and Coolant Circulation - Low Water Level."

ACTIONS DHR loop requirements are met by having one DHR loop OPERABLE and in operation, except as permitted in the Note to the LCO.

A.1 If DHR 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 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 DHR loop requirements are not met, actions shall be taken immediately to suspend the 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 23 ft above the reactor vessel flange provides an adequate available heat sink. Suspending any operation that would increase decay heat load, such as loading a fuel assembly, is prudent under this condition.

BWOG STS B 3.9.4-3 Rev. 3.0, 03/31/04 Attachment 1, Volume 14, Rev. 1, Page 66 of 142

Attachment 1, Volume 14, Rev. 1, Page 67 of 142 DHR and&Coolant Circulation - High Water Level B13.9.4 BASES ACTIONS (continued)

A.3 If DHR loop requirements are not met, actions shall be initiated immediately in order to satisfy DH R looprequirements.

AA4, A.5.ýA.6ag 0

If:no DHR is.in operation, the following actions must be taken:

.a. The equipment hatch must be Closed and.secured with rfoud boltsm.-Fl 000

b. One.door in each air lock must be closeclsnd_.

c. Each penetration providing direct access from the containment atmosphere to the outside atmosphere must be either closed by a.

manua -or automatic isolation valve, blind flange or equivalent, or verified to be capable of being closed by aLQ. aE i 0

,Containment Purge and Exhaust Isolation System.

0 With DHR 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 canmbe closed so that the dose.1limits are not exceeded.

The Completion Time of 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months allows fixing of most'DHR problems and.

isreasonable, based on the low probability of the coolant boiling in that time.

SURVEILLANCE SR, 3.9.4.1 REQUIREMENTS.

This Surveillance demonstrates that the DHR loop is in operation and circulating reactor coolantý The.flow rate is determined by the flow rate.

necessary to provide sufficient decay heat removal-capability and to prevent thermaland boron stratification in the core., The Frequency of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is sufficient, considering the flow, temperature, pump control, and alarm indications available to the operator in the control room for monitoring the DHR -System.

REFERENCES t FSAR econ

1. UPSAR, Appendix 3D.1.30, Criterion 34- Residual Heat Removal 0001 BVWOG STS B 3.9.4-4 Rev. 3.0,03/31/04 0

Attachment 1, Volume 14, Rev. 1, Page 67 of 142

Attachment 1, Volume 14, Rev. 1, Page 68 of 142 B 3.9.4 0 INSERT I A Containment Purge and Exhaust Isolation System consists of a containment purge and exhaust noble gas monitor, including all automatic actuations resulting from a high radiation signal (i.e., the shutting down of the containment purge and exhaust supply and exhaust fans and closure of the associated inlet and outlet dampers), and one containment purge and exhaust isolation valve in each penetration flow path, which is capable of being manually closed from the control room.

0 Insert Page B 3.9.4-4 Attachment 1, Volume 14, Rev. 1, Page 68 of 142

Attachment 1, Volume 14, Rev. 1, Page 69 of 142 JUSTIFICATION FOR DEVIATIONS ITS 3.9.4 BASES, DHR AND COOLANT CIRCULATION - HIGH WATER LEVEL

1. 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.

2. Editorial changes made for clarity or to be consistent with the format of the ITS.

3. The brackets have been removed and the proper plant specific information/value has been provided.

4. Changes are made to reflect changes made to the Specification.

5. These punctuation corrections have been made consistent with the Writer's Guide for the Improved Standard Technical Specifications, TSTF-GG-05-01, Section 5.1.3.

6. Changes have been made to be consistent with the Specification.

7.. The wording has been modified since Section 3.5 does not provide requirements for the DHR function.

8. This redundant sentence has been deleted. The operation requirement is already discussed in the first paragraph of the LCO Bases.

0 Davis-Besse Page 1 of 1 Attachment 1, Volume 14, Rev. 1, Page 69 of 142

Attachment 1, Volume 14, Rev. 1, Page 70 of 142 W Specific No Significant Hazards Considerations (NSHCs) 0 0

Attachment 1, Volume 14, Rev. 1, Page 70 of 142

Attachment 1, Volume 14, Rev. 1, Page 71 of 142 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.9.4, DHR AND COOLANT CIRCULATION - HIGH WATER LEVEL There are no specific NSHC discussions for this Specification.

0 0

Davis-Besse Page 1 of 1 Attachment 1, Volume 14, Rev. 1, Page 71 of 142

Attachment 1, Volume 14, Rev. 1, Page 72 of 142 ATTACHMENT 5 ITS 3.9.5, DHR AND COOLANT CIRCULATION - LOW WATER LEVEL Attachment 1, Volume 14, Rev. 1, Page 72 of 142

Attachment 1, Volume 14, Rev. 1, Page 73 of 142 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) 0 0

Attachment 1, Volume 14, Rev. 1, Page 73 of 142

Attachment 1, Volume 14, Rev. 1, Page 74 of 142 ITS 3.9.5 ITS I REFUEL *IG OPERATIONS LOl WATER LEVEL LIMITING CONDITION FOR OPERATION LAO1 M0 F and one loop shall be in operat ion - "

LCO 3.9.5 3.9.8.2 Two inde endent DHR loops shall be OPERABLE. Add proposed LCO NOTE 1 A02 Add proposed LCO NOTE 2 APPLICABILITY: MODE 6asserrblies

(~~fuel when the water level ,above the top of the ý_rradiate4_

spirted within the reactor presue v~essel. is:

- 01 L01*/

less than 23 feet. fla.nge ACTION: M02 ACTIONS A a. With less than the required DHR loops OPERABLE., inmediately initiate and B corrective action to return the required loops to OPERABLE status as soon as possible. BA.adB5frtoip eral Add proposed Required Actions B.1,oops op.3, A03 I b.. The. proy-rsions of Specifif~ation 3.U.3 are/fot app lcab e.*-A

- - " I ~Add proposed A C7TION B for " * / '

loop not in operation M0-SURVEILLANCE REQUIREMENTS SR 3.9.5.1 4.9.8.2 At least one DHR loop shall, be detemined to be in'operation per Speci..

fication 4.9.8.1. , The inactiyv loop shall be detemined to be OPERABLE perL 1speci fcati/on 4.0.:*

SAdd proposed SR 3.9.5.2 M04_(/

1 enonna e e y owersourpae" may be inoperabl'or eac DHR 6oop. A02 DAVIS-BESSE, UNIT 1, 3/4 9-8a Amendment No. 38 Page 1 of 2 Attachment 1, Volume 14, Rev. 1, Page 74 of 142

Attachment 1, Volume 14, Rev. 1, Page 75 of 142 ITS 3.9.5 ITS REFUELING OPERATIONS 314.9.8 DECAY HEAT REMOVAL AND COOLANT CIRCULATION ALL WATER LEVELS LIMITING CONDITION FOR OPERATION 3.9.8.1 At least one decay heat removal loop shall be in operation.

I APPLICAB1LITY: .MODE 6 when the water level above the top of the irradiated fuel assemblies seated within the reactor. pressure vessel is.

> 23 feet.

ACTION- See ITS 3.9.4 J

a. With less than one decay heat removal loop in operation, except as provided in b below, suspend all operations involving an increase in the reactor decay heat load or a reduction in boron concentration of the Reactor Coolant System. Close all containment penetrations providing direct access from the containment atmosphere to the outside atmosphere within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

b. The decay heat removal loop may be removed from operation for up to one hour per 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months period during the performance of CORE ALTERATIONS in the vicinity of the reactor pressure vessel (hot) legs.

c. The provisions ofSpecification 3A.3. are. not applicable.

SURVEILLANCE REQUIREMENTS SR 3.9.5.1 4.9.8..] Surveillance at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months shall verify at least one decay heat removal loop to be in operation and circulating. reactor coolant through the reactor core:

a. At a flow ate of > 2800 gpm, w enever a reduction in Reactor Coolant Syrtem boron concentration is being madeý.

b. At a flow rate such that the co e outlet temperatu e is maintained L03

< 140"F, rovided no reduction in Reactor Coolant ystem boron concentra ion is being made.

Water of a lower boron concen ration than the existing RCS concentr tion may be added to the RCS, with the lowrate of reactor oolant through the.R S less than 2800 g m, provided that the bor n concentration of th water to be added is equal to or greater than the boron conce tration correspondi g to the more res.tric ive reactivity condi ion specified in S ecification 3.9.1.

PAVIS-BESSE, UNIT 1. 3/4 9-8 Amendment No. S8,188-Page.2 of 2 Attachment 1, Volume 14, Rev. 1, Page 75 of 142

Attachment 1, Volume 14, Rev. 1, Page 76 of 142 DISCUSSION OF CHANGES ITS 3.9.5, DHR AND COOLANT CIRCULATION - LOW WATER LEVEL ADMINISTRATIVE CHANGES A01 In the conversion of the Davis-Besse 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-1430, Rev. 3.1, "Standard Technical Specifications-Babcock and Wilcox Plants" (ISTS).

These changes are designated as administrative changes and are acceptable because they do not result in technical changes to the CTS.

A02 CTS 3.9.8.2 is modified by footnote *, which states that the normal or emergency power source may be inoperable for each DHR 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.

A03 CTS 3.9.8.2 Action a states that with less than the required DHR loops OPERABLE, immediately initiate corrective action to return the required DHR 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 the 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.

A04 CTS 3.9.8.2 Action b states, "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. CTS 3.9.8.2 and ITS 3.9.5 are only applicable in MODE 6. 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 M01 CTS 4.9.8.2 requires that at least one DHR loop be determined to be in operation per Specification 4.9.8.1, the DHR loop flow rate verification. However, CTS Davis-Besse Page 1 of 6 Attachment 1, Volume 14, Rev. 1, Page 76 of 142

Attachment 1, Volume 14, Rev. 1, Page 77 of 142 DISCUSSION OF CHANGES ITS 3.9.5, DHR AND COOLANT CIRCULATION - LOW WATER LEVEL LCO 3.9.8.2 does not require a DHR loop to be in operation; it just requires two DHR loops to be OPERABLE, and no Actions are provided if a DHR loop is not in operation. ITS 3.9.5 requires one of the DHR loops to be in operation, as modified by the LCO 3.9.5 Note 1 allowance. In addition, ITS 3.9.5 ACTION B provides the actions when the required DHR loop is not in operation. This changes the CTS by providing requirements for one DHR loop to be in operation and appropriate actions when the DHR loop is not in operation.

