Text

Volume 9, Revision 0, Davis-Besse, Unit 1 - Improved Technical Specifications Conversion, ITS Section 3.4 Reactor Coolant System.
	ML072200472
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Site:	Davis Besse
Issue date:	08/03/2007
From:	; FirstEnergy Nuclear Operating Co
To:	; Office of Nuclear Reactor Regulation
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Attachment 1, Volume 9, Rev. 0, Page 1 of 415 ATTACHMENT 1 VOLUME 9 DAVIS-BESSE IMPROVED TECHNICAL SPECIFICATIONS CONVERSION ITS SECTION 3.4 REACTOR COOLANT SYSTEM (RCS)

Revision 0 Attachment 1, Volume 9, Rev. 0, Page 1 of 415

Attachment 1, Volume 9, Rev. 0, Page 2 of 415 LIST OF ATTACHMENTS

1. ITS 3.4.1

2. ITS 3.4.2

3. ITS 3.4.3

4. ITS 3.4.4

5. ITS 3.4.5

6. ITS 3.4.6

7. ITS 3.4.7

8. ITS 3.4.8

9. ITS 3.4.9

10. ITS 3.4.10

11. ITS 3.4.11

12. ITS 3.4.12

13. ITS 3.4.13

14. ITS 3.4.14

15. ITS 3.4.15

16. ITS 3.4.16

17. ITS 3.4.17

18. Relocated Current Technical Specifications Attachment 1, Volume 9, Rev. 0, Page 2 of 415

Attachment 1, Volume 9, Rev. 0, Page 3 of 415 ATTACHMENT 1 ITS 3.4.1, RCS PRESSURE, TEMPERATURE, AND FLOW DEPARTURE FROM NUCLEATE BOILING (DNB) LIMITS Attachment 1, Volume 9, Rev. 0, Page 3 of 415

, Volume 9, Rev. 0, Page 4 of 415 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) , Volume 9, Rev. 0, Page 4 of 415

Attachment 1, Volume 9, Rev. 0, Page 5 of 415 ITS 3.4.1 ITS POWER DISTRIBUTION LIMITS DNB PARAMETERS LIMITING CONDITION FOR OPERATION_

LCO3.4.1 3.2.5 The following DNB related parameters shall be maintained within the limits shown on Table 3.2-2.

a. Reactor Coolant Hot Leg Temperature

b. Reactor Coolant Pressure

c. Reactor Coolant Flow Rate APPLICABILITY: MODE 1 ACTION:

ACTIONA -- ýI-f any parameter above exceeds its limit, restore the parameter to within its Llimit within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or reduce THERMAL POWER to less than 5% of RATED THERMAL ACTION B ER within the next hours. A02 SURVEILLANCE REQUIREMENTS SR 3.4.1.1, SR 3.4.1.2, 4.2.5.1 Each of the parameters of Table 3.2-2 shall be verified to be within SR 3.4.1.3 their limits at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

SR 3.4.1.4 .2"522 itsThelimit Reactor Coolant System total once flow per rate18 shall be determined to be Swithin by measurement at least months.

S---------------------------------NOTE---------------TE-------------------

Not required to be performed until 7 days after stable thermal conditions are established at _>70% RTP.

. . . . . -- L0e 2

DAVIS-BESSE, UNIT I 3/4 2-13 Amendment No. 64-,-1-23, 222 Page 1 of 2 Attachment 1, Volume 9, Rev. 0, Page 5 of 415

Attachment 1, Volume 9, Rev. 0, Page 6 of 415 0

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ý- 'c i 02 zxo6C 0C~O az Attachment 1, Volume 9, Rev. 0, Page 6 of 415

Attachment 1, Volume 9, Rev. 0, Page 7 of 415 DISCUSSION OF CHANGES ITS 3.4.1, RCS PRESSURE, TEMPERATURE, AND FLOW DEPARTURE FROM NUCLEATE BOILING (DNB) LIMITS 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 The CTS 3.2.5 Action requires the unit to reduce THERMAL POWER to "less than" 5% of RATED THERMAL POWER (RTP) within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months if the DNB parameters are not restored to within limit in 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months . ITS 3.4.1 ACTION B requires the power reduction to MODE 2, which is less than or equal to 5% RTP, within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ifthe DNB parameters are not restored to within limit in 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months . This changes the CTS by allowing the unit be at 5% RTP instead of < 5%

RTP. The change in the time period to reach 5% RTP is discussed in DOC L01.

