Text

Volume 14, Revision 0, Davis-Besse, Unit 1 - Improved Technical Specifications Conversion, ITS Section 3.9 Refueling Operations.
	ML072200481
Person / Time
Site:	Davis Besse
Issue date:	08/03/2007
From:	; FirstEnergy Nuclear Operating Co
To:	; Office of Nuclear Reactor Regulation
References
	Download: ML072200481 (138)
	v • d • e

Attachment 1, Volume 14, Rev. 0, Page 1 of 138 ATTACHMENT 1 VOLUME 14 DAVIS-BESSE IMPROVED TECHNICAL SPECIFICATIONS CONVERSION ITS SECTION 3.9 REFUELING OPERATIONS Revision 0 Attachment 1, Volume 14, Rev. 0, Page 1 of 138

Attachment 1, Volume 14, Rev. 0, Page 2 of 138 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. RelocatedlDeleted Current Technical Specifications Attachment 1, Volume 14, Rev. 0, Page 2 of 138

, Volume 14, Rev. 0, Page 3 of 138 ATTACHMENT 1 ITS 3.9.1, BORON CONCENTRATION , Volume 14, Rev. 0, Page 3 of 138

, Volume 14, Rev. 0, Page 4 of 138 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) , Volume 14, Rev. 0, Page 4 of 138

Attachment 1, Volume 14, Rev. 0, Page 5 of 138 ITS 3.9.1 ITS 3/4.9 REFUELING OPERATIONS BORON CONCENTRATION LIMITING CONDITION FOR OPERATION LCO 3.9.1 3.9.1 The boron concentration of all fill 1 portions of the Reactor Coolant e0 System and the refuelino canal shall be maintained unifonm and sufficient to I for uncerta~hties LA01 L {within the limits of the COLR}

APPLICATION TY: MODE 6. Add proposed Applicability Note ACTION:

ACTION A With the requirements of the above specification not satisfied, Immediately suspend all operations Involving JCOR ALTE INS or ýposltive reactivity changes and initiate and continue borat~on Jof2* 7 9P* of 78175 pm oric L lacidl solution or Its leouivalent until &ye is reduced i~ ;0.95. eL0 provisions ot/.%pec~ttcatlo~n 3_..3 are pr~o.applicable.1 SURVEILLANCE REO*UTREMENTSA0 4..Y 1°*.I le aoove con ition snaJI De ete*erineo pr or To:

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

b. WIthdrawal o any safety or regulating ro in excess of 3 feet from its ful y inserted position within.t e reactor pressure vessel.

SR 3.9.1.1 4.9.1.2 The boron concentration of the reactor pressure vessel and the refueling canal shall be determined lby chemic.ana ysis at least once each 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . I LAO2 DAVIS-BESSE, UNIT 1 3/4 9-1 Amendment No. 43,207 Page 1 of 1 Attachment 1, Volume 14, Rev. 0, Page 5 of 138

Attachment 1, Volume 14, Rev. 0, Page 6 of 138 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 Parameter Limits from the Technical Specification to the Core Operating Limits 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. 0, Page 6 of 138

Attachment 1, Volume 14, Rev. 0, Page 7 of 138 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 as a less restrictive removal of detail change because procedural details for meeting Technical Specification Requirements are being removed from the Technical Specifications.

Davis-Besse Page 2 of 4 Attachment 1, Volume 14, Rev. 0, Page 7 of 138

Attachment 1, Volume 14, Rev. 0, Page 8 of 138 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 (Category4 - 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. 0, Page 8 of 138

Attachment 1, Volume 14, Rev. 0, Page 9 of 138 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 (Category 5 - Deletion of Surveillance Requirement) CTS 4.9.1.1 requires the LCO reactivity condition to be determined prior to removing or unbolting the reactor vessel head, and prior to withdrawal of any 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.

Davis-Besse Page 4 of 4 Attachment 1, Volume 14, Rev. 0, Page 9 of 138

Attachment 1, Volume 14, Rev. 0, Page 10 of 138 Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs)

Attachment 1, Volume 14, Rev. 0, Page 10 of 138

Attachment 1, Volume 14, Rev. 0, Page 11 of 138 CTS Boron Concentration 3.9.1 3.9 REFUELING OPERATIONS 3.9.1 Boron Concentration (RCS) and 3.9.1 LCO 3.9.1 Boron concentrations of the Reactor Coolant Systernothe refueling canal 0 and cavity shall be maintained within the limit specified in the 00 COLR.

APPLICABILITY: MODE 6.

NOTE ----

DOC L01 Only applicable to the refueling canaljy connected to the RCS.

when 0

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

Af Suspend positive reactivity Immediately additions.

AND A,. Initiate action to restore Immediately -77TSTF 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 Attachment 1, Volume 14, Rev. 0, Page 11 of 138

Attachment 1, Volume 14, Rev. 0, Page 12 of 138 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.

Davis-Besse Page 1 of 1 Attachment 1, Volume 14, Rev. 0, Page 12 of 138

Attachment 1, Volume 14, Rev. 0, Page 13 of 138 Improved Standard Technical Specifications (ISTS) Bases Markup and Justification for Deviations (JFDs)

Attachment 1, Volume 14, Rev. 0, Page 13 of 138

Attachment 1, Volume 14, Rev. 0, Page 14 of 138 Boron Concentration B 3.9.1 B 3.9 REFUELING OPERATIONS B 3.9.1 Boron Concentration BASES BACKGROUND The limit on the boron concentrations of the Reactor Coolant System S(RCS) the refueling canal an t re ue i cavity 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 refueling. 0 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 kef < 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. UFSAR, Appendix 3D.1.22 [Makeup and Purification GDC 26 CFR 50 endix , 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. The ChemcZAdditionlSystem serves as the system capable of maintaining the reactor subcritical in cold ci) 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 is lo y removed to form the refn cavit . Therefueling canal is re ue in v ar then flooded with borated water from the borated water storage tank into the open reactor vessel by gravity feeding or by the use ofe 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 n th

,refuelicav mix the added concentrated boric acid with the water in 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 and the re in cavil above the COLR limit. 0 BWOG STS B 3.9.1-1 Rev. 3.0, 03131/04 Attachment 1, Volume 14, Rev. 0, Page 14 of 138

Attachment 1, Volume 14, Rev. 0, Page 15 of 138 Boron Concentration B 3.9.1 BASES APPLICABLE During refueling operations, the reactivity condition of the core is SAFETY consistent with the initial conditions assumed for the boron dilution ANALYSES accident in the accident analysis and is conservative for MODE 6. The boron concentration limit specified in the COLR is based on the core reactivity at the beginning of each fuel cycle (the end of refueling) and 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 kf of the core will remain < 0.95 during the refueling operation. Hence, at least a 5% Ak/k margin of safety is established during refueling.

During refueling, the water volume in the spent fuel pool, the transfer canal, the refueling canal, Ithe refueig 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).

LO an The LCO requires that a minimum boron concentration be maintained in LCO ithe RCS, the refueling canall, and the re e ing cay yywhile 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 to an inadvertent criticality during MODE 6.

APPLICABILITY This LCO is applicable in MODE 6 to ensure that the fuel in the reactor vessel will remain subcritical. The required boron concentration ensures a kejf_ 0.95. Above MODE 6, LCO 3.1.1, "SHUTDOWN MARGIN (SDM)." 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 boron concentration are only applicable to the refueling canal anse

}03 this

~ireuehrcavl when se volumeM[Lateconnected to the RCS. When the refueling canalland the refelinq cavi EIO isolated from the RCS, no potential path for boron dilution exists.

BWOG STS B 3.9.1-2 Rev. 3.0, 03131/04 Attachment 1, Volume 14, Rev. 0, Page 15 of 138

Attachment 1, Volume 14, Rev. 0, Page 16 of 138 Boron Concentration B 3.9.1 BASES ACTIONS A.1 A TT C.471 Continuation of CORE ALT zKATIONS orlpositive reactivity additions (including actions to reduce boron concentration) is contingent upon maintaining the unit in compliance with the LCO. Ifthe boron or concentration of any coolant volume in the RCS*the refueling canalF- 0

ýhe refu is less than its limit, all operations involving F TSTF JALTER ONS or positive reactivity additions must be suspended Y47 immediately.

Suspension of CORE ALTEARATIONS and positive reactivity additions ST*F 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 CORE ALTE ATIONS and 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 restore the boron concentration to its required value as soon as possible. In order to raise the boron concentration as soon as possible, the operator should begin boration with the best source available for unit conditions.

Once actionr ha been initiated, mu-st be continued until the boron concentration is restored. The restoration time depends on the amount of 0

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. 0, Page 16 of 138

Attachment 1, Volume 14, Rev. 0, Page 17 of 138 Boron Concentration B 3.9.1 BASES SURVEILLANCE SR 3.9.1.1 REQUIREMENTS This SR ensures the coolant boron concentration in the RCSp and connected portions of the refueling canal and the refling cavityj 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 canal or the ref eing cav to the 0 RCS, this SR must be met per SR 3.0.4. If any dilution activity has . fl occurred while the canal ý reisconnected from the RCS, this SR ensures the correct boron concentration prior to communication with the RCS.

The SR 3.9.1.1 . . um Frequency of once every 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months is ore a reasonable amount of time to verify the boron concentration of representative 0 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.

REFERENCES 1. I10 CFR *5A *,ff1A. GDC 2E.- UFSAR, Appendix 3D.1.22 0

BWOG STS B 3.9.1-4 Rev. 3.0, 03131/04 Attachment 1, Volume 14, Rev. 0, Page 17 of 138

Attachment 1, Volume 14, Rev. 0, Page 18 of 138 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.

Davis-Besse Page 1 of 1 Attachment 1, Volume 14, Rev. 0, Page 18 of 138

Attachment 1, Volume 14, Rev. 0, Page 19 of 138 Specific No Significant Hazards Considerations (NSHCs)

Attachment 1, Volume 14, Rev. 0, Page 19 of 138

Attachment 1, Volume 14, Rev. 0, Page 20 of 138 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. 0, Page 20 of 138

, Volume 14, Rev. 0, Page 21 of 138 ATTACHMENT 2 ITS 3.9.2, NUCLEAR INSTRUMENTATION , Volume 14, Rev. 0, Page 21 of 138

, Volume 14, Rev. 0, Page 22 of 138 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) , Volume 14, Rev. 0, Page 22 of 138

Attachment 1, Volume 14, Rev. 0, Page 23 of 138 ITS 3.9.2 ITS REFUELING OPERATIONS INSTRUMENTATION

'LIMITING.CONDITION FOR OPERATION LCO 3.9.2 3.9.2 Two source range neutron flux monitors, one from each side of the reactor core, shall be OPERABLE.

APPLICABILITY: MODE 6.

ACTION:

ACTION A a. With only one of the required OPERABLE source range neutron flux monitors, LOS I. Immediately suspend CORE ALTERATIONS and

2. Immediately suspend operations that would cause introduction of coolant into the RCS with boron concentration less than the RCS boron concentration requirement of LCO 3.9.1.