The purpose of CTS 3.9.8.2ý is to ensure adequate DHR is OPERABLE for heat removal and coolant circulation. This change is acceptable because it provides the necessary requirements to ensure the above purpose is met. ITS LCO 3.9.5 requires one DHR loop to be in operation, as modified by the LCO 3.9.5 Note 1 allowance. LCO 3.9.5 Note 1 allows all DHR pumps to be removed from operation for < 15 minutes when switching from one train to the other provided the core outlet temperature is maintained > 100 F below saturation temperature, no operations are permitted that would cause introduction of coolant into the RCS with boron concentration less than that required to meet the minimum required boron concentration of LCO 3.9.1, "Boron Concentration," and no draining operations to further reduce RCS water volume are permitted. ITS 3.9.5 ACTION B provides the actions when the required DHR loop is not in operation.

This ACTION requires immediate suspension of operations that would cause introduction of coolant into the RCS with boron concentration less than that required to meet the minimum required boron concentration of LCO 3.9.1 (ITS 3.9.5 Required Action B.1), immediate initiation of action to restore one DHR loop to operation (ITS 3.9.5 Required Action B.2), and requires within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , the equipment hatch to be closed with four bolts (ITS 3.9.5 Required Action B.3), one door in each air lock to be closed (ITS 3.9.5 Required Action B.4), and a verification that 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 a Containment Purge and Exhaust Isolation System (ITS 3.9.5 Required Action B.5). These actions assist in minimizing the consequences of a DHR loop not being in operation. This change is designated as more restrictive because an LCO requirement is being added to the ITS that is not required by the CTS.

M02 CTS 3.9.8.2 requires two DHR loops to be in OPERABLE in MODE 6 when the water level above the top of the irradiated fuel assemblies seated within the reactor pressure vessel is < 23 feet. ITS 3.9.5 requires two DHR loops to be OPERABLE and one in operation when water level is < 23 feet above the top of the reactor vessel flange. This changes the CTS by changing the point at which either one or two DHR loops are required to be OPERABLE and one in operation. The change requiring the DHR loop to be in operation is discussed in DOC M01.

The purpose of CTS 3.9.8.2 is to ensure adequate DHR is OPERABLE for heat removal and coolant circulation. CTS 3.9.8.1 and CTS 3.9.8.2 provide the requirements when water level is > 23 feet and < 23 feet above the top of the irradiated fuel assemblies seated within the reactor pressure vessel, respectively.

When water level is > 23 feet above the top of the irradiated fuel assemblies seated within the reactor pressure vessel, only one DHR loop is required to be in Davis-Besse Page 2 of 6 Attachment 1, Volume 14, Rev. 1, Page 77 of 142

Attachment 1, Volume 14, Rev. 1, Page 78 of 142 DISCUSSION OF CHANGES ITS 3.9.5, DHR AND COOLANT CIRCULATION - LOW WATER LEVEL operation (and essentially OPERABLE). When water level is < 23 feet above the top of the irradiated fuel assemblies seated within the reactor pressure vessel, two DHR loops are required to be OPERABLE, and one must be in operation. In ITS 3.9.4 and ITS 3.9.5, the equivalent ITS requirements, the water level reference point is the top of the reactor vessel flange, not the top of the irradiated fuel assemblies seated within the reactor pressure vessel. Changing this reference point effectively requires a larger complement of DHR loops to be OPERABLE between 23 feet above the top of the irradiated fuel assemblies seated within the reactor pressure vessel and 23 feet above the top of the reactor vessel flange. Therefore, this change is acceptable because more loops will be required to be OPERABLE under certain water level conditions to ensure the decay heat can be removed and the coolant circulated. This change is designated more restrictive because more DHR loops are required OPERABLE in the ITS under certain water level conditions than were required in the CTS.

M03 CTS 3.9.8.2 Action a states that with less than the required DHR loops OPERABLE, immediately initiate corrective action to return the required DHR loops to OPERABLE status as soon as possible. ITS 3.9.5 ACTION B includes the same requirement, but also includes additional requirements when both DHR loops are inoperable. This changes the CTS by requiring additional actions when both DHR loops are inoperable.

The purpose of CTS 3.9.8.2 is to ensure adequate DHR is OPERABLE for heat removal and coolant circulation. This change is acceptable because it provides the necessary requirements to ensure the above purpose is met. ITS 3.9.5 ACTION B provides the actions when both DHR loops are inoperable. This ACTION requires immediate suspension of operations that would cause introduction of coolant into the RCS with boron concentration less than that required to meet the minimum required boron concentration of LCO 3.9.1 (ITS 3.9.5 Required Action B.1), immediate initiation of action to restore one DHR loop to operation (ITS 3.9.5 Required Action B.2), and requires within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , the equipment hatch to be closed with four bolts (ITS 3.9.5 Required Action B.3), one door in each air lock to be closed (ITS 3.9.5 Required Action B.4), and a verification that 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 a Containment Purge and Exhaust Isolation System (ITS 3.9.5 Required Action B.5). These actions assist in minimizing the consequencesof both DHR loops being inoperable. This change is designated as more restrictive because Required Actions are being added to the ITS that are not required by the CTS.

M04 The CTS 3.9.8.2 requires two independent DHR loops to be OPERABLE. ITS SR 3.9.5.2 requires verification every 7 days of correct breaker alignment and that indicated power is available to the required DHR pump not in operation. A Note states that the Surveillance Requirement is not required to be performed until 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after a required DHR 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 DHR loop to be in operation and one DHR loop to be held in readiness should it be needed. This change is Davis-Besse Page 3 of 6 Attachment 1, Volume 14, Rev. 1, Page 78 of 142

Attachment 1, Volume 14, Rev. 1, Page 79 of 142 DISCUSSION OF CHANGES ITS 3.9.5, DHR AND COOLANT CIRCULATION - LOW WATER LEVEL acceptable because it verifies that the DHR 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 LA01 (Type I - Removing Details of System Design and System Description, Including Design Limits) CTS 3.9.8.2 states that two "independent" DHR loops shall be OPERABLE. ITS 3.9.5 requires two DHR loops to be OPERABLE, but does not contain the detail that the loops must be independent. This changes the CTS by moving the detail that the DHR loops are independent to the Bases.

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 the requirement for two DHR loops to be OPERABLE. Also, 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 L01 (Category I - Relaxation of LCO Requirements) ITS 3.9.5 is modified by LCO Note 2, which allows one required DHR loop to be inoperable for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for Surveillance testing, provided that the other loop is OPERABLE and in operation. CTS 3.9.8.2 does not contain this allowance. This changes the CTS by allowing the LCO to not be met under certain situations.

The purpose of 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. ITS 3.9.5 continues to require one DHR loop to be OPERABLE and in operation when using this Note allowance, which will ensure sufficient decay heat removal capability exists. ITS 3.9.5 Note 2 allows normal operational evolutions, i.e., Surveillance testing, to be performed while in the Applicability of the Specification. These Surveillances are necessary to demonstrate DHR System OPERABILITY or OPERABILITY of other systems. Furthermore, the ITS Bases states that prior to making one of the DHR loops inoperable and utilizing this Note allowance, consideration should be given to the existing plant Davis-Besse Page 4 of 6 Attachment 1, Volume 14, Rev. 1, Page 79 of 142

Attachment 1, Volume 14, Rev. 1, Page 80 of 142 DISCUSSION OF CHANGES ITS 3.9.5, DHR AND COOLANT CIRCULATION - LOW WATER LEVEL configuration. This consideration should include time to core boiling, potential for RCS draindown, and RCS makeup capability. These considerations will further minimize the probability and consequences of a loss of the remaining DHR loop while using this Note allowance. This change is designated as less restrictive because less stringent LCO requirements are being applied in the ITS than were applied in the CTS.

L02 (Category 5 - Deletion of Surveillance Requirement) CTS 4.9.8.2 requires verification that the inactive DHR 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 DHR 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 DHR loop OPERABILITY to the Inservice Testing Program. This change is acceptable because it is not necessary to perform inservice testing of a DHR loop to determine if it is OPERABLE, as the system is routinely operated and the DHR loops are instrumented so that degradation can be observed. Significant degradation of the DHR System would be indicated by the DHR 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.

L03 (Category 6- Relaxation Of Surveillance RequirementAcceptance Criteria)

CTS 4.9.8.1 verifies that the DHR loop is in operation and circulating reactor coolant and provides two flow rate requirements. CTS 4.9.8.1.a requires

> 2800 gpm when a reduction in boron concentration is in progress and CTS 4.9.8.1 .b requires a flow rate sufficient to maintain core outlet temperature

< 140OF when a reduction in boron concentration is not in progress. ITS SR 3.9.5.1 requires a similar Surveillance, but does not include a specific flow rate requirement. This changes the CTS by deleting the DHR loop flow rate requirement.

The purpose of CTS 4.9.8.1 is to ensure that the DHR loop is in operation. This change is acceptable because the ITS continues to require a DHR loop to be in operation, and this requirement is verified every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months in ITS SR 3.9.5.1.

During MODE 6 conditions, the reactor is cooled down and the decay heat load varies with time. Therefore, stating a flow rate that must be met at all times is overly conservative with regard to removing the actual decay heat load that is present. Davis-Besse normally maintains temperature < 140OF during MODE 6 operations. As stated in the ISTS Bases, the flow rate is determined by the flow rate necessary to provide efficient decay heat removal capability and prevent thermal and boron stratification in the core. Thus, this will ensure that adequate flow is maintained without a specific flow rate requirement being in the ITS. This Davis-Besse Page 5 of 6 Attachment 1, Volume 14, Rev. 1, Page 80 of 142

Attachment 1, Volume 14, Rev. 1, Page 81 of 142 DISCUSSION OF CHANGES ITS 3.9.5, DHR AND COOLANT CIRCULATION - LOW WATER LEVEL change is designated as less restrictive because a Surveillance Requirement acceptance criterion included in the CTS is not included in the ITS.

Davis-Besse Page 6 of 6 Attachment 1, Volume 14, Rev. 1, Page 81 of 142

Attachment 1, Volume 14, Rev. 1, Page 82 of 142 Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs)

Attachment 1, Volume 14, Rev. 1, Page 82 of 142

Attachment 1, Volume 14, Rev. 1, Page 83 of 142 CTS DHR andCoolant Circulation - Low Water Level 3.9:5 3.9 REFUELING OPERATIONS 3-9.5 Decay Heat Removal (DHR) and Coolant Circulation- Low Water Level 3.9.8.2 LCO 3.9.5 Two DHR loops shall be OPERABLE, and one DHR loop shall bein operation.