This change is acceptable because it results in no technical change to the Technical Specifications. CTS 3.2.5 is applicable in MODE 1, which is greater than 5% RTP. CTS 3.0.1 (and ITS LCO 3.0.1) states that Actions are applicable during the MODES or other conditions specified for the Specification. Therefore, the CTS 3.2.5 Action to be less than 5% RTP ceases to be applicable once the unit enters MODE 2, i.e., at 5% RTP, and the Action is exited. As a result, changing the ACTION to be in "MODE 2" results in no operational difference from the CTS Action. This change is designated as administrative as it results in no technical change to the CTS.

A03 CTS 3.2.5, Table 3.2-2, Note (3) states that the minimum required Table 3.2-2 measured RCS flow rates include a flow rate uncertainty of 2.5%, "and are based on a minimum of 52 lumped burnable poison rod assemblies in place in the core." ITS 3.4.1 does not include the reason for the values of the measured RCS flow rate limits. This changes the CTS by deleting the specific reason for the measured RCS flow rate limit values. The change that moves the uncertainty value (2.5%) to the Bases is discussed in DOC LA01.

License Amendment 91 reduced the minimum RCS flow rate limits and as part of this amendment, added in the reason for the flow rate limits change (the limits were based on having a minimum of 64 lumped burnable poison rod assemblies). The reason was updated as part of License Amendment 135 and changed the number of assemblies to 52. For the current fuel cycle, while the flow rate limits have not been changed, Davis-Besse does not use lumped burnable poison rod assemblies. Therefore, the reason for the measured RCS flow rate limit values currently in the CTS is not correct. However, the basis of the RCS flow rate limits values is not needed to be included in the Technical Specifications to properly control the values - it is only information as to why the specific values were chosen. The ITS 3.4.1 Bases provides sufficient detail to explain the reason for the RCS flow rate limits. Therefore, not including the reason for the RCS flow rate limits has no impact of the technical requirements Davis-Besse Page 1 of 4 Attachment 1, Volume 9, Rev. 0, Page 7 of 415

Attachment 1, Volume 9, Rev. 0, Page 8 of 415 DISCUSSION OF CHANGES ITS 3.4.1, RCS PRESSURE, TEMPERATURE, AND FLOW DEPARTURE FROM NUCLEATE BOILING (DNB) LIMITS and therefore, the change is acceptable. This change is designated as administrative as it results in no technical change to the CTS.

MORE RESTRICTIVE CHANGES M01 CTS 3.2.5 requires that departure from nucleate boiling (DNB) parameters specified in CTS Table 3.2-2, including reactor coolant pressure, be maintained within specified limits. CTS Table 3.2-2 requires the measured reactor coolant system pressure to be > 2062.7 psig for four reactor coolant pump operation and

> 2058.7 psig for three reactor coolant pump operation. ITS LCO 3.4.1 .a requires RCS loop pressure be > 2064.8 psig for four reactor coolant pump operation and ITS LCO 3.4.1 .b requires RCS loop pressure be > 2060.8 psig for three reactor coolant pump operation. These values are also provided in ITS SR 3.4.1.1. This changes the CTS by increasing the DNB reactor coolant pressure parameter limits.