ACTION A, b. With no OPERABLE source range neutron flux monitor, ACTION B I1. Perform AetION a., and + A02

2. Immediately initiate action to restore one source range neutron flux monitor to OPERABLE status, and

3. Once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ver fy that the RCS boron concentration mees the requirement of LCO 3.9.A1, using chmical analysis to determine the boron con/entration of the

[reactor pressure vessI and the refueling al t PromS3...

-- A0e SJRV EUAC EOIEET 4.9.2 As a minimum, two source range neutron flux monitors, one from each side of the reactor core, shall be demonstrated OPERABLE by performance of:

a. Dileted

b. [ eleted 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

d. A CHANNEL CALIBRATION nrior to entry into MOD not performed within the SR 3.9.2.2 la 118 months. Neutron detectors are excluded from CHANNEL CALIBRATION.

DAVIS-BESSE, UNIT 1 3/4 9-2 Amendment No. 172, 269 Page 1 of 1 Attachment 1, Volume 14, Rev. 0, Page 23 of 138

Attachment 1, Volume 14, Rev. 0, Page 24 of 138 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 entered. Therefore, there is no need for the specific statement. This change is designated as administrative because it does not result in technical changes to the CTS.

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. 0, Page 24 of 138

Attachment 1, Volume 14, Rev. 0, Page 25 of 138 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 (Category4 - Relaxation of RequiredAction) 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. 0, Page 25 of 138

Attachment 1, Volume 14, Rev. 0, Page 26 of 138 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.

Davis-Besse Page 3 of 3 Attachment 1, Volume 14, Rev. 0, Page 26 of 138

Attachment 1, Volume 14, Rev. 0, Page 27 of 138 Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs)

Attachment 1, Volume 14, Rev. 0, Page 27 of 138

Attachment 1, Volume 14, Rev. 0, Page 28 of 138 CTS Nuclear Instrumentation 3.9.2 3.9 REFUELING OPERATIONS 3.9.2 Nuclear Instrumentation 3.9.2 LCO 3.9.2 one from each side of the reactor core, Two source range neutron flux monitors1hall be OPERABLE.

0 APPLICABILITY: MODE 6.

ACTIONS CONDITION REQUIRED ACTION [COMPLETION TIME Action a.

Action b A. One Irequiredl source A.1 Sus end C RE Immediately range neutron flux LRexcept the introduction of monitor inoperable. AND I' additions ,coolant

[positive reactivity into the RCS R

A.2 Suspend operations that Immediately would cause introduction of coolant into the RCS with boron concentration less 0D0 TT than required to meet the boron concentration of LCO 3.9.1.

Action b B. Two Irequiredl source B.1 Initiate action to restore one Immediately range neutron flux source range neutron flux monitors 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. 0, Page 28 of 138

Attachment 1, Volume 14, Rev. 0, Page 29 of 138 CTS Nuclear Instrumentation 3.9.2 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

BWVOG STS 3.9.2-2 Rev. 3.0, 03/31/04 Attachment 1, Volume 14, Rev. 0, Page 29 of 138

Attachment 1, Volume 14, Rev. 0, Page 30 of 138 JUSTIFICATION FOR DEVIATIONS ITS 3.9.2, NUCLEAR INSTRUMENTATION

1. ISTS LCO 3.9.2 has been changed to include the statement "one from each side of the reactor core." This is consistent with the current licensing basis. Davis-Besse has four source range neutron flux monitors available to meet the LCO, two from the Reactor Protection System (RPS) source range neutron flux monitors and two from the Post Accident Monitoring (PAM) source range neutron flux monitors. The RPS source range neutron flux monitors are located on each side of the reactor core, approximately 1800 apart. The PAM source range neutron flux monitors are located in the same vicinity as the RPS source range neutron flux monitors. Requiring at least one to be OPERABLE on each side of the core ensures that the neutron flux is appropriately monitored during refueling operations.

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 1 Attachment 1, Volume 14, Rev. 0, Page 30 of 138

Attachment 1, Volume 14, Rev. 0, Page 31 of 138 Improved Standard Technical Specifications (ISTS) Bases Markup and Justification for Deviations (JFDs)

Attachment 1, Volume 14, Rev. 0, Page 31 of 138

Attachment 1, Volume 14, Rev. 0, Page 32 of 138 Nuclear Instrumentation B 3.9.2 B 3.9 REFUELING OPERATIONS B 3.9.2 Nuclear Instrumentation

[ Reactor Protection System (RPS) and Post Accident Monitoring (PAM) Instrumentation I D A C' CC BACKGROUND The source range neutron flux monitors are used during refueling operations to monitor the core reactivity condition. The installed source range neutron flux monitors are part of the lNuclear Ins§1 mentation I' 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 installed~souc range neutron flux monitors areFBF3 det ctors loperating in the Oroportional region ofthe gas filled detector Oharacteristic curve. The detectors monitor the neutron flux in counts per second. The instrument range cover si-decades of neutron flux ,1E+6 cps)[*"* 1E-1 cps to 1[51% instrurre'nt accuracyI. The detectors also provide continuous visual indication in the control room land an au fble alarm to alert operators to a RPS possible dilution accident. The [N Ss designed in accordance with the criteria presented in Reference 1. +lf used, portable detectors should be functionally equivalent to the installed N S 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 accidentlor an imprerly Ioade Ifuel asdembl . The safety analysis of the unco role boron dilution accident is described in Reference 2. The analysis of theu boron dilution accident shows that the normally available SDM would not be lost, and there is sufficient time for the operator to take corrective action.

The source range neutron flux monitors satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii).

LCO This LCO requires two source range neutron flux monitors'OPERABLE to INSERT 3 Q ensure that redundant monitoring capability is available to detect changes in core reactivity. C)

APPLICABILITY In MODE 6, the source range neutron flux monitors must be OPERABLE 0 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

same installedsource 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 Monitorinq (PAM) Instrumentation."

BWOG STS B 3.9.2-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 14, Rev. 0, Page 32 of 138

Attachment 1, Volume 14, Rev. 0, Page 33 of 138 B 3.9.2 O INSERT 1 two high sensitivity proportional counters (BF3 chambers) located on opposite sides of the core.

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 (1E-1 cps to 1E+5 cps). The detectors also provide continuous visual indication in the control room and an audible indication to alert operators.

O INSERT 3 one from each side of the core, be O INSERT 4 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.

Insert Page B 3.9.2-1 Attachment 1, Volume 14, Rev. 0, Page 33 of 138

Attachment 1, Volume 14, Rev. 0, Page 34 of 138 Nuclear Instrumentation B 3.9.2 BASES ACTIONS A.1 and A.2 With only one frequired*]source range neutron flux.monitor OPERABLE, redundancy has been lost. Since these instruments are the only direct means of monitoring core reactivity conditions, ICORE ALTERATIONS TSTF and introduction of coolant into the RCS with boron concentration less positive reactivity -471 than required to meet the minimum boron concentration of LCO 3.9.1 additions must be suspended immediately. Suspending positive reactivity additions that could result in failure to meet the minimum boron concentration limit is required to assure continued safe operation. Introduction of coolant inventory must be from sources that have a boron concentration greater than Ftwhat 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. Performance of Required Action A.1 shall not preclude completion of movement of a component to a safe position.

B.1 With no frequired*]source range neutron flux monitor OPERABLE, action to restore a monitor to OPERABLE status shall be initiated immediately.

Once initiated, action shall be continued until a source range neutron flux monitor is restored to OPERABLE status.

B.2 With no required source range neutron flux monitor OPERABLE, there is no direct means oS detecting changes in core reactivity. However, since TSTF ICORE ALTFTIONS and positive reactivity additions are not to be -471 made, the core reactivity condition is stabilized until the source range neutron flux monitors are OPERABLE. This stabilized condition is determined by performing SR 39.1.1 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 changes in boron concentration would be 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. 3.0, 03/31/04 Attachment 1, Volume 14, Rev. 0, Page 34 of 138

Attachment 1, Volume 14, Rev. 0, Page 35 of 138 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 parameter on other channels. Itlis based on the assumption that the two indication channels should be consistent with core conditions. Changes in fuel loading and core geometry can result in significant differences between source range channels, but each channel should be consistent with its local conditions.

The Frequency of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is consistent with the CHANNEL CHECK Frequency specified similarly for the same instruments in LCO 3.3.9.

SR 3.9.2.2 SR 3.9.2.2 is the performance of a CHANNEL CALIBRATION every M118Mmonths. This SR is modified by a Note stating that neutron detectors 0 RS are excluded from the CHANNEL CALIBRATION. The CHANNEL _ channel I CALIBRATION for the source range Lnu ea is a complete check and re- R.S-cabinet (

adjustment of the channels, from thelpre-agplifierinput to the indicator The 18 month Fre:luency is based on the need tq perform this Surveillance duririg the conditions that apply durhg a plant outage.

(D Operating experience has shown these components usually pass the Surveillance when performed at the [D8*month Frequency. _.

REFERENCES 1. 110CFR50 Ap dix A, GDC 13 GDC 26, GDt 28. and GDC 24.

00 E'-* UFSAR, Appendices 3D.1.9. 3D.1.16, 3D.1.17. 3D.1.18, 3D.1.19. 3D.1.20, and 3D.1.25 2.-- FSAR, Section [ . 15.2.4 and Appendix 4B al and for the PAM source range channels is complete check of the instrument channel l 0 BVWOG STS B 3.9.2-3 Rev. 3.0, 03/31/04 Attachment 1, Volume 14, Rev. 0, Page 35 of 138

Attachment 1, Volume 14, Rev. 0, Page 36 of 138 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. Changes are made to be consistent with changes made to the Specification. Davis-Besse has four source range neutron flux monitors available to monitor core reactivity conditions during refueling. Two from the Reactor Protection System (RPS) source range neutron flux monitors and two from the Post Accident Monitoring (PAM) source range neutron flux monitors. Any combination of the two, as long as there is one from each side of the reactor core, is allowed.