DOCs M01 1. All DHR pumps may be removed from operation for < 15 minutes and L01 when switching from one train to another provided:

a. The core outlet temperature is maintained > 10 degrees F below saturation temperatureM_ F-1 0

b. No operations are permitted that would cause introduction of coolant into the Reactor Coolant Systemfvith boron (R(S concentration less than that requiredto meet the minimum required boron concentration of LCO 3.9.1,

_ on

c. No draining operations to further reduce RCS water volume are permitted.

2. One required DHR loop may be inoperable for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for surveillance testing, provided that the other DHR loop is OPERABLE and in operation.

APPLICABILITY: MODE 6 with the water level < 23ft above the top of reactor vessel flange.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME Action a A. Less than required A.1 Initiate action to restore Immediately number of DHR loops DH R loop to OPERABLE OPERABLE. status.

OR A.2 Initiate action to establish Immediately

> 23 ft of water above the top of reactor vessel flange.

BWOG STS 3.9.5-1 Rev. 3.0, 03/31/04 0

Attachment 1, Volume 14, Rev. 1, Page 83 of 142

Attachment 1, Volume 14, Rev. 1, Page 84 of 142 CTS DHR and Coolant Circulation - LowWateriLevel 3.9.5 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME Action a, B. No DHR loop B.1 Suspend operations that Immediately DOC M01 OPERABLE or in would cause introduction of operation. coolant into the RCS with boron concentration less than required to meet the boron concentration of LCO 3.9.1.

AND B.2 Initiate action to restore one Immediately DHR loop to OPERABLE status and to operation.

AND I,

B.3 Close equipment hatch and 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months secure with [our bolts. 0 AND B.4 Close one door in each air 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months lock.

AND Verify B.S*] Ieeach penetration 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months providing direct access from the containment atmosphere to the outside is either closed atmospherehvith a manual or automatic isolation valve, blind flange, or equivalen FW I orj from next page BWOG STS 319.5-2 Rev. 3.0, 03/31/04 0

Attachment 1, Volume 14, Rev. 1, Page 84 of 142

Attachment 1, Volume 14, Rev. 1, Page 85 of 142 CTS DHR and Coolant Circulation - Low Water Leve!

3.9.5 move to previous page ACýTIONS (,contjiued)

CONDITION REQUIRED ACTION . ]_tCOMF*LETI ON TIME.

Action a, DOC M01 IB5.25

/

Vei:W7each penetrationnis 0

capable of being closed by Containment Purge and Exhaust.Isolation.System, SURVEILLANCE REQUIREMENTS ..........

SURVEILLANCE FREQUENCY 4.9.8.1, 4.9.8.2 SR 39,5.1 Verify one DHR loop is in operation. 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months s-DOC M04 *SR 3.9.5.2 Verify correct breaker alignment and indicated 7days poZweravailable to the required DHR pump that is not in operation.

F 2 Not required to be performed until 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after a required pump is not in operation.

I 0 BWOG STS 3.9.5-3 Rev. 3.0, 03/31/04 Attachment 1, Volume 14, Rev. 1, Page 85 of 142

Attachment 1, Volume 14, Rev. 1, Page 86 of 142 JUSTIFICATION FOR DEVIATIONS ITS 3.9.5, DHR AND COOLANT CIRCULATION - LOW WATER LEVEL

1. These punctuation corrections have been made consistent with the Writer's Guide for the Improved Standard Technical Specifications, TSTF-GG-05-01, Section 5.1.

2. Editorial change to be consistent with the format of the ITS.

3. The brackets have been removed and the proper plant specific information/value has been provided.

4. ISTS 3.9.5 Required Actions A.5.1 and A.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 A.5.1 or Required Action A.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 3.9.3, 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.5.1 and A.5.2 have been combined into a single Required Action in ITS 3.9.5 Required Action A.5. Furthermore, since ISTS 3.9.3 has not been adopted, the term OPERABLE has been deleted as requested by the NRC and as documented in RAI 200801161532.

5. TSTF-265 was previously approved and incorporated in NUREG-1430, Rev. 2, in similar SRs (e.g., ISTS SRs 3.4.5.2, 3.4.6.2, 3.4.7.3, and 3.4.8.2). Consistent with TSTF-265, a Note is added to ISTS SR 3.9.5.2 that permits the performance of the SR to verify correct breaker alignment and power availability to be delayed until 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 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 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to perform the breaker alignment verification is acceptable because the pump was in operation, which demonstrated OPERABILITY, and because 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months is currently allowed by invoking SR 3.0.3.

This is a new Surveillance Requirement not required in CTS 3.9.8.2.

0 Davis-Besse Page 1 of 1 Attachment 1, Volume 14, Rev. 1, Page 86 of 142

Attachment 1, Volume 14, Rev. 1, Page 87 of 142 Improved Standard Technical Specifications (ISTS) Bases Markup and Justification for Deviations (JFDs)

Attachment 1, Volume 14, Rev. 1, Page 87 of 142

Attachment 1, Volume 14, Rev. 1, Page 88 of 142 DHR and Coolant Circulation - Low Water Level B 39.5 B 3.9 REFUELING OPERATIONS B'3.9.5 Decay Heat Removal (DHR) and Coolant Circulation - Low Water. Level BASES BACKGROUND The purposes of the DHR System in MODE 6 are to remove decay heat.

UFSAR, Appendnd nd ible heat from the Reactor Coolant System (RCS), as. required 3D.1.30 (Ref. 1) Jb3i to provide mixing of borated coolant, to provide sufficient coolant.circulation to minimize the. effects of a boron dilution accident, and, WD to prevent boron, stratification (Ref4.-). Heat is removed from the.RCS by----ier _

circulating reactor coolant through the DHR heat e *anger swhere thece heat is transferred to the.Cornponent Cooling Water System t heat cs'tanserr The, coolant is then returned to the RCS via, the S core flood nozzles cold Operation of the DHR System for normal cooldown/decay heat removal is manually accomplished.from theicontrol room, The heat.

ooes removal the DHR rate is adjusted by control ha~xb'grs n v~sn of the flow h of eat'ef reactorcharter(sY; coolant through Mixing: *'

of the reactor coolant is maintained by this continuous circulation of, reactor coolant through'the DHR System.

APPLICABLE If the. reactor coolant temperature is not maintained below 200F, boiling SAFETY. of the reactor coolant'could result. This could lead to inadequate cooling 0 ANALYSES of the reactor fuel dueito-resulting loss of coolant in 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, and because of the possible additi6n~of water to the reactor vessel with a lower boron concentration than is.required to.keep'the reactor'subcritical. The:loss.of reactor coolant and the reduction' of boron concentration in the reactor coolant would eventually challenge theintegrity of the fuel cladding, which is a fission product barrier. Two trains of the DHR System are required to be OPERABLE, and one is required to be in operation, to prevent this challenge.

The DHR System satisfies Criterion 4 of 10 CFR 50.36(c)(2)(ii).

LCO i t In MODE 6, with the water'level < 23 ft above the top of the reactor vessel Lp flange, two HR loops must be OPERABLE. Additionally, one DHR loop' must be in operation to'provide:

0

a. Removal.of decay heat,.I7E) 00

b. Mixing of borated coolant to minimize the possibility of criticality tjfi.-V 0

Ic. Indi9otion ot reactpr coolant teytiperature. I BVWOG STS B 3.9.5-1 Rev. 310, 03/31/04 0

Attachment 1, Volume 14, Rev. 1, Page 88 of 142

Attachment 1, Volume 14, Rev. 1, Page 89 of 142 DHR and Coolant .Circulation - Low Water Level B 3.9.5 BASES LCO (continued)

This LCO is modified.by.two Notes. Note 1 permits the DHR pumps to be removed from operation for -<15minutes when switching from one trainto another. The circumstances for stopping both DHR pumps are to be limited to situations when the outage time is short aind the core outlet (5 temperatre is mintaind > 10 degrees F below satu0ration.temperatur4 The Noteinto coolant prohibits the RCSboron with dilutio*{¶]draining operations.

boron concentrations, introduction less thanbyrequired of to meet the minimum boron-concentration of LCO 3.9.1 when DHR forced flow is stopped. "Boron Concentration,"

Note2 allows.one DHR loop to be.inoperable for a period.of 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months provided the otherloop is OPERABLE and in operation. Prior to declaring the loop existinqgptant inoperable,TFhis configuration., consideration should be given consideration-should to the include*

___ core time to/boil is short, there is no//draining operation to furthh reduce INSERT I RCS watef-level and that.the ca p0ility exists to inect borated ter into the react r vessel. 1This permits surveillancetests to be performed on the inoperable Ioopduring a time when these tests are safe and.possib!e.

AnOPERABLE DHR loop consists of a DHR pump, a heat echanaee ()

valves, pipingj instruments, and controls to ensure an OPERABLE flow pat h land to d~termine th ow endtperature. The flow path starts in one of~the RCS hot legs and is returned to the ,

Both DHR pumps be aligned to the Refueling Wa Storage Tank to ISR2 support filling or ining the refueling cavity or forbpormrance of required testin.

APPLICABILITY Two DHR loopsare required to be OPERABLE, and one 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 DHR System in other MODES are covered by LCOs in Section 3.4,iReactor CoolantSystem (RCM , an ection -eren ore ooing ste DHR loop requirements in MODE 6; with the water 00 level >_23.ft above the top of the reactor vessel flange, are located in LCO 3.9.4,"'Decay Heat Removal (DHR) and Coolant Circulation - High Water Level."

ACTIONS A.1 and A,2 With fewer than the:required-loops OPERABLE, action shall be immediately initiated and continued until the DHR loop is restored to OPERABLE status or until -> 23 ft of water level is established above the reactor vessel flange. When the water level is established at a 23 ft above the reactor vessel flange, the Applicability will change to that of BWOG STS B 3.9.572 Rev. 3.0, 03/31/04 Attachment 1, Volume 14, Rev. 1, Page 89 of 142

Attachment 1, Volume 14, Rev. 1, Page 90 of 142 B 3.9.5 O INSERT 1 time to core boiling, potential for RCS draindown, and RCS makeup capability.

INSERT 2 Additionally, since the DHR System is a manually operated system (i.e., it is not automatically actuated), each DHR loop is OPERABLE if it can be manually aligned (remote or local) to the decay heat removal mode.

Insert Page B 3.9.5-2 Attachment 1, Volume 14, Rev. 1, Page 90 of 142

Attachment 1, Volume 14, Rev. 1, Page 91 of 142 DHR and Coolant Circulation - Low Water.Level B 3.9.5 BASES ACTIONS (continued)

LCO"3.9.4, and only one DHR loop is required to be OPERABLE and in operation. An immediate Completion Time is necessary for an operator

to. initiate corrective actions to restore the required *forced circulation or w* ter level.