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. 0, Page 36 of 138

Attachment 1, Volume 14, Rev. 0, Page 37 of 138 Specific No Significant Hazards Considerations (NSHCs)

Attachment 1, Volume 14, Rev. 0, Page 37 of 138

Attachment 1, Volume 14, Rev. 0, Page 38 of 138 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. 0, Page 38 of 138

, Volume 14, Rev. 0, Page 39 of 138 ATTACHMENT 3 ITS 3.9.3, CONTAINMENT PENETRATIONS , Volume 14, Rev. 0, Page 39 of 138

, Volume 14, Rev. 0, Page 40 of 138 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) , Volume 14, Rev. 0, Page 40 of 138

Attachment 1, Volume 14, Rev. 0, Page 41 of 138 IITS 3.9.3 REFUELING OPERATIONS CONTAINMENT PENETRATIONS LIMITING CONDITION FOR OPERATION LCO 3.9.3 3.9.4 The containment penetrations shall be in the folowing status:

aLhe equipment hatch cover closed and held in lace by a minimum of four bol except the L01 LCO 3.9.3.a Iequipme.t hatch may be open providefir the requirements of Speification *. .12 are Z, Lsatisfi*,

LCO 3.9.3.b b. A minimum of one door in each air lock closed,[but both doors of the ulontainmnent personnel air lc~k may be open provided that at least one personnel aii lock door is capable of being closed and aýdesign ted individual is available immediately *utside the personnel air lock to close the door,land LCO 3.9.3.c c. Each penetration providing direct access from the containment atmosphere to the atmosphere outside containment shall be either:

1. Closed by a manual or automatic isolation valve, blind flange, or equivalent, or containnient purge -" LC "

2. Be capable of being closed from *econtrol room by an OPERABLE L01*._

and exhaust valve/upon receipt ofta higýýUdathaon s~ig~nal fron2the containment purge*-

Fand exhaust sygtem noble gas mo)nitoXf INER 1 A02 APPLICABILITY: During 00 ALýTI(S o movement o irradiated fuel within the containment.L0 ACTION:

ACTION A a. With the requirements of the above specification not satisfied, immediately suspend all operations involving K-ORr, ALTF/RATIQ'NS o* movement of irradiated fuel in the containment.

b. With the require nts of Specification 3.9. .c not satisfied for the c tainment purge and A03 exhaust syste close at least one of the is ation valves for each of he purge and exhaust penetrations poviding direct access from e containment atmospgre to the outside atmosphere 'thin one hour.

lc. The pr ~isions of Specifatidn 3.0.3 are ot applicable.

SURVEILLANCE REQUIREMENTS SR 3.9.3.1, 4.9.4 Each of the above required containment penetrations shall be determined to be either in its SR 3.9.3.2 required condition or capable of being closed y an OPERABLE containment ur e and exhaust valve, ýwithin 100 ho~utws'por to the o! an'd at leastlonee pr 7days during RE-ALE*nN~ r movement of irradiated fuel in the containment, by:L0 SR 3.9.3.1 a. Verifying the penetrations are in their required condition, or SR 3.9.3.2 b. Verifying that with the containment purge and exhaust system in operation, and the LAO1 containment purge/and exhaust system 96ble gas monitor capable of oviding a high radiation signal t the control room, ttafter initiauion of the high diation signal, the .-

containment purge and exhaust isolation valves can be closed fom tho-dontrol room DAVIS-BESSE, UNIT 1 314 9-4 Amendment No. 186, 202, 221,251 Page 1 of 2 Attachment 1, Volume 14, Rev. 0, Page 41 of 138

Attachment 1, Volume 14, Rev. 0, Page 42 of 138 3.9.3 S0 INSERT 1

-- .NOTE------------ --------------

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

Insert Page 3/4 9-4 Page 2 of 2 Attachment 1, Volume 14, Rev. 0, Page 42 of 138

Attachment 1, Volume 14, Rev. 0, Page 43 of 138 DISCUSSION OF CHANGES ITS 3.9.3, CONTAINMENT PENETRATIONS 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 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.

A03 CTS 3.9.4 Action b provides actions to be taken when Specification 3.9.4.c is not met, and requires the associated penetration to be isolated within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

However, CTS 3.9.4 Action a provides the requirements when any portion of Specification 3.9.4 is not met, including Specification 3.9.4.c, and requires, in part, the immediate suspension of movement of irradiated fuel assemblies within containment. Under similar conditions, ITS 3.9.3 ACTION A also requires the immediate suspension of movement of recently irradiated fuel assemblies within containment. This changes the CTS by deleting a redundant Action. The change related to only requiring suspension of recently irradiated fuel is discussed in DOC L02.

Completing the action required by CTS 3.9.4 Action a will place the unit outside the Applicability of the LCO. Therefore, CTS 3.9.4 ACTION b is not required once CTS 3.9.4 Action a is complete. Since CTS 3.9.4 Action a has an immediate completion time and CTS 3.9.4 Action b allows one hour to complete the action, CTS 3.9.4 Action a would be completed prior to the expiration of the CTS 3.9.4 Action b completion time. Therefore, CTS 3.9.4 Action b is redundant to CTS 3.9.4 Action a and has been deleted. This change is designated as administrative because it does not result in any technical changes to the CTS.

A04 CTS 3.9.4 Action c states that the provisions of Specification 3.0.3 are not applicable. ITS 3.9.3 does not include this exception. This changes the CTS by deleting the specific exception to Specification 3.0.3.

This change is acceptable because it results in no technical change to the Technical Specifications. ITS LCO 3.0.3 (which is equivalent to CTS 3.0.3) specifically states that it is not Applicable in MODE 6, which is the MODE in which ITS 3.9.3 is required to be Applicable (since movement of irradiated fuel assemblies within the containment can only occur in MODE 6). Therefore, this exception to CTS 3.0.3 is redundant and unnecessary. This change is designated as administrative because it does not result in a technical change to the CTS.

Davis-Besse Page 1 of 4 Attachment 1, Volume 14, Rev. 0, Page 43 of 138

Attachment 1, Volume 14, Rev. 0, Page 44 of 138 DISCUSSION OF CHANGES ITS 3.9.3, CONTAINMENT PENETRATIONS MORE RESTRICTIVE CHANGES None RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES LA01 (Type 1 - Removing Details of System Design and System Description, Including Design Limits) CTS 3.9.4.c.2 requires each penetration be capable of being closed "from the control room" by an OPERABLE containment purge and exhaust valve "upon receipt of a high radiation signal from the containment purge and exhaust system noble gas monitor." CTS 4.9.4.b requires a verification that the "containment purge and exhaust system noble gas monitor capable of providing a high radiation signal to the control room, that after initiation of the high radiation signal," the containment purge and exhaust isolation valves can be closed "from the control room." ITS LCO 3.9.3.c.2 and SR 3.9.3.2 do not state the specific type of signal or that the valves must be operated from the control room, but only require the valves to be manually closed. This changes CTS by moving the type of actuation signal (i.e., high radiation from the containment purge and exhaust noble gas monitor) and that the valves are operated from the control room 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 to verify that appropriate equipment can be actuated upon receipt of an actuation signal. 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 of the ITS. 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 (Category2 - Relaxation of Applicability) 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. Furthermore, as described in DOC A02, 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 Davis-Besse Page 2 of 4 Attachment 1, Volume 14, Rev. 0, Page 44 of 138

Attachment 1, Volume 14, Rev. 0, Page 45 of 138 DISCUSSION OF CHANGES ITS 3.9.3, CONTAINMENT PENETRATIONS atmosphere to the outside atmosphere to be unisolated under administrative controls. ITS 3.9.3 is applicable during movement of recently irradiated fuel assemblies within containment. In addition, ITS 3.9.3 requires the equipment hatch to be closed and one door in each air lock to be closed; no allowances to open the equipment hatch or open both doors in an air lock is provided.

Furthermore, the proposed CTS Note allowance is not being retained in ITS 3.9.3. This changes the CTS by eliminating requirements for Containment Penetrations during CORE ALTERATIONS and reduces the Applicability to only when moving "recently" irradiated fuel assemblies in lieu of all irradiated fuel assemblies. This change also deletes the allowance for the equipment hatch to be open under a specific condition, both air lock doors to be open under a specific condition, and the allowance to unisolate certain penetration flow paths under administrative controls.

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 deleted.

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 specifically being maintained in ITS 3.3.15, the proposed deletion of CORE ALTERATIONS is justified. Furthermore, reducing the Applicability 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 Plant limits (well within 10 CFR 100). Due to the addition of the term "recently," the allowance to open the equipment hatch, both doors in an air lock, or a penetration flow path providing direct access from the containment atmosphere to the outside atmosphere is not retained. Currently, the equipment hatch and both air lock doors are only opened after the reactor had been in the shutdown condition for 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . Furthermore, the License Amendment request adding the proposed CTS Note also describes that the Note will not be used until the unit has been shutdown for 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . Since 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months is the time limit for the term "recently," the CTS allowance is redundant and can be deleted. This change is designated as less restrictive because the ITS LCO requirements are applicable in fewer operating conditions than in the CTS.

L02 (Category 7- Relaxation Of Surveillance Frequency, Non-24 Month Type Change) CTS 4.9.4 states that specified containment penetration Surveillances shall be performed, in part, "within 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months prior to the start of' the specified conditions in the Applicability. ITS SR 3.9.3.1 and ITS SR 3.9.3.2 do not include the "within 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months prior to the start of" Frequency. This changes the CTS by eliminating the stipulation that the Surveillances be met within 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months prior to entering the conditions specified in the Applicability.

The purpose of CTS 4.9.4 is to verify the equipment required to meet the LCO is OPERABLE. This change is acceptable because the new Surveillance Frequency provides an acceptable level of equipment reliability. ITS SR 3.0.1 states "SRs shall be met during the MODES or other specified conditions in the Davis-Besse Page 3 of 4 Attachment 1, Volume 14, Rev. 0, Page 45 of 138

Attachment 1, Volume 14, Rev. 0, Page 46 of 138 DISCUSSION OF CHANGES ITS 3.9.3, CONTAINMENT PENETRATIONS Applicability for the individual LCOs, unless otherwise stated in the SR."

Therefore, the ITS requires that the Surveillances must be met prior to the initiation of movement of irradiated fuel. For CTS 4.9.4, the periodic Surveillance Frequency for verifying containment penetrations are in the required status is acceptable during the conditions specified in the Applicability, and is also acceptable during the period prior to entering the conditions specified in the Applicability. This change is designated as less restrictive because Surveillances will be performed less frequently under the ITS than under the CTS.

L03 (Category 7 - Relaxation Of Surveillance Frequency, Non-24 Month Type Change) CTS 4.9.4 and CTS 4.9.4.b include a Surveillance Frequency of "once per 7 days" during conditions specified in the Applicability for performing a Surveillance of the containment purge supply and exhaust isolation system. This test is only required if the penetrations are not in their required conditions since either CTS 4.9.4.a and 4.9.4.b is required to be performed. The ITS SR 3.9.3.2 Frequency for the same requirement is 24 months. ITS SR 3.9.3.2 is also modified by a Note that states that SR 3.9.3.2 is not required to be met for containment purge supply and exhaust valve(s) in penetrations that are closed to comply with LCO 3.9.3.c.1. This changes the CTS by changing the Surveillance Frequency from 7 days to 24 months.