B.1 If no DHR loopis in operation or no .DHR Ioop is. OPERABLE, there will beino forced circulation to provide mixingto establish *uniform boron concentrations; 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 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.

B.2 If no DHR loop is in operation or no DHR loop is OPERABLE,.actions shall be initiated immediately and continueddtouterrupio to restore one DHR loop to OPERABLE status and operation. Since the unit is in Conditions A and B concurrently, the restoration of two OPERABLE DHR loops and one operating DHR loop should be accomplished expeditiously.

If no DHR loop is OPERABLE or in operation, alternate actions-shall have been initiated immediately under Condition A to establish _>23 ft of water above the top of the reactor vessel flange. Furthermore, when the LCO cannot be fulfilled, alternate decay heat removal methods, as specified in the unit's Abnormal and Emergency Operating Procedures, should be Shigh pressure injection, *implemented. This includes decay heat removal using the c arging orl makeup, or other ýý aft ijcinpm/thogteCeialand Volume Control Syste inecio souces writh consideration for the boron concentration. The method used to remove decay heat should be the most prudent as well as the safest choice, based upon unit conditions. The choice could be different if the reactor vessel head is in place rather than removed.

If no DHR is in operation, the following actions must be taken:

a. The equipment hatch must be closed and secured with fourl bolts[:tL*L ; ( hI(

BWOG STS B 3.9.5-3 Rev. 3.0, 03/31/04 Attachment 1, Volume 14, Rev. 1, Page 91 of 142

Attachment 1, Volume 14, Rev. 1, Page 92 of 142 DHR and Coolant Circulation'- Low Water Level B 3.9.5 BASES ACTIONS (continued)

b. One. door in each air lock must be closec 0 c- Each penetration providing direct access from the containment athosphere'to the outside atmosphere must be either closed by a.

manual or automatic isolation valve, blind flange, or e uivalent, or verified to be capable of being. closed by n<QEPABl Containment Purge and Exhaust Isolation System. I 00 With DHR, loop requiremennts not met, the potential exists for the coolant to boilzand release radioactive gasto theýcontainmentatmosphere.

Performing the actions'stated above ensures that all containment penetrations are either closed or can-be closed so that the dose limits are not exceeded.

The Completibn Time of 4,hours allows fixing of most DHR problems and is reasonable, based on thellow probability oftthe.coolant boiling in that time.

SURVEILLANCE SR 3.9,5.1 REQUIREMENTS This Surveillance demonstrates that one.DHR loop is in operation. The flowrate is determined by the flow rate necessary to provide efficient decay heat removal capability and to prevent thermal and boron stratification in the core.,

In addition, during operation of the DHR loop with theýWater level in the vicinity of the reactor vessel nozzles, the DHR loop flow rate determination~must also consider.the DHR pump suction requirement.

The Frequencyof 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is sufficient, considering the flow, temperature, pump control, and alarm indications available to the operator to: monitor, the. DHRI.System in:the control room.

SR 39;5,2 Verification that the required pump is OPERABLE ensures that an additional DHR pump can be placed in operation, if needed, to maintain decay heat removal and reactor coolant circulation. Verification is perfoirmed by verifying proper breaker alignment and Power available to the required pump. The Frecquency of 7 days is considered reasonable in Eacceptable Viewof other administrative controls available and has been shown to be by operating:experience. 09ý REFERENCES2 K4FSAR, Section 'ý' ___ 1. UFSAR, Appendix 3D.1.30,

=5fCntriterion 34- Residual Heat Removal.

000(D BWOG STS B 319.5-4 Rev. 3:0, 03/31/04 Attachment 1, Volume 14, Rev. 1, Page 92 of 142

Attachment 1, Volume 14, Rev. 1, Page 93 of 142 B 3.9.5 (DINSERT 3 0

A Containment Purge and Exhaust Isolation System consists of a containment purge and exhaust noble gas monitor, including all automatic actuations resulting from a high radiation signal (i.e., the shutting down of the containment purge and exhaust supply and exhaust fans and closure of theassociated inlet and outlet dampers), and one containment purge and exhaust isolation valve in each penetration flow path, which is capable of being manually closed from the control room.

INSERT 4 This SR is modified by a Note that states the SR is not required to be performed until 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after a required pump is not in operation.

Insert Page B 3.9.5-4 Attachment 1, Volume 14, Rev. 1, Page 93 of 142

Attachment 1, Volume 14, Rev. .1, Page 94 of 142 JUSTIFICATION FOR DEVIATIONS ITS 3.9.5 BASES, DHR AND COOLANT CIRCULATION - LOW WATER LEVEL

1. 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.

2. Editorial changes made for clarity or to be consistent with the format of the ITS.

3. These punctuation corrections have been made consistent with the Writer's Guide for the Improved Standard Technical Specifications, TSTF-GG-05-01, Section 5.1.3.

4. Changes have been made to be consistent with the Specification.

5. The brackets have been removed and the proper plant specific information/value has been provided.

6. 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.

7. The wording has been modified since Section 3.5 does not provide requirements for the DHR function.

8. Change made to reflect the Specification. ITS 1.3 does not state that Actions with an "immediate" Completion Time must be performed without interruption.

9. Changes are made to reflect changes made to the Specification.

0 10. Changes made to be consistent with similar words in ITS 3.9.4 Bases. The proposed words clearly define that the standby DHR loop is not required to be in the DHR mode to be considered OPERABLE since the DHR System is a manually operated and controlled system.

0 Davis-Besse Page 1 of 1 Attachment 1, Volume 14, Rev. 1, Page 94 of 142

Attachment 1, Volume 14, Rev. 1, Page 95 of 142 Specific No Significant Hazards Considerations (NSHCs)

Attachment 1, Volume 14, Rev. 1, Page 95 of 142

Attachment 1, Volume 14, Rev. 1, Page 96.of 142 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.9.5, DHR AND COOLANT CIRCULATION " LOW WATER LEVEL There are no specific NSHC discussions for this Specification.

Davis-Besse Page 1 of 1 Attachment 1, Volume 14, Rev. 1, Page 96 of 142

, Volume 14, Rev. 1, Page 97 of 142 ATTACHMENT 6 ITS 3.9i6, REFUELING CANAL WATER LEVEL , Volume 14, Rev. 1, Page 97 of 142

Attachment 1, Volume 14, Rev. 1, Page 98 of 142

Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) 0 Attachment 1, Volume 14, Rev. 1, Page 98 of 142

Attachment 1, Volume 14, Rev. 1, Page 99 of 142 ITS 3.9.6 ITS REFUELING OPERATIONS WATER LEVEL - REACTOR VESSEL LIMITING CONDITION FOR OPERATION LCO 3.9.6 3.9.10 As a minimum. 23 feet of water shall be maintained over the top of irradiat flange d fuel asspblies sea 0ed within the reactor pressure vessel.

irradiated APPLICABILITY: During movement offue assemblies Or co r rods.5 A02 within Ithe re'ac~r pressure ves i1 " DiE ,M ACTION : " ' containment M ACTION A With the requirements of the above specification not satisfied, suspend irradiated all operation involving movement of fuel assemblies or cdro' roL0 1he provisi ons of 50ec~ifcation 3..* L02 within~he re'a c rajre not applilcabl./

5ressure :v~eýssel.

" containmentl 0 SURVEILLANCE SURVEILLANCE REQUIRtMENTS SR 3.9.6.1 4.9.10t The water level shall be determined to be at least its minimum L03 required depth 6within 2 hourr to the start af and at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months during movement of fuel assemblieslo r rod sl with in the reactor pressure vessel. .0 DAVIS-BESSE, UNIT 1 3/4 9-10 Page 1 of I Attachment 1, Volume 14, Rev. 1, Page 99 of 142

Attachment 1, Volume 14, Rev. 1, Page 100 of 142 DISCUSSION OF CHANGES ITS 3.9.6, REFUELING CANAL WATER LEVEL ADMINISTRATIVE CHANGES A01 In the conversion of the Davis-Besse 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-1430, Rev. 3.1, "Standard Technical Specifications-Babcock and Wilcox Plants" (ISTS).

These changes are designated as administrative changes and are acceptable because they do not result in technical changes to the CTS.

A02 CTS 3.9.10 is applicable during movement of fuel assemblies or control rods within the reactor pressure vessel while in MODE 6. 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 L01. The change from within "the reactor pressure vessel" to within "containment" is discussed in DOC M02. The change eliminating control rods is discussed in DOC L02.

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.

A03 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 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 M01 CTS 3.9.10 requires a minimum of 23 feet of water be maintained over the top of irradiated fuel assemblies seated within the reactor pressure vessel. ITS 3.9.6 requires 23 feet of water be maintained above the top of the reactor vessel flange. This changes the CTS by increasing the amount of water that must be in the refueling canal during fuel movement.

Refueling canal water level is required to ensure the consequences of a design basis refuel accident remain within the bounds of the radiological dose calculations. Since the fuel handling accident could occur anywhere in the refueling canal, the water level in the reactor vessel and refueling canal must be at least 23 feet above the top of the reactor vessel flange. Therefore, the increased water level requirement is acceptable. This change is also being made for consistency with the requirements of NUREG-1430, Rev. 3.1. This Davis-Besse Page 1 of 4 Attachment 1, Volume 14, Rev. 1, Page 100 of 142

Attachment 1, Volume 14, Rev. 1, Page 101 of 142 DISCUSSION OF CHANGES ITS 3.9.6, REFUELING CANAL WATER LEVEL change is designated as more restrictive because it adds new requirements to the CTS.

M02 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 "reactor pressure vessel." The ITS 3.9.6 Applicability is 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 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 L01. The change eliminating MODE 6 is discussed in DOC A02. The change eliminating control rods is discussed in DOC L02.

The purpose of CTS 3.9.10 is to ensure the water level is greater than or equal to that assumed in the fuel handling accident analysis. This change is acceptable because a fuel handling accident could occur not just within the reactor pressure vessel, but also within the containment. For example, an irradiated fuel assembly could be dropped in the refueling canal or onto the reactor vessel flange, not over the reactor vessel. While this location is not the drop location assumed in the fuel handling accident, it is consistent with the reason for the water level change discussed is DOC M01. 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 L01 (Category 2 - Relaxation of Applicability) CTS 3.9.10 states that at least 23 feet of water shall be maintained over the top of irradiated fuel assemblies seated within the reactor pressure vessel 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 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 canal 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 by restricting the Applicability and ACTIONS from movement of any "fuel.

assemblies" within the reactor pressure vessel to movement of "irradiated fuel Davis-Besse Page 2 of 4 Attachment 1, Volume 14, Rev. 1, Page 101 of 142

Attachment 1, Volume 14, Rev. 1, Page 102 of 142 DISCUSSION OF CHANGES ITS 3.9.6, REFUELING CANAL WATER LEVEL assemblies" within containment. The change eliminating MODE 6 is discussed in DOC A02. The change from within "the reactor pressure vessel" to within "containment" is discussed in DOC M02. The change eliminating control rods is discussed in DOC L02.

The purpose of CTS 3.9.10 is to ensure the 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.

L02 (Category 2 - Relaxation of Applicability) CTS 3.9.10 states that at least 23 feet of water shall be maintained over the top of irradiated fuel assemblies seated within the reactor pressure vessel 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 movement of fuel assemblies or control rods within the pressure vessel if the water level requirement is not met. CTS 4.9.10 requires a determination of the water level during the movement of fuel assemblies or control rods. ITS 3.9.6 states the refueling canal water level shall be maintained

> 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 are applicable during control rod movement. The change to "irradiated fuel assemblies" from "fuel assemblies" is discussed in DOC L01. The change eliminating MODE 6 is discussed in DOC A02. The change from within "the reactor pressure vessel" to within "containment" is discussed in DOC M02.

The purpose of CTS 3.9.10 is to ensure the 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.

L03 (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 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months prior to the start of" and at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months thereafter during movement of fuel assemblies or control rods within the reactor pressure vessel. ITS SR 3.9.6.1 requires verification that the refueling canal water level is within limit every 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . This changes the CTS by reducing the Frequency for verifying water level from 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months before entering the Applicability of the LCO to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months before entering the Applicability of the LCO.

Davis-Besse Page 3 of 4

,Attachment 1, Volume 14, Rev. 1, Page 102 of 142

Attachment 1, Volume 14, Rev. 1, Page 103 of 142 DISCUSSION OF CHANGES ITS 3.9.6, REFUELING CANAL WATER LEVEL The purpose of CTS 4.9.10 is to ensure that the 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 provides an acceptable level of equipment reliability. The Frequency of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 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).

Furthermore, ITS SR 3.0.4 requires the Surveillance to be met within the specified Frequency prior to entering the Applicability of the LCO. Thus, ITS SR 3.9.6.1 will be performed within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months prior to movement of irradiated fuel assemblies within containment. Therefore, changing the Frequency from 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months before moving fuel assemblies to within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 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.

0 Davis-Besse Page 4 of 4 Attachment 1, Volume 14, Rev. 1, Page 103 of 142

Attachment 1, Volume 14, Rev. 1, Page 104 of 142 Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs)-

0 Attachment 1, Volume 14, Rev. 1, Page 104 of 142

Attachment 1, Volume 14, Rev. 1, Page 105 of 142 Refueling Canal Water Level CTS 319.6 39.: REFUELING OPERATIONS 3.9.6 Refueling Canal Water Level 3.9.10 LCO 3.9.6 Refueling ca nal water Ileve yl shalilbe maintaine d .>!23.It above tihetop of, the reactor vessel flange.

APPLICABILITY: During movement of irradiated fuelassemblies within containment.

ACTIONS

'CONDITION REQUIRED ACTION COMPLETION'TIMEý Action A. Refueling cwater level not within limit, A1 Suspend movement of irradiated fuel 'assemblies Immediately 0

within containmenLt.

SURVEILLANCE REQUIREMENTS ,_,"

SURVEILLANCE FREQUENCY 4.9.10 SR 3.9:6.1 Verify refueling canal water level is Ž.23 ft above the 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months top of reactor vessel flange.,

BWOG STS 3.96-1 Rev. 3.0, 03/31/04 0

Attachment 1, Volume 14, Rev. 1, Page 105 of 142

Attachment 1, Volume 14, Rev. 1, Page 106 of 142 JUSTIFICATION FOR DEVIATIONS ITS 3.9.6, REFUELING CANAL WATER LEVEL

1. Changed to be consistent with the LCO statement.

Davis-Besse Page 1 of 1 Attachment 1, Volume 14, Rev. 1, Page 106 of 142

Attachment 1, Volume 14, Rev. 1, Page 107 of 142 Improved Standard Technical Specifications (ISTS) Bases Markup and Justification for Deviations (JFDs)

Attachment 1, Volume 14, Rev. 1, Page 107 of 142

Attachment 1, Volume 14, Rev. 1, Page 108 of 142 Refueling Canal Water Level B:3A9.6 B 3.9 REFUELING OPERATIONS B 3.9.6 Refueling Canal Water Level BASES BACKGROUND The movement of irradiated: fuel assemblies within containment requires a minimum water level.f 23 ft above the top of the:reactor vesefange.,reactor fg vessel )

During refueling, this maintains sufficient water level in the 1contdinmenT, the refueling canal, the fuel transfer canal,lthe ref ing cavity and the 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 within 10 CFR 100 limits; as provided by the guidance of Reference 3.

APPLICABLE During movement of irradiated fuel assemblies, the water level in the SAFETY ANALYSES refueling canal ui cavit is an initial condition design parameter in the analysis of the fuel handling accident in containment 0

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 c Reference 2. With a minimum water level of 23 ft, and a minimum decay L -7: 1 hours1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months prior to fuel handling, the analysis and test programs demonstrate that the iodine release due to a postulated fuel handling 0

accident is adequately captured by the water, and offsite doses are maintained within allowable limits (Ref. 3).

Refueling canal water level satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii).

LCO L A minimum refueling water level of 23 ft above the reactor vessel flange is required to ensure that the radiological consequences of a 0

postulated fuel handling accident inside containment are within acceptable limits as provided by 10 CFR 100.

BWOG STS B 3.9.6-1 Rev. 3.0, 03/31/04 0

Attachment 1, Volume 14, Rev. 1, Page 108 of 142

Attachment 1, Volume 14, Rev. 1, Page 109 of 142 Refueling Canal Water Level B3.9.6 BASES APPLICABILITY LCO'3.926 is applicable when moving irradiated fuel assemblies within the containment. The LCO minimizes the possibility of a fuel handling accident in containment thatis beyond the assumptions of the safety analysis. If irradiated fuel is not present in conta inment, therecan be no significant radioactivity release as a result of..a postulated fuel handling accident. Requirements for fuel handling accidents in the spent fuel pool are covered by LCO 3:7.14)Fuel t ]Pool Water. Level," 0 ACTIONS A-1 With a water level of.< 23 ft above the top.of the reactor vessel flange, all operations involving movement of irradiated. fuel assemblies shall be:

suspended immediately to ensure that a fuel handling accident cannot occur.

The suspension of fuel movement shall not preclude comipletion of movement of a component to a safe position.

SURVEILLANCE SR 39.6.1 REQUIREMENTS Verification of a minimum water'level of 23Aft above the top of the reactor vessel flange ensures that the design basis for the postulated fuel handling accidentanalysisduring refueling operations is met. Water'at:

the requiredlevel above the topof.the reactor vessel flange limits the.

consequences of damaged fuel rods that are postulated to result from a postulated fuel handling accident inside containment (Ref, 2).

The Frequency of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months is based on engineering judgment and is considered adequate in view of the..large.v olume: of water -andthe normal procedural controls of valve positions, .which make significant unplanned level changes unlikely..

REFERENCES 1. Regulatory Guide 1.25i March 23, 1972.

2t1-'A FSAR Section 11-4.7f 00

3. 10CFR 100.10.

BWOG STS B.3.9.6-2 Rev. 3.0, 03/31104 0

Attachment 1, Volume 14, Rev. 1, Page 109 of 142

Attachment 1, Volume 14, Rev. 1, Page 110 of 142 JUSTIFICATION FOR DEVIATIONS ITS 3.9.6 BASES, REFUELING CANAL WATER LEVEL

1. 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.

2. The brackets have been removed and the proper plant specific information/value has been provided.

3. Change made to reflect changes made to the Specification.

Davis-Besse - Page 1 of 1 Attachment 1, Volume 14, Rev. 1, Page 110 of 142

Attachment 1, Volume 14;' Rev. 1, Page 111 of 142 Specific No Significant Hazards Considerations (NSHCs)

Attachment 1, Volume 14, Rev. 1, Page 111 of 142

Attachment 1, Volume 14, Rev. 1, Page 112 of 142 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.9.6, REFUELING CANAL WATER LEVEL There are no specific NSHC discussions for this Specification.

Davis-Besse Page 1 of 1 Attachment 1, Volume 14, Rev. 1, Page 112 of 142

, Volume 14, Rev. 1, Page 113 of 142 ATTACHMENT 7 RELOCATED/DELETED CURRENT TECHNICAL SPECIFICATIONS , Volume 14, Rev. 1, Page 113 of 142

Attachment 1, Volume 14, Rev. 1, Page 1,14 of 142 CTS 3/4.9.4, CONTAINMENT PENETRATIONS 0

Attachment 1, Volume 14, Rev. 1, Page 114 of 142

Attachment 1, Volume 14, Rev. 1, Page 11.5 of 142 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) 0 Attachment 1, Volume 14, Rev. 1, Page 115 of 142

Attachment 1, Volume 14, Rev. 1, Page 116 of 142 L@ CTS 3/4.9.4 F eLG OPERATIONS AINMENT PENETRATIONS L ING CONDITION FOR OPERATION 3.9.4 Th, containment penetrations shall be in the following status:

a. The iprient hatch cover closed and held in place by a miniimum of four bolts, except the equip. ,thatch maybe open, provided the requirements of Specification 3.9.12 are b; A minimum oone door in each air lock closed, butboth doors.of the cbntainnent personnel air I may be open provided that at least one personnel air lock door is capable of being closed an a designated individualis available irmmediately.outside the personnel air lock to close the or,'and

c. Each penetration providg dirbct access from the containment atmosphere to the atmosphere outsidee ntai nt shall be either:

L. Closed by a manual or auto tic isolation valve, blind flange, or equivalent,.or rge 1 _.

th m ten e c o ntain pu ti on s ig n a l fro m

2. Be capable of being closed recefro

'pon a': h ra d ia ig ipt o thehcontrol roomINbyEanT OPERABLE containment purge 0

a n d exh aust .va lve gas~m bru 0r.

and exhaust yst: im~noble APPLICABILITY: During CORE ALTERATI S or movement of irradiated fuel within the

ACTION:,

a. With the requirements of the above specification not s tisfied, immediately suspendall.

operations involving CORE ALTERATIONS or move, nt of irradiated fuel in the

b. With therequirements of Specification 3.9.4. not satisfied fo the containment purge and exhaust system, close at least one of the isolationvalves for eac of the purge and exhaust penetrations rproviding ho ur. access from the containment atmosp ere tothe outside eithin 0 n e direct o sph e w

.atm Specification 3,0.31aremnot applicable.\ \ ,

c. The provisions of ,

REQUIREMENTS"-"

SURVEILLANCE to ~beither in its.

penetrations. shall be determnined required containment

'4.9A Each of the abovce required condition or capable of being closed by an OPERABLE containment purge d exhaust and at least onceper 7 days during COREe valve, within 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months prior to the start of fuel in the containment, by; :

ALTERATIONS or movement of irradiated

a. Verifying the penetrations are in their required condition, or

b. Verifying that with thezcontainment purge and exhaustsystem in operation, and the containmentpurge and exhaust y oblegas monitor capable~of providinga high.

radiation signal to the control room, tat after initiation of the high radiation signal, the, containnment purge and exhau.tisolation valves can be closed from the control room.