The purpose of CTS 4.9.4 and CTS 4.9.4.b is to verify the equipment required to meet the LCO is OPERABLE. This change is acceptable because the new Surveillance Frequency ensures that it provides an acceptable level of equipment reliability. The CTS Applicability for this Specification is during CORE ALTERATIONS or movement of irradiated fuel within the containment. Thus, CTS SR 4.9.4.b only has to be performed during these conditions. Specifically, the current Frequency of "within 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months prior to the start of' CORE ALTERATIONS or movement of irradiated fuel in the containment is performed prior to commencing refueling operations. Since the Davis-Besse is operating on a 24 month refueling outage cycle, this effectively means that the interval between the end of one refueling outage (when the Surveillance would be performed for the last time during the refueling outage) and the beginning of the next refueling outage (when the Surveillance would be performed for the first time for the next refueling outage) is approximately 24 months. Thus, this 24 month Frequency provides an appropriate degree of assurance that the valves can be closed, when needed. This change is designated as less restrictive because Surveillances will be performed less frequently under the ITS than under the CTS.

Davis-Besse Page 4 of 4 Attachment 1, Volume 14, Rev. 0, Page 46 of 138

Attachment 1, Volume 14, Rev. 0, Page 47 of 138 Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs)

Attachment 1, Volume 14, Rev. 0, Page 47 of 138

Attachment 1, Volume 14, Rev. 0, Page 48 of 138 CTS Containment Penetrations 3.9.3 3.9 REFUELING OPERATIONS 3.9.3 Containment Penetrations 3.9.4 LCO 3.9.3 The containment penetrations shall be in the following status:

3.9.4.a a. The equipment hatch closed and held in place by four bolts l5-o 0

3.9.4.b b. One door in each air lock is [caable beir closedz1 and 00 3.9.4.c c. Each penetration providing direct access from the containment atmosphere to the outside atmosphere either:

1. Closed by a manual or automatic isolation valve, blind flange, or equivalent-orj7 0

2. Capable of 4eing closed by an OPERABLE Aontainment purge and fxhaustJsolation i 0 NOTE--'.--- ...............-

Penetration flow p (s) providing direct access from t e containment atmosphere to th outside atmosphere may be uniso ted under 0

administrative c ntrols.

APPLICABILITY: During movement ofrgrecentlyqjirradiated fuel assemblies within containment.

0 1

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME Action a A. One or more A.1 Suspend movement of Immediately containment penetrations not in

[Jrecentlj irradiated fuel assemblies within 0

required status. containment.

BVWOG STS 3.9.3-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 14, Rev. 0, Page 48 of 138

Attachment 1, Volume 14, Rev. 0, Page 49 of 138 CTS Containment Penetrations 3.9.3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY 4.9.4. SR 3.9.3.1 Verify each required containment penetration is in 7 days 4.9.4.a the required status.

4.9.4, SR 3.9.3.2 ............--------.... NOTE -....--------------------

4.9.4.b Not required to be met for containment purge and exhaust valve(s) in penetrations closed to comply with LCO 3.9.3.c.1.

be manually placed in Verify each required containment purge and

[vlv ac tes to

.the isolation position F months 00 Ian actual or sjtiplted actuation siaonal.

BWOG STS 3.9.3-2 Rev. 3.0, 03/31/04 Attachment 1, Volume 14, Rev. 0, Page 49 of 138

Attachment 1, Volume 14, Rev. 0, Page 50 of 138 JUSTIFICATION FOR DEVIATIONS ITS 3.9.3, CONTAINMENT PENETRATIONS

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

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

3. The containment purge and exhaust valves are not automatically actuated by a high radiation signal. When a high radiation signal is alarmed, the unit operators are required to manually close the containment purge and exhaust valves. Therefore ISTS SR 3.9.3.2 has been modified to reflect this design, consistent with the current licensing basis. Since the valves are manually closed, there is no reason to mention the actuation signal. The capability to close the valves is not affected by the presence, or lack of presence, of an alarm. The alarm itself is tested as part of ITS 3.3.15.

4. The ISTS LCO 3.9.3.b bracketed allowance that the door in each air lock is only required to be "capable of being" closed has not been retained in ITS LCO 3.9.3.b.

The ISTS LCO Note allowing penetration flow path(s) to be unisolated has also not been retained in ITS 3.9.3. These allowances have not been retained since the Applicability of ITS 3.9.3 is during movement of "recently" irradiated fuel assemblies.

Currently, both air lock doors are only opened after the reactor had been in the shutdown condition for 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . Furthermore, the Note also cannot be used until the unit has been shutdown for 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . Since 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months is the time limit for the term "recently," the CTS allowance is redundant and can be deleted.

Davis-Besse Page 1 of 1 Attachment 1, Volume 14, Rev. 0, Page 50 of 138

Attachment 1, Volume 14, Rev. 0, Page 51 of 138 Improved Standard Technical Specifications (ISTS) Bases Markup and Justification for Deviations (JFDs)

Attachment 1, Volume 14, Rev. 0, Page 51 of 138

Attachment 1, Volume 14, Rev. 0, Page 52 of 138 I

All changes are unless otherwise noted 9 Containment Penetrations B 3.9.3 B 3.9 REFUELING OPERATIONS 8 3.9.3 Containment Penetrations BASES BACKGROUND During movement of [recentlyJirradiated fuel assemblies within containment, a release of fission product radioactivity within containment 0

will be restricted from escaping to the environment when the LCO requirements are met. In MODES 1, 2, 3, and 4, this is accomplished by maintaining containment OPERABLE as described in LCO 3.6.1, "Containment." In MODE 6, the potential for containment pressurization as a result of an accident is not likely; therefore, requirements to isolate the containment from the outside atmosphere can be less stringent. The LCO requirements are referred to as "containment closure" rather than "containment OPERABILITY." Containment closure means that all potential escape paths are closed or capable of being closed. Since there is no potential for containment pressurization, the Appendix J leakage criteria and tests are not required.

The containment serves to contain fission product radioactivity that may be released from the reactor core following an accident, such that offsite radiation exposures are maintained well within the requirements of 10 CFR 100. Additionally, the containment provides radiation shielding from the fission products that may be present in the containment atmosphere following accident conditions.

The containment equipment hatch, which is part of the containment pressure boundary, provides a means for moving large equipment and close and closed components into and out of containment. During movement of[ecentlyfJ irradiated fuel assemblies within containment, the equipment hatch must 0

0 b held in place by at least four bolts. Good engineering practice dictates that the bolts required by this LCO be approximately equally spaced.

The containment air locks, which are also part of the containment pressure boundary, provide a means for personnel access during MODES 1, 2, 3, and 4 unit operation in accordance with LCO 3.6.2, "Containment Air Locks." Each air lock has a door at both ends. The doors are normally interlocked to prevent simultaneous opening when containment OPERABILITY is required. During periods of unit shutdown when containment closure is not required, the door interlock mechanism may be disabled, allowing both doors of an air lock to remain open for extended periods when frequent containment entry is necessary. During movement of recentlyMirradiated fuel assemblies within containment, 0

containment closu e is required therefore, the door i/terlock mechanism may remain disab ed, but ýne air lock door must always remain 00 Of§ziI closed.

BWOG STS B 3.9.3-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 14, Rev. 0, Page 52 of 138

Attachment 1, Volume 14, Rev. 0, Page 53 of 138 All changes are a unless otherwise noted 9 Containment Penetrations B 3.9.3 BASES BACKGROUND (continued)

The requirements on containment penetration closure ensure that a release of fission product radioactivity within containment will be restricted

[INSERT 1 to within regulatory limits.

IThe Contain, ent Purge and Exhas A, yste 0m incuer w sbytm.

_ _ _ _ _ _ _ __8 inch purge penetration and a "M inch exhaust penetration. e secon dj.ýbsystem, or mi purge system, Tin/ludes an [8] inch purge penetrition and an [81 inghn exhaust\

penetrat6n. _ During MODES 1, 2, 3, and 4, the two valves in each of the no--ma position.purge and exhaust penetrations are secured in the closed The l(o valves 'in each of th two minipurge penetra ions can\

be openedVn /rmittently but a re clo s* automatically by the *ngineered\

Safety Feat re Actuation System (E./&AS). Neither of the s bsystems is\

Isubject to flSpecification in MODF_/5.

INSERT 2 In MODE 6, larcb air exchangers are /ecessary to cond ct refueling operations. Th* normal [42] inch pu e system is used r this purpose, and all four vales are closed on a r actor building (RB* high radiation signal in accodance with LCO 3.3. 5, "Reactor Buildin) (RB) Purge Isolation- Hi Radiation.

The other containment penetrations that provide direct access from containment atmosphere to outside atmosphere must be isolated on at least one side. Isolation may be achieved by an OPERABLE automatic isolation valve or by a manual isolation valve, blind flange, or equivalent. by an engineering Equivalent isolation methods must be approved n may include use of a, evaluation material that can provide a temporary, atmospheric pressure ventilation barrier for the other containment penetrations during fuel movements

[involving handling recently irradiated fuelE](Ref. 1).

APPLICABLE During moveme t of [recently] irradiat d fuel assemblies thin SAFETY containment, th most severe radiolo cal consequences esult from a ANALYSES fuel handling a ident [involving hand ng recently irradiat d fuel]. The fuel handling ac ident is a postulated vent that involves amage to irradiated fuel ( ef. 2). Fuel handling accidents, analyze in Ref. 3, include droppin a single irradiated f el assembly and ha dling tool or a heavy object o to other irradiated fuI assemblies. Th e r quirements of INSERT 3 LCO 3.9.6, "R ueling Canal Water vel," in conjunction with minimum decay time of [ 00] hours prior to [irr diated] fuel mover nt [with 0

containment cl sure capability or a inimum decay time of [X] days without contai ment closure capabili ], ensures that the release of fission product radio ctivity subsequent to fuel handling accident results in doses that are within the requireme ts specified in 10 C:R 100. The acceptance ii its for offsite radiatio exposure are cont med in Ref. 2.

Containment enetrations satisfy C terion 3 of 10 CFR 0.36(c)(2)(ii).

BWOG STS B 3.9.3-2 Rev. 3.0, 03/31/04 Attachment 1, Volume 14, Rev. 0, Page 53 of 138

Attachment 1, Volume 14, Rev. 0, Page 54 of 138 B 3.9.3 O INSERT I The Containment Purge and Exhaust System is part of the Containment Vessel Air Purification and Cleanup System. The Containment Purge and Exhaust System is designed to provide clean fresh air to the containment vessel. This is accomplished via O INSERT 2 In MODE 6 when access to the containment is desired, the Containment Purge and Exhaust System is aligned to containment as a source of clean fresh air. The containment purge and exhaust supply and exhaust fans shut down and associated inlet and outlet dampers close on a high radiation signal (LCO 3.3.15, "Containment Purge and Exhaust Isolation -

High Radiation"). The containment purge and exhaust valves are then manually closed by operator action.

O INSERT 3 Containment closure is not assumed in the fuel handling accident inside containment (Ref.2). 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, containment closure is not needed. While the Technical Specifications do not include a requirement on decay time, the Technical Requirements Manual (TRM) includes a requirement that no fuel movement will commence until 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months after shutdown. This ensures this assumption (that containment closure is not needed) is valid. However, the containment penetrations are required to be in the required status during movement of recently irradiated fuel within the containment to mitigate the consequences of a fuel handling accident if it occurs prior to the unit being shut down for 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

Maintaining the containment penetrations in the required status is not assumed in MODE 6. However, they can provide mitigation of a fuel handling accident that occurs prior to the unit being shut down for 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . Therefore, containment penetrations are being retained in the Technical Specifications.