DAVIS-BESSE, UNIT 1 .314 9-4 Amendment N6. 186,202,221,251 Page 1 of 2 Attachment 1, Volume 14, Rev. 1, Page 116 of 142

Attachment 1, Volume 14, Rev. 1, Page 117 of 142 CTS 3/4.9.4 L** INSERT I I

/

Penetration flow pat s) providing direct access from the cont nment atmosphere to the utside atmosphere may be unisolated u ner administrative controls.

Insert Page 3/4 9-4 Page 2 of 2 Attachment 1, Volume 14, Rev. 1, Page 117 of 142

Attachment 1, Volume 14, Rev. 1, Page 118 of 142 DISCUSSION OF CHANGES CTS 3/4.9.4, CONTAINMENT PENETRATIONS ADMINISTRATIVE CHANGES A01 This change to CTS 3.9.4 is provided in the Davis-Besse ITS consistent with License Amendment Request No. 06-0002, submitted to the USNRC for approval in FENOC letter Serial Number 3301, from Mark B. Bezilla (FENOC) to USNRC, dated February 12, 2007. As such, this change is administrative.

MORE RESTRICTIVE CHANGES None RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES None LESS RESTRICTIVE CHANGES L01 (Category I - Relaxation of LCO Requirement) CTS 3.9.4 is applicable during CORE ALTERATIONS and movement of irradiated fuel within the containment.

CTS 3.9.4.a allows the equipment hatch to be open provided the requirements of CTS 3.9.12 (the Spent Fuel Pool Area Emergency Ventilation System) are satisfied and CTS 3.9.4.b allows both airlock doors to be opened under certain provisions. CTS 3.9.4.c provides the requirements for containment penetrations and requires either the penetrations to be isolated by a manual or automatic valve, blind flange, or equivalent, or to be capable of being closed by an OPERABLE containment purge and exhaust valve upon receipt of a high radiation signal. Furthermore, as described in DOC A01, a new Note is proposed to be added to the CTS by another License Amendment request. The proposed Note allows penetration flow paths providing direct access from the containment atmosphere to the outside atmosphere to be unisolated under administrative controls. The ITS does not include this Technical Specification. This changes the CTS by eliminating requirements for Containment Penetrations during CORE ALTERATIONS and when moving irradiated fuel assemblies.

The purpose of the requirements in CTS 3.9.4 is to ensure that a release of fission product radioactivity within containment will be restricted to within regulatory limits during CORE ALTERATIONS and movement of irradiated fuel within the containment. The Applicability of CORE ALTERATIONS is not required since the only accident postulated to occur during CORE ALTERATIONS that is postulated to result in fuel cladding integrity damage is a fuel handling accident. Since the only accident postulated to occur during CORE ALTERATIONS that results in a significant radioactive release is the fuel handling accident, and this Applicability is already specified in the Applicability, it Davis-Besse Page 1 of 2 Attachment 1, Volume 14, Rev. 1, Page 118 of 142

Attachment 1, Volume 14, Rev. 1, Page 119 of 142 DISCUSSION OF CHANGES CTS 3/4.9.4, CONTAINMENT PENETRATIONS is redundant and not necessary. Reducing the Applicability of this Specification to "recently" irradiated fuel assemblies is justified based upon the accident analysis demonstrating that after 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months of radioactive decay, offsite doses resulting from a fuel handling accident remain below the Standard Review Plan limits (well within 10 CFR 100). Once the Applicability has been reduced to require this Specification only during "recently" irradiated fuel assemblies, it is not necessary to maintain the Specification at all. CTS 3.9.3 and ITS 3.9.3 specifically prohibit fuel movement prior to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months after a reactor shutdown.

Thus, after 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , fuel handling would be allowed without any need for the Containment Penetration Specification. That is, the requirements of this Specification are not needed to meet any of the criteria of 10 CFR 50.36(d)(2)(ii).

Containment isolation is not assumed in the fuel handling accident inside containment as documented in UFSAR Section 15.4.7 and Table 15.4.7-4a. The fuel handling accident inside containment assumes no fuel movement prior to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months after shutdown; thus, the Containment Closure Specification, CTS 3/4.9.4 is not needed to be maintained in the ITS since ITS 3.9.3 prohibits fuel movement prior to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . This change is designated as less restrictive because the redundant CTS LCO requirements for Containment Closure are not being maintained.

0 Davis-Besse Page 2 of 2 Attachment 1, Volume 14, Rev. 1, Page. 119 of 142

Attachment 1, Volume 14, Rev. 1, Page 120 of 142 Specific No Significant Hazards Considerations (NSHCs)

Attachment 1, Volume 14, Rev. 1, Page 120 of 142

Attachment 1, Volume 14, Rev. 1, Page 121 of 142 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS CTS 3/4.9.4,, CONTAINMENT PENETRATIONS There are no specific NSHC discussions for this Specification, Davis-Besse Page 1 of 1 Attachment 1, .Volume 14, Rev. 1, Page 121 of 142

Attachment 1, Volume 14, Rev. 1, Page 122 of 142 CTS 3/4.9.6, FUEL HANDLING BRIDGE OPERABILITY Attachment 1, Volume 14, Rev. 1, Page 122 of 142

, Volume 14, Rev. 1, Page 123 of 142 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) , Volume 14, Rev. 1, Page 123 of 142

Attachment 1, Volume 14, Rev. 1, Page 124 of 142 CTS 3/4.9.6 REFUELING OPE___ATIONS

FUEL HANDLING BRIDGE OPERABILITY

ýiLIMTN CONDITION FOR.OPERAT4O*

3.9.6 The control rod hoist a d fuel assembly hoist of the fuel ha d*ing b..dge shall be used for movem nt of control rods or fuel assemblie and shall be OPERABLE with:

a. The control rod hoi t having:

I. A minimum cap city of 3000 pounds, and

2. An overload utoff limit of <_ 2650 pounds.

b. The fuel assembly hoist having:

1. A minimum c pacity of 3000 pounds, and

.2. An overloa cutoff limit of < 2700 pounds.

APPLICABILITY: Durlng'm vement of control rods or fuel assemb ies within the reactor pressure ves el..

ACTION:

With the requirements or control rod hoist and/or fuel ass ly hoist OPERABILITY not satisf ed. suspend use of any inoperable con rol rod hoist and/or fuel assembly ist from operations involving the mov ent of control rods or fuel ssemblies within the reactor pressure essel. The provisions of Specifi ation 3.0.3 are not applicable.

SURVEILLANCE REQJIR ENTS 4.9.6.1 Each cont. I rod hoist used for movement of contr 1 rods or fuel assemblies within t e reactor pressure vessel shall be d nstrated OPERABLE within 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months p ior to the start of such operations by performing a hoist load test of at least 3000 pounds and demonstrating an automatic load cutoff when the co trol rod hoist load exceeds 2650 pound 4.9.6.2 Each fue assembly hoist used for movement of c ntrol rods or fuel assemblies within the reactor pressure vessel shall be d onstrated OPERABLE within 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months prior to the start of such operations performing a load test of at east 3000 pounds and demonstrating an utomatic load cutoff when the uel assembly hoist load exceeds 2700 p unds.

DAViS-BESSE, U1IT 1 3/4 9-6 Amn dment No. 135 Page 1 of 1 Attachment 1, Volume 14, Rev. 1, Page 124 of 142

Attachment 1, Volume 14, Rev. 1, Page 125 of 142 DISCUSSION OF CHANGES CTS 3/4.9.6, FUEL HANDLING BRIDGE OPERABILITY ADMINISTRATIVE CHANGES None MORE RESTRICTIVE CHANGES None RELOCATED SPECIFICATIONS R01 CTS 3.9.6 states that the control rod hoist and fuel assembly hoist of the fuel handling bridge shall be used for movement of control rods or fuel assemblies and shall be OPERABLE with:

a. The control rod hoist having:

1. A minimum capacity of 3000 pounds, and

2. An overload cutoff limits of < 2650 pounds.

b. The fuel assembly hoist having:

1. A maximum capacity of 3000 pounds, and

2. An overload cutoff limit of < 2700 pounds.

OPERABILITY of the fuel handling bridge hoists ensures that the equipment used to handle fuel within the reactor pressure vessel functions as designed and that the equipment has sufficient load capacity for handling fuel assemblies and/or control rod assemblies. Although the interlocks designed to provide the above capabilities can prevent damage to the refueling equipment and fuel assemblies, they are not assumed to function to mitigate the consequences of a design basis accident. This specification does not meet the criteria for retention in the ITS; therefore, it is not included in the ITS. This changes the CTS by relocating this Specification to the Technical Requirements Manual (TRM).

This change is acceptable because CTS 3.9.6 does not meet the 10 CFR 50.36(c)(2)(ii) criteria for inclusion into the ITS.

10 CFR 50.36(c)(2)(ii) Criteria Evaluation:

1. Fuel Handling Bridge OPERABILITY is not used for, nor capable of, detecting a significant abnormal degradation of the reactor coolant pressure boundary prior to a design basis accident (DBA).

2. Fuel Handling Bridge OPERABILITY is not a process variable, design feature, or operating restriction that is an initial condition of a DBA or transient.

Davis-Besse Page 1 of 2 Attachment 1, Volume 14, Rev. 1, Page 125 of 142

Attachment 1, Volume 14, Rev. 1, Page 126 of 142 DISCUSSION OF CHANGES CTS 3/4.9.6, FUEL HANDLING BRIDGE OPERABILITY

3. Fuel Handling Bridge OPERABILITY is not part of a primary success path in the mitigation of a DBA or transient.

4. As discussed in B&W Owners Group Technical Report 47-1170689-00 (Appendix A pages A-89 and A-90), Fuel Handling Bridge OPERABILITY was found to be non-significant risk contributor to core damage frequency and offsite releases. Davis-Besse has reviewed this evaluation, considers it applicable to Davis-Besse Nuclear Power Station, and concurs with the assessment.