Insert Page B 3.9.3-2 Attachment 1, Volume 14, Rev. 0, Page 54 of 138

Attachment 1, Volume 14, Rev. 0, Page 55 of 138 B 3.9.3 Containment Penetrations B 3_9.3 BASES LCO -* con~tainme~nt E IW prsonnel F'8 NOTEairlock do( rs open and The allowance have penetration flov paths with direct acc ss from the contair ment atmosphere to *he outside atmosphe re to be unisolated Iuring fuel movement and CORE A-'TERATIONKSis basedon (1) c* nfirrnatory dose calculations of fuel ha~ndling accide pt as approved by tt e NRC staff TSF 47 which indicate *ceptabl railgij I consequences anI 2)emitmlts fothliesetimplement acoept ble administrat~ive procedures that ensu inthe event of a r;fueling accident (even thoug3h he conta iiment fissio 1 product control fu ction is not required to r et acceptable dose ,sequences) - that t e open airlock can and will b:.promptly' closed folh wing containment, acuation and that the open penetration(s) an an 1will be promptly cl, sed. The time to close such pq netrations or combin* tion of penetrations shall be included in the confirn atory dose alutirs.

This LCO limits the consequences of a fuel handling accident Tinvolving (7 handling recently irradiated fuelfJin containment by limiting the potential escape paths for fission product radioactivity from containment. The LCO requires any penetration providing direct access from the containment atmosphere to the outside atmosphere to be closed except for the OPERABLE containment purge and exhaust penetrations a the4 by an operator if a Icontainment p ovr-nnel airlocksj[* For the OPERABLE containment purge '-' "-'

Containment Purge and and exhaust penetrations, this LCO ensures that these penetrations are s received haudioation isolablelbythe B solationl signal. IThe OPERABILITY/

requiremen or this LCO ensure tha t e automatic purge a d exhaust valve clsre times specified in the F/SAR can be achieved a/d therefore 0 meet the ssumptions used in the s fety analysis to ensure'eleases through te valves are terminated ýuch that radiological doqes are within the acceptance limit. J The LCO is moctied by a Note allowiqg penetration flow ilaths with direct access from men atmosp ere to the outside a:mosphere to be unisolated und r administrative contr Is. Administrative ntrols ensure that 1) appropriate personnel are av re of the open stats of the penetration floi path during CORE ONS or rrbvement of irradiated fuel ssemblies within co ainment, and 2) sp cified individuals are designatel and readily availabl to isolate the flow ath in the event of a fuel hancfling accident. /

The containmert personnel airlock dolrs may be open du'ing movement of [recently] irra/i iated fuel in the contj inment provided thi t one door is capable of beiog closed in the event/of a fuel handling a fcident. Should a fuel handling #ccident occur inside I!ontainment, one personnel airlock 0 door will be coosed following an evakuation of containmpnt.

BWOG STS B 3.9.3-3 Rev. 3.0, 03/31/04 Attachment 1, Volume 14, Rev. 0, Page 5 °o6ttir t Penetrations B 3.9.3

Attachment 1, Volume 14, Rev. 0, Page 56 of 138 B 3.9.3 0 INSERT 4 The equipment hatch is required to be closed and held in place by four bolts.

Insert Page B 3.9.3-3 Attachment 1, Volume 14, Rev. 0, Page 56 of 138

Attachment 1, Volume 14, Rev. 0, Page 57 of 138 Containment Penetrations B 3.9.3 BASES APPLICABILITY The containment penetration requirements are applicable during movement of t*ecently* irradiated fuel assemblies within containment because this is when there is a potential for the limiting fuel handling 0

accident. In MODES 1, 2, 3, and 4, containment penetration MODE requirements are addressed by LCO 3.6.1. In MODES 5 and6wen (

movement of irradiated fuel assemblies within containment is not being conducted, the potential for a fuel handling accident does not exist.

While in MODE 6 without 1o1 110 y, ue -ecay,a. oav a fuel handling I involving

ýhaedlingreently irradiated fuel (i.e., fuel that has occus.ed art of a 72 hou 0 conainentglosre apailiy.]Therefore, under these conditions no E~ff requirements are placed on containment penetration status.

The addition of- e term "recently" as.ociated'S-REVIEWE NOTEwith handli g irradiated fuel in all of the cont inment function Tec nical Specification r quirements is only applicable o those licensees wh, have demonstrate by analysis that after suffici nt radioactive decay as occurred, off-sit doses 0

resulting from a fuel handling acciden remain below the tandard Review Plan limits (well within 10 CFR 100).

Additionally, lic nsees adding the ter "recently" must m ke the following commitment ich is consistent with UMARC 93-01, R vision 4, Section 11.36. "Safety Assessmen for Removal of Eq ipment from Service During Shutdown Conditions " subheading "Con inment -

Primary (PWR Secondary (BWR)."

"The following uidelines are include in the assessmen of systems removed from ervice during move nt of irradiated fue:

- During f el handlinglcore alter tions, ventilation s tem and radiation onitor availability (as defined in NUMAR 91-06) should be asses ed, with respect to filt ation and monitorin of releases from the fuel. ollowing shutdown, r dioactivity in the fu I decays away fairly rapi ly. The basis of the echnical Specificati n operability amend ntis the reduction in oses due to such d cay. The goal of maintaini g ventilation system nd radiation monit r availability is to reduce d ses even further belo that provided by t e natural decay.

- A singl normal or contingen y method to prompt y close primary or seconda containment penetr tions should be de eloped. Such prompt thods need not co pletely block the pe etration or be capable f resisting pressure.

BVVOG STS B 3.9.3-4 Rev. 3.0, 03/31/04 Attachment 1, Volume 14, Rev. 0, Page 57 of 138

Attachment 1, Volume 14, Rev. 0, Page 58 of 138 B 3.9.3 0 INSERT 5 in progress, the LCO does not apply. This is because the accident analysis has demonstrated 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).

Insert Page B 3.9.3-4 Attachment 1, Volume 14, Rev. 0, Page 58 of 138

Attachment 1, Volume 14, Rev. 0, Page 59 of 138 B 3.9.3 Containment Penetrations B 3.9.3 BASES APPLICABILITY (continued)

The purpose of he "prompt methods" mentioned above a e to enable ventilation syst ms to draw the relea e from a postulated fuel handling accident in the proper direction such that it can be treate and 0 monitored." "_ _

_--------t-_-e~ ........... . ... ._. _ . . . ......

ACTIONS A.1 With the containment equipment hatch, air locks, or any containment penetration that provides direct access from the containment atmosphere to the outside atmosphere not in the required status, including the being manually closed

_ontainment

-- u oma Ourge and Vxhaustisolation Syt__iot capable of uaion when the purge and exhaust valves are open, the 0

unit must be placed in a condition in which the isolation function is not needed. This is accomplished by immediately suspending movement of DecentlyE irradiated fuel assemblies within containment. Performance of 0 these actions shall not preclude moving a component to a safe position.

SURVEILLANCE SR 3.9.3.1 REQUIREMENTS This Surveillance demonstrates that each of the containment penetrations required to be in its closed position is in that position. The Surveillance on the open purge and exhaust valves will demonstrate that the valves are not blocked from closing. Also the Surveillance will demonstrate that each valve operator has motive power, which will ensure each valve is 0 capable of being closed by an OPERABLE purge isolationl The Surveillance is performed every 7 days during movement ofPecentlyl 0

irradiated fuel assemblies within the containment. The Surveillance interval is selected to be commensurate with the normal duration of time to complete fuel handling operations. A surveillance before e start of refueling operatio's will provide two irthree surveillance v, ifications 0 during the applicqble period for thi CO.

As such, this Si1veillance ensures th t a postulated fuel andling accident [involving handling recently/irradiated fuel] that releases fission product radioa*tivity within the cont*inment will not res It in a release of significant fision product radioactivity to the environ nt in excess of those recomrended by Standard eview Plan Sectio 15.7.4 (Ref. 3).

BWOG STS B 3.9.3-5 Rev. 3.0, 03/31/04 Insert Page B 3.9.3-4 Attachment 1, Volume 14, Rev. 0, Page 59 of 138

Attachment 1, Volume 14, Rev. 0, Page 60 of 138 Containment Penetrations B 3.9.3 BASES SURVEILLANCE REQUIREMENTS (continued)

S R 3.9.3.2 1 from the control room can be manually placed in _This Surveillance demonstrates that each containment purge and exhaust tes te its isolation position Ion manual i ition or on an actual a my dvalveactv require met .I o r si mu la te d-hih LGO ra d ia 3tio3.-15 n .-s i gn a, l"ý TBPge h e J3_mo n th F re q u e n cy ma in t a in s 24 consistency with other similar[ESFAS instrumentation isolation - and valve testing Hg h R and a CHAN ia tinc the N/LFUNCTIONAL TE ST every 92 isolation tec insru, ntation requires a HANNEL CHECK to roscde days to ever e ure n 12thehours rnt rateo vh n ery ming re a bengcosdfrma cissintelpodcLtI nnelv fu eli o O i aureiap A IYsbl o a ing efulo after a tpo nstuaefuel)ae m nhsandlin udingrofm CHANNEL p thecontai*nmnt.

Iroduct ra ioactivity fro The systeimen l t.o 8mit aes o s 0ve The SR is modified by a Note stating that this Surveillance is not required to be met for valves in isolated penetrations. The LCO provides the option to close penetrations in lieu of requiring capability.

actuation 0

REFERENCES 1. GPU Nuclear Safety Evaluation SE-0002000-001, Rev. 0, May 20, 1988.

2>. NFSAR Sectionn15.7 G a e 1 ,. uy19 1.

00 0.NUREG-0800 in 15.7.4, RevL. 1 ýJul .1981.1 0 BWOG STS B 3.9.3-6 Rev. 3.0, 03/31/04 Attachment 1, Volume 14, Rev. 0, Page 60 of 138

Attachment 1, Volume 14, Rev. 0, Page 61 of 138 JUSTIFICATION FOR DEVIATIONS ITS 3.9.3 BASES, CONTAINMENT PENETRATIONS

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. Changes made to be consistent with the Specification.

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

5. This Reviewer's Note has been deleted. This information is for the NRC reviewer to be keyed in to what is needed to meet this requirement. This is not meant to be retained in the final version of the plant specific submittal. In addition, the TSTF-471T changes to the Reviewer's Note are not shown.

6. Typographical error corrected.

7. Editorial changes for clarity and consistency.

8. This information is discussing the Applicable Safety Analysis for this Specification and is not necessary to be included in a Surveillance Requirement discussion.

Davis-Besse Page 1 of 1 Attachment 1, Volume 14, Rev. 0, Page 61 of 138

Attachment 1, Volume 14, Rev. 0, Page 62 of 138 Specific No Significant Hazards Considerations (NSHCs)

Attachment 1, Volume 14, Rev. 0, Page 62 of 138

Attachment 1, Volume 14, Rev. 0, Page 63 of 138 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.9.3, CONTAINMENT PENETRATIONS There are no specific NSHC discussions for this Specification.

Davis-Besse Page 1 of 1 Attachment 1, Volume 14, Rev. 0, Page 63 of 138

Attachment 1, Volume 14, Rev. 0, Page 64 of 138 ATTACHMENT 4 ITS 3.9.4, DHR AND COOLANT CIRCULATION - HIGH WATER LEVEL Attachment 1, Volume 14, Rev. 0, Page 64 of 138

, Volume 14, Rev. 0, Page 65 of 138 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) , Volume 14, Rev. 0, Page 65 of 138

Attachment 1, Volume 14, Rev. 0, Page 66 of 138 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 in operation-' MMO OPERABLE and I Erraated APPLICABILITY: MODE 6 when the water level above the top of the Vuel assemhL~seated within thelreactor pressure vessel is M02

> 23 feet. flange ACTION:

a. With less than one decay heat removal loop in operation, except as in -

ACTION A provided in b below, suspend all operations ,involving an increase e Iracordcay heat load orja reduction in boo ocnration of the Ratr Cool ant System./ LIose al I containmenlL penetrations provioin-g-\

fd-reLL acceýss Ir/ t e ot in e t a m s h re t h u sde atmospherq - L01 1within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .*Y*./

b.heat The deay *on Tor Up to one] 0 LCO 3.9.4 hour per 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months period during the performancofOR ALTERATIONS in the L- * "

Note ývicinit of the reactor pressure vessel (hot) legs. "*-M4 S

1. The prp,,isions of Specifi/extion 3.0.3 are/lot applicab~le.A00 SURVEILLANQF REnUIREMENTS 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 SR 3.9.4.1 decay heat removal loop to be in operation and circulating reactor coolant tnrougIn thne reactor core:

a. At a flow rate of 5 2800 gpm w enever a reductio in Reactor cOlant ystem oron oncen ration is being made.

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

< 140"F, provide no reduction in Reactor Coo ant System boron concentration being made.

Water of a lower boron concentration than the existing RCS M0S Required concentration may be added to the RCS, wit t, flowrate o" Action A,1 reactor coolan gh the ess t an ,O00 gpm, provide that the Doron concentration of 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 I 3/4 9-8 Amendment No. 88. 188 Page 1 of 1 Attachment 1, Volume 14, Rev. 0, Page 66 of 138

Attachment 1, Volume 14, Rev. 0, Page 67 of 138 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. 0, Page 67 of 138

Attachment 1, Volume 14, Rev. 0, Page 68 of 138 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. 0, Page 68 of 138

Attachment 1, Volume 14, Rev. 0, Page 69 of 138 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

< 1401F 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. 0, Page 69 of 138

Attachment 1, Volume 14, Rev. 0, Page 70 of 138 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. The Required Actions are consistent with the actions taken for containment closure in CTS 3.9.4 and ITS 3.9.3. This change is designated as less restrictive because less stringent Required Actions are being applied in the ITS than were applied in the CTS.

Davis-Besse Page 4 of 4 Attachment 1, Volume 14, Rev. 0, Page 70 of 138

Attachment 1, Volume 14, Rev. 0, Page 71 of 138 Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs)

Attachment 1, Volume 14, Rev. 0, Page 71 of 138

Attachment 1, Volume 14, Rev. 0, Page 72 of 138 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.

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 Systemýýith bron concentration less than that required to meet the minimum required boron 0

concentration of LCO 3 .9 .1ct , "Boron Concentration.'

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 Attachment 1, Volume 14, Rev. 0, Page 72 of 138

Attachment 1, Volume 14, Rev. 0, Page 73 of 138 CTS DHR and Coolant Circulation - High Water Level 3.9.4 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME A.4 Close equipment hatch and 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months Action a secure with aour bolts. 0 AND A.5 Close one door in each air 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months lock.

AND V A.62 j each penetration 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months providing direct access from the containment is either closed - atmos here to the outside atmosphere with a manual or automatic isolation valve, blind flange, or equivalen A.6. Veri each enetratio is capableo beingcose bby 0

an OPERABLE Containment Purge and Exhaust Isolation System.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY 4.8.9.1 SR 3.9.4.1 Verify one DHR loop is in operation and circulating reactor coolant at a flow rate of _[2800gpm.

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months 0

BWOG STS 3.9.4-2 Rev. 3.0, 03/31/04 Attachment 1, Volume 14, Rev. 0, Page 73 of 138

Attachment 1, Volume 14, Rev. 0, Page 74 of 138 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.

Davis-Besse Page 1 of 1 Attachment 1, Volume 14, Rev. 0, Page 74 of 138

Attachment 1, Volume 14, Rev. 0, Page 75 of 138 Improved Standard Technical Specifications (ISTS) Bases Markup and Justification for Deviations (JFDs)

Attachment 1, Volume 14, Rev. 0, Page 75 of 138

Attachment 1, Volume 14, Rev. 0, Page 76 of 138 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 in MODE 6 are to remove decay heat UFSAR, Appendix I and sensible heat from the Reactor Coolant System (RCS), as required 3D.1.30 (Ref. 1) by- 34, to provide mixing of borated coolant, to provide sufficient coolant circulation to minimize the effects of a boron dilution accident, and 0

Stoýpreveintboron stratification (ýRef. ). Heat is removed from the RcCSby circulating reactor coolant through the DHR heat ey)6hanger(s), where the cooer heat is transferred to the Component Cooling Water System -

lheat -e an er). The coolant is then returned to the RCS via the R S (core flood nozzles col - s. Operation of the DHR System for normal cooldown or decay heat removal is manually accomplished from the control room. The heat C) removal rate is adjusted by control of the flow of reactor coolant through coolers the DHR heat excbahnger(s) and byj:assing the heat_ ý(changer(s.

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 - 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 Attachment 1, Volume 14, Rev. 0, Page 76 of 138

Attachment 1, Volume 14, Rev. 0, Page 77 of 138 DHR andCoolant 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 loopis 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 heatsl.4ý 00

b. Mixing of borated coolant to minimize the possibility of criticalityaiqHD Ic. Indiotion of reacr coolant t94nperature. 0 An OPERABLE DHR loop includes a DHR pump, alheatqichange

0 since the DHR System is pathland to dltermine th Ylow end t-erature. The flow path starts in ,

a manually operated vavs piig'ntuetadcnrl oesr nOEAL lw0 system (i.e.. it is nnt lt oneE of the RCS hot legs and is returned to the IRCS cod legs core flood nozzles }.

sytei-..i_

automatically actuated). Additionally, each DHR loop is cons ered OPERABLE if it canb cay he manually aligned (remote or Iocalitthe shutdov coolin-ode Mr Iremoval o fecay heat* Operation of one subsystj*m can maintain the Ireactor coolant temperature as required.""

The LCO is modified by a Note that allows the 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 an 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months period, provided no operations are permitted that would dilute the RCS boron concentration by introduction of coolant into the RCS with boron concentration less than Boron required to meet the minimum boron concentration of LCO 3 .9 .1w Boron .Concentration."

concentration reduction with coolant at boron concentrations less than required to 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 heat is removed by natural convection to the large mass of water in the refueling Iana0 BVVOG STS B 3.9.4-2 Rev. 3.0, 03/31/04 Attachment 1, Volume 14, Rev. 0, Page 77 of 138

Attachment 1, Volume 14, Rev. 0, Page 78 of 138 DHR and Coolant Circulation - High Water Level B 3.9.4 BASES APPLICABILITY One DHR loop must be OPERABLE and in operation in MODE 6, with the water level a 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 movement in 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 Se ion 3.5 E r en C e Coolin sys[e. (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 LCO 3.9.5, "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. 0, Page 78 of 138

Attachment 1, Volume 14, Rev. 0, Page 79 of 138 DHR and Coolant Circulation - High Water Level B 3.9.4 BASES ACTIONS (continued)

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

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

a. The equipment hatch must be closed and secured with foul bolts,-,Ej 00

b. One door in each air lock must be closed*and

c. Each penetration providing direct access from the containment atmosphere to the outside atmosphere must be either closed by a manual or automatic isolation valve, blind flange, or equivalent, or 0

©D verified to be capable of being closed by an OPERABLE Containment Purge and Exhaust Isolation System. {_

Wth 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 can be closed so that the dose limits 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 is reasonable, 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 coolant. The flow rate is determined by the flow rate necessary to provide sufficient decay heat removal capability and to prevent thermal and 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 uDF7"--.FSAR, Section 1l. UFSAR, Appendix 3D.1.30. ] 000 BWOG STS B 3.9.4-4 Rev. 3.0, 03/31/04 Attachment 1, Volume 14, Rev. 0, Page 79 of 138

Attachment 1, Volume 14, Rev. 0, Page 80 of 138 B 3.9.4 0 INSERT 1 An OPERABLE Containment Purge and Exhaust Isolation System consists of an OPERABLE containment purge and exhaust noble gas monitor (LCO 3.3.15, "Containment Purge and Exhaust Isolation - High Radiation"), 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 OPERABLE containment purge and exhaust isolation valve in each penetration flow path, which is capable of being manually closed from the control room.

Insert Page B 3.9.4-4 Attachment 1, Volume 14, Rev. 0, Page 80 of 138

Attachment 1, Volume 14, Rev. 0, Page 81 of 138 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.

Davis-Besse Page 1 of 1 Attachment 1, Volume 14, Rev. 0, Page 81 of 138

Attachment 1, Volume 14, Rev. 0, Page 82 of 138 Specific No Significant Hazards Considerations (NSHCs)

Attachment 1, Volume 14, Rev. 0, Page 82 of 138

Attachment 1, Volume 14, Rev. 0, Page 83 of 138 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.

Davis-Besse Page 1 of 1 Attachment 1, Volume 14, Rev. 0, Page 83 of 138

Attachment 1, Volume 14, Rev. 0, Page 84 of 138 ATTACHMENT 5 ITS 3.9.5, DHR AND COOLANT CIRCULATION - LOW WATER LEVEL Attachment 1, Volume 14, Rev. 0, Page 84 of 138

, Volume 14, Rev. 0, Page 85 of 138 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) , Volume 14, Rev. 0, Page 85 of 138

Attachment 1, Volume 14, Rev. 0, Page 86 of 138 ITS 3.9.5 ITS REFLIEL ING OPERAT IONS LOtW NATER LEVEL LIMITING CONDITION FOR OPERATION and- oneloop shleinoperaton LA01 M0l LCO 3.9.5 3.9.8.2 Two jindýAndent]DHR loops shall be OPERABLE. Add proposed LCO NOTE1"A02 Add proposed LCO NOTE 2 APPLICABILITY: MODE Ifuel 6assemb~lies when the waterswtCled level withinabove the the top of reactor prsthe Vradiate./

vSse 1~~ is (0 k M01 less than 23 feet. fag ACTION:

ACTIONS A a. With less than the required DHR loops OPERABLE, immediately initiate and B corrective action to return the required loops to OPERABLE status as soon as possible. proposed Required Ation A.2 f'Add Add proposed Required Actions B'I, 8b.3.