Since the 10 CFR 50.36(c)(2)(ii) criteria have not been met, the Fuel Handling Bridge Operability LCO and associated Surveillance may be relocated out of the Technical Specifications. The Fuel Handling Bridge Operability will be relocated to the TRM. The TRM is currently incorporated by reference into the UFSAR, thus any changes to the TRM are made under 10 CFR 50.59, which ensures changes are properly evaluated. This change is designated as relocation because the LCO did not meet the criteria in 10 CFR 50.36(c)(2)(ii) has been relocated to the TRM.

REMOVED DETAIL CHANGES None LESS RESTRICTIVE CHANGES None Davis-Besse Page 2 of 2 Attachment 1, Volume 14, Rev. 1, Page 126 of 142

Attachment 1, Volume 14, Rev. 1, Page 127 of 142 Specific No Significant Hazards Considerations (NSHCs)

Attachment 1, Volume 14, Rev. 1, Page 127 of 142

Attachment 1, Volume 14, Rev. 1, Page 128 of 142 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS CTS 3/4.9.6, FUEL HANDLING BRIDGE OPERABILITY There are no specific NSHC discussions for this Specification.

Davis-Besse Page 1 of 1 Attachment 1, Volume 14, Rev. 1, Page 128 of 142

Attachment 1, Volume 14, Rev. 1, Page 129 of 142 ATTACHMENT 8 IMPROVED STANDARD TECHNICAL SPECIFICATIONS (ISTS)

NOT ADOPTED IN THE DAVIS-BESSE ITS Attachment 1, Volume 14, Rev. 1, Page 129 of 142

, Volume 14, Rev. 1, Page 130 of 142 ISTS 3.9.3, CONTAINMENT PENETRATIONS , Volume 14, Rev. 1, Page 130 of 142

Attachment 1, Volume 14, Rev. 1, Page 131 of 142 W Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs) 0 Attachment 1,,Volume 14, Rev. 1, Page 131 of 142

Attachment 1, Volume 14, Rev. 1, Page 132 of 142 0

Containment Penetrations 3¸ 3.9 REFUELING OPERATIONS 3.9.3 Containment Penetrations LC 319.3 , The containment penetrations shall be in the following status:

a. The equipment hatch closed and held in place by four bolts,.

b. One door in each air lock is[capable of being] closed, and c: Each penetration providing direct access from the containment atmosphere to the outside atmosphere either:

1. Closed by amanual or automatic isolation valve, blind flange, orequivalent or Capable of being closed by an OPERABLE Containment Purgeand Exhaust Isolation System.

- *- - -NOTE- ----- -.--------- ------ I-------

Penetration f w path(s) providing direct access from the containment atmosphere to e.eoutside%.atmosphere. may be unisolated under administrative co rols.

IF APPLICABILITY: During movement of fre-cen] irradiated fuel'assemble-wii ithin containment.

ACTIONS, CONDITION REQUIRED ACTI0 COMPLETION TIME A, One or more A.1 Suspend movement of Immediately.

containment (recently] irradiated fuel penetrations not in assemblies within required status. containment.

BVVOG STS 3.9.3-1 Rev. 3.0,,03/31104 II Attachment 1, Volume 14, Rev. 1, Page 132 of 142

Attachment 1, Volume 14, Rev. 1, Page 133 of 142 Containment Penetrations SURVEILLANCE REQUIREMENTS

SURVEILLANCE ' FREQUENCY "Nk S 3.9.3,1 Verify each required containment penetration is in 7 days

~the requited status.

SR 3.933. .. . . NOTE ---.............---------------

Not required to be met for containmentL purge and exhaust valve(s) in penetrations closedto comply With LCO a..aci.

Ven' each-reqbired containment purge and. [18] months exha 'tvalve actuates to the isolation position on an Bctu or simulated actuation signal.

BWVOGSTS1932R.3.,0/14 Attachment 1, Volume 14, Rev. 1, Page 133 of 142

Attachment 1, Volume 14, Rev. 1, Page 134 of 142 JUSTIFICATION FOR DEVIATIONS ISTS 3.9.3, CONTAINMENT PENETRATIONS

1. This ISTS Specification is not being maintained as described in the Discussion of Changes for CTS 3/4.9.4.

Davis-Besse Page 1 of 1 Attachment 1, Volume 14, Rev. 1, Page 134 of 142

Attachment 1, Volume 14, Rev. 1, Page. 135 of 142 Improved Standard Techni1call Specifications (ISTS) Bases Markup and Justification for Deviations (JFDs)

Attachment 1, Volume 14, Rev. 1, Page 135 of 142

Attachment 1, Volume 14, Rev. 1, Page 136 of 142 0

Containment Penetrations B 3.9.3 B 3.9 REFUELING OPERATIONS Bý3.9.3 Containrent Penetrations.

BA. S, BACK OUND During movernent of [recently] irradiated fuel assemblies within containment a release of fission product radioactivity.within containmtent:

will be restricted from escaping to the'environment when the LCO requirements are met. In MODES 1, 2, 3, and 4, thisis accomplished by maintaining containment OPERABLE as described in LCO 3.611, "Containment.' InMODE 6, the potential for containment pressurization as a result of an accident isnot likely; therefore, requirements to isolate he,containment from theoutside atmosphere can 'be less stringent. The 0 requirements'are referred to as 'containment closure' rather than "co ainment OPERABILITY." Containment closuremeans that all poten'al escape paths are closed or capable of being closed. Sirice there is

potential for containment pressurization, the Appendix J leakage te'ria and tests are not required.

The contain nt serves to contain fission product radioactivity that may be released'fro the. reactor'core-following an accident, such'that offsite radiation exposur s are maintained well within the requirements!of 10.CFR 100. Addit nally, the containment provides radiation shielding from the fission produts that may be present in the containment atmosphere following a ident conditions.

Thecontainment equipmen atch, which is part-of the~containment pressure boundary,.proVides, means for moving large equipment and components into and out of con inment. During movement of [recently]

irradiated fuel assemblies within rntainment, the equipment hatch must be held in place by at least four bol . Good engineering practice dictates that~the bolts requiredby this LCO be pproximately equally spaced.

The containment air locks, which are also artlof the containment pressure boundary, provide a means for pe onnel access during MODES 1, 2, 3, and 4 unit operation in accor nce with LCO 3.6,2, "ContainmentAir Locks." Each air lock has a d r at both ends. The doors are normally interlocked to prevent simulta ous opening when containment OPERABILITY is required. During per 0ds of unit shutdown when containment closure is not required, the door in rlock mechanism may be disabled, allowing both doors of an airlock to re in open for extended periods when frequent containment entry is nec sary. During movementof [recently] irradiated fuel assemblies within con inme nt, containamentclosurelis required; therefore, the door interlock echanism may remain disabled, but one air lock door must always remain apable of being] closed.

BWOG STS B 3.9.3-1 Rev. 3.0, 03131104 Attachment 1, Volume 14, Rev. 1, Page 136 of 142

Attachment 1, Volume 14, Rev. 1, Page 137 of 142 0

0 Containment Penetrations B.3.9.3.

BASES BACKGROUND (continued)

The requirementson.containment penetration closure ensure that a release of fission product radioactivity within containment will be restricted to within regulatory limits.

The Contairiment:Purge and Exhaust System includes two subsystems.

The normal subsystem includes a (42] inch purge penetration and, a (42] inch exhaust penetration. The secornd subsystem, or minipurge system, includes an.[8] inchpurge penetration and an [8] inch exhaust penetration. During MODES 1 2,.3, and 4,.the two valves in each of the normal purge and exhaust penetrations are secured.in the closed sition'; The twolValves:in each of~the two minipurge penetrations can, b., opened intermittently but are closed automatically by the Engineered Saf Feature.Actuation System (ESFAS). Neither of the subsystems is subje to a;Specification in MODE 5.

InMODE' large'air exchangers are necessary to conduct refueling operations'. he normal [42] inch purge system is used for this purpose, and all four vayaes areclosed on a reactorbuilding (RB) high radiation signalinaccord .ce with LCO 3.3..15.' Reactor Building (RB) Purge Isolation - High R iation.

Theother containmen enetrations that provide direct access from containment atmrospher to outsideatmosphere must be isolated on at:;

least o sie. dne Isolation y be achieved by an OPERABLE automatic isolation valve or by'a manu 1,isolation valve, blind flange, or equivalent.

EquivaleInt'isoatin methodsr ust be approved and may-include use of a material that can provide:a.temp rary, atmospheric pressure ventilation barrier for the other containment.p netrations during fuel movements

[involAvig handling recentil irradiate fuel] (Ref. 1).

APPLICABLE During movement of [recently] irradiate uel assemblies within SAFETY, .containment,. the. most severe radiological onsequences result from a ANALYSES fuel handling accident [involving handling re ntly irradiated fuel]. The fuel handling accident is a postulated event th involvesdamage to irradiatedfuel.(Ref. 2)ý Fuel handling accidents, nalyzed in Ref. 3,.

include dropping a single irradiated fuel assembly nd handling tool or a heavy objectonto. other irradiated fuel assemblies. e requirements of LCO3.9.6, "Refueling Canal WaterLevel," in conjunct n with minimum decay time of (100] hours prior to [irradiated] fuel move nt [with containment closure capability or a minimum decay time o X] days without containment closure capability], ensures that the rele se of fission product radioactivity subsequent to afuel handling accident re,Its in doses that are within the requirements specified in 10 CFR 100. he acceptance limits for offsite radiation exposure are contained in R 2.

Containmentpenetrationssatisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii).

BVWOG STS B3.9:3-2 Rev. 3:0, 03/311104 Attachment 1, Volume 14, Rev. 1, Page 137 of 142

Attachment 1, Volume 14, Rev. 1, Page 138 of 142 0

Containmernt Penetrations B3.9.3 BASES LCO ---------

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

The allowance to have containment personnel airlock doors open and penetration flow paths with direct access from the containment atmosphere to the outside atmosphere to be unisolated during fuel movement and CORE ALTERATIONS is based on (1) confirma-tory dose

calculatiohs of a fuel handling accident as approvedby the NRC staff

'Which indicate acceptable'radiological consequences and (2) commitments from the licensee to implement acceptable,

.administrative procedures thatensure in the event of a. refueling accident (even though the containment fission product control function is not required t9 meet acceptable dose consequences) that the open airlock can'and willfbe: promptly closed following containment evacuation and:

that:the open penetration(s) can. and will be promptly closed. The timeto

.osesuch.penetrations or combination of.penetrations shall be included in- e, onfirmratory. dose calculations.