Ad poosd euie AtinA. 8.4, and 8.5 for two inoperable i1oops A03 bTe proy-sions, of .Secif1 lofl J.U.J areR~appcicable.A0

-[Add proposed ACTION B for loop not in operation SURVEILLANCE REQUIREMENTS one DHR loop shall be determined to be in operation per Sped-SR 3.9.5.1 4.9.8.2 At least one DHR loop shall be determined to be in 'operation per Speci-fication 4.9.8.1. The inactivX loop shall be deterimined to be OPERABLE perL ispecificati/On 4.0.

Add proposed SR 3.9.5.2 M,,

J*The normal gy-emergency power sour-e may be inoperable aor eac DHR 1oop. A DAVIS-BESSE, UNIT 1. 3/4 9-8a Amendment No. 38 Page 1 of 2 Attachment 1, Volume 14, Rev. 0, Page 86 of 138

Attachment 1, Volume 14, Rev. 0, Page 87 of 138 ITS 3.9.5 ITS REFUELING OPERATIONS 3/4.9.8 DECAY HEAT REMOVAL AND COOLANT CIRCULATION ALL WJATER LEVELS LIMITING CONDITION FOR OPERATION 3.9.8.1 At least one decay heat removal loop shall be in operation.

APPLICABILITY: 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

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 of Specification 3.0.3 are not applicable.

F V r PFn hlT oFM rNTrr MII rIN SURVEIIIANCr RrInUTRPMrNTý SR 3.9.5.1 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 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 Sy tem boron concentrat on is being made .

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 restric ive reactivity condi ion specified in S ecification 3.9.1.

PAVIS-BESSE, UNIT I 3/4 9-8 AmendmenL No. *8,1B8 Page 2 of 2 Attachment 1, Volume 14, Rev. 0, Page 87 of 138

Attachment 1, Volume 14, Rev. 0, Page 88 of 138 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. 0, Page 88 of 138

Attachment 1, Volume 14, Rev. 0, Page 89 of 138 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 3.9.8.2 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 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 > 10°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 an OPERABLE 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. 0, Page 89 of 138

Attachment 1, Volume 14, Rev. 0, Page 90 of 138 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 an OPERABLE Containment Purge and Exhaust Isolation System (ITS 3.9.5 Required Action B.5). These actions assist in minimizing the consequences of 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. 0, Page 90 of 138

Attachment 1, Volume 14, Rev. 0, Page 91 of 138 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. 0, Page 91 of 138

Attachment 1, Volume 14, Rev. 0, Page 92 of 138 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 < 1401F 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. 0, Page 92 of 138

Attachment 1, Volume 14, Rev. 0, Page 93 of 138 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. 0, Page 93 of 138

Attachment 1, Volume 14, Rev. 0, Page 94 of 138 Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs)

Attachment 1, Volume 14, Rev. 0, Page 94 of 138

Attachment 1, Volume 14, Rev. 0, Page 95 of 138 CTS DHR and Coolant 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 be in operation.

.............................. IMU I- :t*,

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 temperature M.*; 0

b. No operations are permitted that would cause introduction of coolant into the Reactor Coolant Systemtvith boron 0

concentration less than that required to meet the minimum required boron concentration of LCO 3.9.1, ' * "Boron Concentration;" 00

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 < 23 ft 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 Attachment 1, Volume 14, Rev. 0, Page 95 of 138

Attachment 1, Volume 14, Rev. 0, Page 96 of 138 CTS DHR and Coolant Circulation - Low Water Level 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 DH R loop to OPERABLE status and to operation.

AND B.3 Close equipment hatch and 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months secure with jfoulbolts. 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 V.erify B.52 each penetration 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months providing direct access from the containment iseither closed - atmosphere to the outside atmosphere~with a manual 0 or automatic isolation valve, blind flange, or equiva from next page BWOG STS 3.9.5-2 Rev. 3.0, 03/31/04 Attachment 1, Volume 14, Rev. 0, Page 96 of 138

Attachment 1, Volume 14, Rev. 0, Page 97 of 138 CTS DHR and Coolant Circulation - Low Water Level 3.9.5 move to previous page ACTIONS (continued) A CONDITION REQUIRED ACTION COMPLETION TIME Action a, DOC M01 1B.5.2 Ve eifyach penetration is" 0

capable of being close by an OPERABLE Containment Purge and Exhaust Isolation System.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY 4.9.8.1, 4.9.8.2 SR 3.9.5.1 Verify one DHR loop is in operation. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months DOC M04 SR 3.9.5.2 rVerify correct breaker alignment and indicated 7 days Ppower available to the required DHR pump that is nnot in operation.

--NOTE -.------.-..-........---

L 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.

J

} 0 BWOG STS 3.9.5-3 Rev. 3.0, 03/31/04 Attachment 1, Volume 14, Rev. 0, Page 97 of 138

Attachment 1, Volume 14, Rev. 0, Page 98 of 138 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

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.

Davis-Besse Page 1 of 1 Attachment 1, Volume 14, Rev. 0, Page 98 of 138

Attachment 1, Volume 14, Rev. 0, Page 99 of 138 Improved Standard Technical Specifications (ISTS) Bases Markup and Justification for Deviations (JFDs)

Attachment 1, Volume 14, Rev. 0, Page 99 of 138

Attachment 1, Volume 14, Rev. 0, Page 100 of 138 DHR and Coolant Circulation - Low Water Level B 3.9.5 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, Appendix and sensible heat from the Reactor Coolant System (RCS), as required 3D.1.30 (Ref. 1 ib) , to provide mixing of borated coolant, to provide sufficient coolant circulation to minimize the effects of a boron dilution accident, and 1 W to prevent boron stratification (Ref. 7. Heat is removed from the RCS by circulating reactor coolant through the DHR heat e)*,dhanger(s,,where the .o .

0_____)

heat is transferred to the Component Cooling Water System core flood nozzles eat canger. The coolant is then returned to the RCS via the F cold es. peration of the DH R System for normal cooldown/decay 0 0 heat removal is manually accomplished from the control room. The heat removal rate is adjusted by control of the flow of reactor coolant through ce the DHR eat excla'nger(s) and bypassing the heat e~changer(s)l. Mixing n2 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 due to 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 addition of water to the reactorvessel 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 the integrity 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 In MODE 6, with the water level < 23 ft above the top of the reactor vessel independent flange, two DHR loops must be OPERABLE. Additionally, one DHR loop must be in operation to provide:

0

a. Removal of decay heats HI* 00

b. Mixing of borated coolant to minimize the possibility of criticalityJ- -.

0 Ic. Indidtion of reacterr coolant teyfiperature. I 0 BWOG STS B 3.9.5-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 14, Rev. 0, Page 100 of 138

Attachment 1, Volume 14, Rev. 0, Page 101 of 138 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 5 15 minutes when switching from one train to another. The circumstances for stopping both DHR pumps are to be limited to situations when the outage time is short and the core outlet 5 or temperature The Note prohibitsis maintained > 10 degrees F below saturation temperaturer boron dilutiorl"9*draining operations by introduction of 0

coolant into the RCS with boron concentrations less than required to meet the minimum boron concentration of LCO 3.9.1 when DHR forced flow is stopped. i Note 2 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 other loop is OPERABLE and in operation. Prior to declaring the loop inoperable, consideration should be given to the existing plant configuration. This consideration should inclýude týt

_core RCS time wvate tooil level isand short,_there is no/draining that the cap ility existsoperation to furthet ater to inject borated/ reduce into the reacr vessel. This permits surveillance tests to be performed on the inoperable loop during a time when these tests are safe and possible.

An OPERABLE DHR loop consists of a DHR pump, a heat chane, valves, piping, instruments, and controls to ensure an OPERABLE flow cooler Q path land to determine thyow end tefterature[ The flow path starts in one of the RCS hot legs and is returned to the C e. 0 Both DHR support pumps filling or/diningbethe aligned to the refueling or for Wat Refueling cavity Storageof Tank to p/formnance

~required testinV4.

APPLICABILITY Two DHR loops are 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 D System in other MODES are covered by LCOs in Section 3.4,;Reactor Coolant System (RCg. an ection 3 . er en ore oolin 000 se OCS. DHR loop requirements in MODE 6, with the water 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 a 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 3.9.5-2 Rev. 3.0, 03/31/04 Attachment 1, Volume 14, Rev. 0, Page 101 of 138

Attachment 1, Volume 14, Rev. 0, Page 102 of 138 B 3.9.5 (O INSERT 1 time to core boiling, potential for RCS draindown, and RCS makeup capability.

S 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. 0, Page 102 of 138

Attachment 1, Volume 14, Rev. 0, Page 103 of 138 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 water level.

B..1 If no DHR loop is in operation or no DHR loop is OPERABLE, 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.

B.2 If no DHR loop is in operation or no DHR loop is OPERABLE, actions shall be initiated immediatelyond continued ou erru ion 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 a 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 removalProcedures, should beor using thelcnarging Shigh pressure injection." implemented. This includes decay heat makeup, or other ' -safety injection pumlt through the Chemical/and Volume Control SystemI (

inLco s.u'ces, with 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 withrfourlIboltsfy 0 )

BWOG STS B 3.9.5-3 Rev. 3.0, 03/31/04 Attachment 1, Volume 14, Rev. 0, Page 103 of 138

Attachment 1, Volume 14, Rev. 0, Page 104 of 138 DHR and Coolant Circulation - Low Water Level B 3.9.5 BASES ACTIONS (continued)

b. One door in each air lock must be closed an.d 0

c. Each penetration providing direct access from the containment atmosphere to the outside atmosphere must be either closed by a manual or automatic isolation valve, blind flange, or equivalent, or verified to be capable of being closed by an OPERABLE Containment 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 can be closed so that the dose limits 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 is reasonable, based on the low probability of the coolant boiling in that time.

SURVEILLANCE SR 3.9.5.1 REQUIREMENTS This Surveillance demonstrates that one DHR loop is in operation. The flow rate 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 the water level in the vicinity of the reactor vessel nozzles, the DHR loop flow rate determination must also consider the DHR pump suction requirement.

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

SR 3.9.5.2 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 performed by verifying proper breaker alignment and power available to the required pump. The Frequency of 7 days is considered reasonable in view of other administrative controls available and has been shown to be ERT 4 acceptable by operating experience.

09 REFERENCES .FSAR, Section

l. UFSAR, Appendix 3D.1.30.) 0 00 BWOG STS B 3.9.5-4 Rev. 3.0, 03/31/04 Attachment 1, Volume 14, Rev. 0, Page 104 of 138

Attachment 1, Volume 14, Rev. 0, Page 105 of 138 B 3.9.5 O INSERT 3 An OPERABLE Containment Purge and Exhaust Isolation System consists of an OPERABLE containment purge and exhaust noble gas monitor (LCO 3.3.15, "Containment Purge and Exhaust Isolation - High Radiation"), 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 OPERABLE 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. 0, Page 105 of 138

Attachment 1, Volume 14, Rev. 0, Page 106 of 138 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.

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.

Davis-Besse Page 1 of 1 Attachment 1, Volume 14, Rev. 0, Page 106 of 138

Attachment 1, Volume 14, Rev. 0, Page 107 of 138 Specific No Significant Hazards Considerations (NSHCs)

Attachment 1, Volume 14, Rev. 0, Page 107 of 138

Attachment 1, Volume 14, Rev. 0, Page 108 of 138 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. 0, Page 108 of 138

, Volume 14, Rev. 0, Page 109 of 138 ATTACHMENT 6 ITS 3.9.6, REFUELING CANAL WATER LEVEL , Volume 14, Rev. 0, Page 109 of 138

, Volume 14, Rev. 0, Page 110 of 138 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) , Volume 14, Rev. 0, Page 110 of 138

Attachment 1, Volume 14, Rev. 0, Page 111 of 138 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 irradia d fuel assp blies sea)/ed within the reactor pressure L01 vessel'__ý

ýirra~dia Iedl APPLICABILITY: During movemen ofuel a ies o co ro r s A02 within Fthe reac r pressre vessel M ACTO.N: " containment ACTION A With the requirements of the above specification not satisfied, sus end ir--

all operation involving movement of fuel assemblies r ro ro -- 02 within~the re.actaE res sure vesse . line provisions of Sdecification 3.0.32 are not applicabMe.0 SA03 SURVEILLANCE REQUIREMENTS SR 3.9.6.1 4.9.10 The water level shall be determined to be at least its minimum L03 required depth Viithin 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .=ert*r to the start ot anolt least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months during movement of fuel assemblies rod within the reactor pressure vessel.

DAVIS-BESSE, UNIT 1 3/4 9-10 Page 1 of 1 Attachment 1, Volume 14, Rev. 0, Page 111 of 138

Attachment 1, Volume 15, Rev. 0, Page 112 of 138 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 15, Rev. 0, Page 112 of 138

Attachment 1, Volume 15, Rev. 0, Page 113 of 138 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 DOG L01. The change eliminating MODE 6 is discussed in DOG A02. The change eliminating control rods is discussed in DOG 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 DOG M01. This change is designated as more restrictive because it will prohibit operations that are not prohibited in the GTS.

RELOGATED SPECIFICATIONS None REMOVED DETAIL GHANGES 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 15, Rev. 0, Page 113 of 138

Attachment 1, Volume 15, Rev. 0, Page 114 of 138 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 ifthe 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 15, Rev. 0, Page 114 of 138

Attachment 1, Volume 15, Rev. 0, Page 115 of 138 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.

Davis-Besse Page 4 of 4 Attachment 1, Volume 15, Rev. 0, Page 115 of 138

Attachment 1, Volume 14, Rev. 0, Page 116 of 138 Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs)

Attachment 1, Volume 14, Rev. 0, Page 116 of 138

Attachment 1, Volume 14, Rev. 0, Page 117 of 138 Refueling Canal Water Level CTS 3.9.6 3.9 REFUELING OPERATIONS 3.9.6 Refueling Canal Water Level 3.9.10 LCO 3.9.6 Refueling canal water level shall be maintained .'-23 ft above the top of the reactor vessel flange.

APPLICABILITY: During movement of irradiated fuel assemblies within containment.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME Action A. Refueling p water level not within limit, A.1 Suspend movement of irradiated fuel assemblies Immediately 0

within containment.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY 4.9.10 SR 3.9.6.1 Verify refueling canal water level is >_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 3.9.6-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 14, Rev. 0, Page 117 of 138

Attachment 1, Volume 14, Rev. 0, Page 118 of 138 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. 0, Page 118 of 138

Attachment 1, Volume 14, Rev. 0, Page 119 of 138 Improved Standard Technical Specifications (ISTS) Bases Markup and Justification for Deviations (JFDs)

Attachment 1, Volume 14, Rev. 0, Page 119 of 138

Attachment 1, Volume 15, Rev. 0, Page 120 of 138 Refueling Canal Water Level B 3.9.6 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 of 23 ft above the top of the reactor vessel flange.,reactor vessel During refueling, this maintains sufficient water level in the n the refueling canal, the fuel transfer canal,1therefing cavityr 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. 1 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 nd the ueling cav 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 Reference 2. With a minimum water level of 23 ft, and a minimum decay time ot. hours 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 I canal I A minimum refueling-[ water level of 23 ft above the reactor vessel 0

flange is required to ensure that the radiological consequences of a 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 Attachment 1, Volume 15, Rev. 0, Page 120 of.138

Attachment 1, Volume 14, Rev. 0, Page 121 of 138 Refueling Canal Water Level B 3.9.6 BASES APPLICABILITY LCO 3.9.6 is applicable when moving irradiated fuel assemblies within the containment. The LCO minimizes the possibility of a fuel handling accident in containment that is beyond the assumptions of the safety analysis. If irradiated fuel is not present in containment, there can 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 ta 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 completion of movement of a component to a safe position.

SURVEILLANCE SR 3.9.6.1 REQUIREMENTS Verification of a minimum water level of 23 ft above the top of the reactor vessel flange ensures that the design basis for the postulated fuel handling accident analysis during refueling operations is met. Water at the required level above the top of the reactor vessel flange limits the consequences of damaged fuel rods that are postulated to result from a 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 volume of water and the normal procedural controls of valve positions, which make significant unplanned level changes unlikely.

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

-- 2..FSAR Section 00

3. 10CFR100.10.

BWOG STS B 3.9.6-2 Rev. 3.0, 03/31/04 Attachment 1, Volume 14, Rev. 0, Page 121 of 138

Attachment 1, Volume 14, Rev. 0, Page 122 of 138 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. 0, Page 122 of 138

Attachment 1, Volume 14, Rev. 0, Page 123 of 138 Specific No Significant Hazards Considerations (NSHCs)

Attachment 1, Volume 14, Rev. 0, Page 123 of 138

Attachment 1, Volume 14, Rev. 0, Page 124 of 138 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. 0, Page 124 of 138

, Volume 14, Rev. 0, Page 125 of 138 ATTACHMENT 7 RELOCATED/DELETED CURRENT TECHNICAL SPECIFICATIONS , Volume 14, Rev. 0, Page 125 of 138

, Volume 14, Rev. 0, Page 126 of 138 CTS 3/4.9.3, DECAY TIME , Volume 14, Rev. 0, Page 126 of 138

, Volume 14, Rev. 0, Page 127 of 138 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) , Volume 14, Rev. 0, Page 127 of 138

, Volume 14, Rev. 0, Page 128 of 138 CTS 3/4.9.3 Page 1 of 1 , Volume 14, Rev. 0, Page 128 of 138

Attachment 1, Volume 14, Rev. 0, Page 129 of 138 DISCUSSION OF CHANGES CTS 3/4.9.3, DECAY TIME ADMINISTRATIVE CHANGES None MORE RESTRICTIVE CHANGES None RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES LA01 (Type 4 - Removal of LCO, SR, or other TS requirementto the TRM, UFSAR, ODCM, QAPM, IST Program, or liP) CTS 3.9.3 requires the reactor to be subcritical for at least 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . The ITS does not include the requirements for decay time. This changes the CTS by moving the explicit decay time requirements from the Technical Specifications to the Technical Requirements Manual (TRM).

The removal of these details from the Technical Specification is acceptable because this type of information is not necessary to provide adequate protection of public health and safety. The purpose of CTS LCO 3.9.3 to ensure that sufficient time has elapsed to allow the radioactive decay of the short-lived fission products in the irradiated fuel consistent with the assumptions used in the fuel handling accident analysis. Although CTS 3.9.3 satisfies Criterion 2 of the Technical Specifications Selection Criteria in 10 CFR 50.36(c)(2)(ii) (for radioactive decay assumptions in the fuel handling accident), the requirements for decay time following subcriticality will always be met for a refueling outage because of the operations required prior to moving irradiated fuel in the reactor vessel (e.g., containment entry, removal of vessel head, removal of vessel internals, etc.) Also, this change is acceptable because the removed information will be adequately controlled in the TRM. The TRM is 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 a less restrictive removal of detail change because a requirement is being removed from the Technical Specifications.

LESS RESTRICTIVE CHANGES None Davis-Besse Page 1 of 1 Attachment 1, Volume 14, Rev. 0, Page 129 of 138

Attachment 1, Volume 14, Rev. 0, Page 130 of 138 Specific No Significant Hazards Considerations (NSHCs)

Attachment 1, Volume 14, Rev. 0, Page 130 of 138

Attachment 1, Volume 14, Rev. 0, Page 131 of 138 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3/4.9.3, DECAY TIME There are no specific NSHC discussions for this Specification.

Davis-Besse Page 1 of 1 Attachment 1, Volume 14, Rev. 0, Page 131 of 138

Attachment 1, Volume 14, Rev. 0, Page 132 of 138 CTS 3/4.9.6, FUEL HANDLING BRIDGE OPERABILITY Attachment 1, Volume 14, Rev. 0, Page 132 of 138

, Volume 14, Rev. 0, Page 133 of 138 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) , Volume 14, Rev. 0, Page 133 of 138

Attachment 1, Volume 14, Rev. 0, Page 134 of 138 CTS 3/4.9.6 REFUELIN__

G OPERATIONS .

FUEL HANDLING BRIDGE OPERABILITY LIMITING CONDITION FOR OPERATIO 3.9.6 The control rod hoist a d fuel assembly hoist of the fuel ha dung brdge shall be used for movem nt of control rods or fuel assemblie and shall be OPERABLE with:

a. The control rod hoi t having:

1. 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: During m vement of control rods or fuel assemb ies within the reactor pressure Yes 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 h 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 RE/UIR 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. U IT 1 3/4 9-6 Am ddment No. 135 Page 1 of 1 Attachment 1, Volume 14, Rev. 0, Page 134 of 138

Attachment 1, Volume 14, Rev. 0, Page 135 of 138 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. 0, Page 135 of 138

Attachment 1, Volume 14, Rev. 0, Page 136 of 138 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. 0, Page 136 of 138

Attachment 1, Volume 14, Rev. 0, Page 137 of 138 Specific No Significant Hazards Considerations (NSHCs)

Attachment 1, Volume 14, Rev. 0, Page 137 of 138

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

ML072200481

Text

Navigation menu

Search