This. Lc. limits the consequences of.a fuel handling accident [involving handling icently irradiated fuel] in containment by limiting the potential escapepat for fission product radioactivity.from containment. The LCO requiresany' netration providing direct access from the containment

.a mospere to eouts ide atmosphere to be closed except for the OPERABLE con tnment purge"and'exhaust penetrations [and the containment perso el airlocks]ý For the OPERABLE containment purge and exhaust Openetra ns, this LCO ensures that these penetrations are isolable by the RB pur isolation signal. The OPERABILITY requirements forhis:LC ensure that the automatic purge and exhaust valve closure ties:'specifi in the FSAR can be achieved and therefore meet the assumptios.used i the safetyanalysis to ensure releases through the valves are termina 'd such that radiologica Idoses are within

.the acceptance limit.

The LCOis modified by a Note allovg penetration flow paths with direct access from the containment atmosph re to the outside atmosphere to be unisolated under administrative controls. Administrative controls ensure th~at 1) ppropriate~personnel are awareo he, open' status of the-penetration flow path during COREALTER IONS or movement of ilthin containment ,nd 2) specified individuals are designated and readily available to isolate t flow path in the.event.

of a fuel handling, accident.

The containment personnel airlock doors may be ope during movement of (recently] irradiated*fuel in the containment provided at one door is capable of being closed in the event of a fuel handling a dent. Should a fuel handling accident occur inside containment, one perso el airlock door~will be closed following an evacuation of containment, BWOG:STS B 3.93-3 Rev. 3.0, 03/3 4 Attachment 1, Volume 14, Rev. 1, Page 138 of 142

'Attachment 1, Volume 14, Rev. 1, Page 139 of 142 0

Containment Penetrations B 3.9.3 BASES APPLICABILITY The containment penetration'requirementseare applicable during movement of [recently] irradiated fuel assemblies withincontainrment because this is when there is a potential fortthe limiting fuel,'handling accident. In MODES 1.2,; 3, and 4;..containment penetration requirements are addressed by LCO"3.1. In MODES 5 and 6, when.

movement of irradiated fuel as;sembliesvwithin .containment is not being conducted, the potential for a fuel handling accident does not exist.

[Additionally, dueto radioactive decay, a fuel handling accident involving handling recentlyirradiated fuel (iLe., fuel thathas occupied part of a.

critical reactor core within the previous VXdays) will result in doses that are well within the guideline values specified in 10-CFR, 100 even without cohtain'ent closure capability.] Therefore under these conditions no requirements are placed on containment penetration status.

---- R EVIEWER'SNOTE -------------

Th ddition of the term "recently" associated with handling irradiated fuel inrall f the containmeit function Technical Specification iequirements is only ap licable to those licensees-who. have demonstrated by analysis

ýthatafte ufficient radioactive decay has occurred, off-site doses:

resulting fr m a fuel handling accident remain below the Standard, Review Plan limits ( IIwithin 1 :CFR 100).

Additionally; lice sees'adding the term "recently"-must make the following commitment whic is consistent with NUMARC 93-01, Revision 4, Section 113.6.5 "S ety Assessment for: Removal of.Equipment from Service During Shutd wn Conditions," subheading "Containment -

Primary (PVVR)iSeco:n (BVR)."

"The following guidelines a included in the assessment of systems removed from service during ovementof irradiated fuel:

SDuring fuel handling/core a erations, ventilation system and.

radiation monitor availability (, defined in NUMARC 91 -06) should be assessed, with respect to filt .rtionand monitoring of releases from the fuel., Followin shutdo r activity in the fuel decays away fairly rapidly. The basis of the Techical Specification operability amendmentistýt reduction in doses ue to such decay. The goal of mystem and radi tion monitor availability is.to reduce dosesev~en further below that pro. ided by the natural decay, A single normal or contingency method to p mPtly close primary or secondary containmentpenetrations should b developed. Such

.prompt methods'need not cornpltely block the netration-or be capable of resisting pressure.

BVVOG STS B.3.0~3-4 Rev. 3.0. 0'&31/104 Attachment 1, Volume 14, Rev. 1, Page 139 of 142

Attachment 1, Volume 14, Rev. 1, Page 140 of 142 .

0

containment Penetrations B 3.9.3 BASES APPLICABILITY (continued)

The purpose:of the "prompt.methods" mentioned above areto -enable ventilation systems to draw the release from a postulated fuel handling accident'in the proper direction such thatit can be treated and mon itored."

\ ---------

ACTIO AI W*ith the containmentfequipment hatch, air locks, or any containment penetration that provides directaccess-from the containment atmosphere to the outside atmosphere not in the required status, including.the Containment Purge and Exhaust Isolation System not capable of.

automatic actuation-when the purge and exhaust valves are open. the nit must be placed ina condition in which the isolation function is not ne ded. This is accomplished by immediately suspending movement of

[re ntly],irradiated fuel assemblies within-containment. Performance of these ctions shall not preclude moving a component to a safe position..

SURVEILLANCE SR 3.9.3.

REQUIREMENTS ii This Surveilla demonstrates that each of the containment penetrations required tobe i'ts closed position is in that position. The Surveillance on the openmpurge and exhaust-valves will demonstrate that the valves are notblocked fro closing, Alsothe Surveillance will demonstrate that each valve operator h 'motive power, which will ensure each valve iSr capable of being close y an OPERABLE.automatic RB purge isolation signal.

The Surveillance i1sperforme very 7 days during movement of.[recently],

irradiated fuel asemblies'within hecontainrment. The Surveillance interval is selected to be commen rate with the normal duration of time to complete fuel handling operation A. surveillance before the start of refueling operations willpr~ovide two o hree surveillance verifications during the applicable period for this LC As such, this Surveillance ensures that a po ulated fuel:handling accident (involving handling recently irradiate .*uel].that: releases fission.

product radioactivity within thecontainment will . t result in a release of.

significant fission product radioactivity to the envir ment in excess of those recommended by'Standard Review Plan Sect n 15714.(Ref. 3).

BWVOG STS B.3.9:3-5 Rev: 3.0, 03/3 04 11 Attachment 1, Volume 14, Rev. 1, Page 140 of 142

Attachment 1, Volume 14, Rev. 1, Page 141 of 142 Containmehnt Penetrations BR3.9.3, BASES SURVEILLANCE REQUIREMENTS (continued)

SR 3.9.3.2 This Surveillance demonstrates thateach containment purgepand exhaust valve actuates to its isolation position on manual initiation or'on an actual or simulated high radiation:signal. The 18 month Frequehcy: maintairs consistency with othersimilarESFAS instrumentation and valve testing requirements. In LCO 3.3.15, "RB Purge Isolation -:High Radiation," the isolation instrumentation requires a CHANNEL CHECK every.12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and a CHAN NEL FU NCTIONAL TEST every 92 days to ensure the.

channel OPERABILITY during:refueling operations. Every 18 months'a CHANNEL CALIBRATION is'performed.: The system actuation response

time is demonstrated every 18 months, dudng refueling, on.a, STAGGERED TEST BASIS. SR 3.6.3.5 demonstrates that theisoliation*

im:e 0f each valve is in accordance with the Inservice Testing Program r' uirements. TheseSurveillances performed during MODE 6 will enisure tha 'the valves are capable of closing after.a' postulated fuel handling accid nt [involving handling recently irradiated fuel] to limita. release of fission roduct radioactivity from the containment.

The SR is- odified by a Note:stating that this Surveillance is not required to be met for. alves in isolated penetrations. The LCO provides the option toctose

netrations in lieu.of requiring automatic actuation

.capability, REFERENCES t. GPU NuclearSa ty. Evaluation SE-0002000-001, Rev. 0, May 20.1988,

2. FSAR, Section[].

3. NUREG-0800, Section 15. .4, Rev. 1, July 1981.

BWVOG STS B 319.3-6 Rev 30,0 104 Attachment 1, Volume 14, Rev. 1, Page 141 of 142

Attachment 1, Volume 14, Rev. 1, Page 142 of 142 JUSTIFICATION FOR DEVIATIONS.

ISTS 3.9.3 BASES, CONTAINMENT PENETRATIONS

1. This ISTS Specification Bases is not being maintained as described in the Discussion of Changes for CTS 3/4.9.4.

Davis-Besse Page 1 of 1 Attachment 1, Volume 14, Rev. 1, Page 142 of 142

v t e Letter Sequence Request
TAC:MD6398, Increase Flexibility in Mode Restraints, Addition of LCO 3.0.8, Inoperability of Snubbers, Deletion of E BAR Definition and Revision to RCS Specific Activity Tech Spec, Changes to Reflect Revision of 10 CFR 50.55A, Limit Inservice Testing Program SR 3.0.2 Application to Frequencies of 2 Years or Less, Eliminate Second Completion Times Limiting Time from Discovery of Failure to Meet an LCO, Correct Example 1.4-1, Containment Structural Integrity, TSTF-28, Correct LCO 3.0.6 Bases, Add Restriction for Lowest LOOP Tave During Mode 2 Physics Testing (Approved, Closed)
Initiation Request, Request, Request, Request, Request, Request, Request Acceptance, Acceptance, Acceptance, Acceptance Supplement, Supplement Administration Meeting, Meeting, Meeting Questions 18 June 2008 20 June 2008 Answers 23 July 2008 3 September 2008 Results Approval Other: L-08-161, Copy of Applicable Portions of the NRC and Davis-Besse Improved Technical Specifications Conversion Website - Sections 3.0, 3.1, 3.2, 3.4, and 3.6 and Chapter 4.0, L-08-241, Attachment 1, Applicable Portions of the U.S. Nuclear Regulatory Commission (NRC) and Davis-Besse Nuclear Power Station (DBNPS) Improved. Technical Specification (ITS) Conversion Website, Section 3.5 RAIs, L-08-338, Copy of Improved Technical Specifications (ITS), L-08-378, Implementation of Improved Technical Specification Amendment No. 279, ML073531253, ML081350505, ML082000451, ML082270661, ML082270662, ML082270663, ML082270664, ML082270665, ML082270666, ML082270667, ML082270669, ML082270670, ML082600577, ML082840099, ML082900630, ML082900882, ML082910125
MONTHYEAR2008-02-04 [Table View]

L-08-240, Improved Technical Specification Conversion License Amendment Request, Volume 14, Revision 1, ITS Section 3.9, Refueling Operations

Text

Navigation menu

L-08-240, Improved Technical Specification Conversion License Amendment Request, Volume 14, Revision 1, ITS Section 3.9, Refueling Operations

Text

Navigation menu

Search