The limits on the DNB related parameters specified in CTS 3.2.5 assure that each of the parameters is maintained within the normal steady state envelope of operation assumed in the transient and accident analyses. The proposed ITS limits are consistent with the UFSAR initial assumptions and have been analytically demonstrated adequate to maintain a minimum DNB ratio greater than the minimum allowable DNB ratio throughout each analyzed transient. For the current and previous operating cycles, in order to offset the slight non-conservatism for the reactor coolant pressure parameter in the CTS, a DNB penalty has been assessed against the retained DNB margin in the reload licensing analyses. With implementation of the proposed values in the ITS, this offset will no longer be necessary for future core reload analyses. The proposed change is acceptable because it replaces the current CTS values with corrected values that are more conservative. This change is designated as more restrictive because more limiting DNB RCS loop pressure limits are required in the ITS than are required in the CTS.

RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES LA01 (Type I - Removing Details of System Design and System Description, Including Design Limits) CTS Table 3.2-2 Note (3) states, in part, that "These minimum required measured flows include a flow rate uncertainty of 2.5%." ITS 3.4.1 does not include this specific detail. The details of the Note are moved to the Bases of the applicable Surveillance, ITS SR 3.4.1.4. This changes the CTS by moving the details in CTS Table 3.2-2 Note (3) 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 Davis-Besse Page 2 of 4 Attachment 1, Volume 9, Rev. 0, Page 8 of 415

Attachment 1, Volume 9, Rev. 0, Page 9 of 415 DISCUSSION OF CHANGES ITS 3.4.1, RCS PRESSURE, TEMPERATURE, AND FLOW DEPARTURE FROM NUCLEATE BOILING (DNB) LIMITS protection of public health and safety. The ITS still retains the information and 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 3 - Relaxation of Completion Time) The CTS 3.2.5 Action requires the unit to reduce THERMAL POWER to < 5% of RTP within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months if the DNB parameters are not restored to within limit in 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months . ITS 3.4.1 ACTION B requires the power reduction to < 5% RTP (MODE 2) within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months if the DNB parameters are not restored to within limit in 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months . This changes the CTS by extending the time for the unit to be placed outside the Applicability of the Specification. The change in the THERMAL POWER value is discussed in DOC A02.

The purpose of the CTS 3.2.5 Action is to limit the time the unit can be outside of the DNB parameter limits and remain within the Applicability of the Specification.

This change is acceptable because the Completion Time is consistent with safe operation under the specified Condition, considering the OPERABLE status of the redundant systems or features. This includes the capacity and capability of remaining systems or features, a reasonable time for repairs or replacement, and the low probability of a DBA or transient occurring during the allowed Completion Time. The change extends the time from 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months that the unit is allowed to be outside the DNB parameter limits and be in the Applicability of the Specification. This change is designated as less restrictive because additional time is allowed to restore parameters to within the LCO limits than was allowed in the CTS.

L02 (Category 7- Relaxation of Surveillance Frequency - Non-24 Month Type Change) CTS 4.2.5.2 requires RCS total flow rate be determined to be within limits once per 18 months. ITS SR 3.4.1.4 requires the same Surveillance, but includes a Note to allow the performance to be delayed for up to 7 days after stable thermal conditions are established at > 70% RTP. This changes the CTS by delaying performance of the Surveillance until adequate conditions exist to perform the Surveillance.

The purpose of CTS 4.2.5.2 is to ensure the RCS total flow rate instrumentation is properly calibrated using a precision calorimetric heat balance. The change is acceptable because the new Surveillance Frequency continues to ensure a precision calorimetric heat balance is performed. This change delays the performance of the precision calorimetric heat balance for up to 7 days after stable thermal conditions are established at > 70% RTP. This change is necessary since a precision heat balance necessary to perform the proper calibration is not obtainable at low power conditions when thermal power is not stable (i.e., power or flow are changing). At low power conditions, the AT across Davis-Besse Page 3 of 4 Attachment 1, Volume 9, Rev. 0, Page 9 of 415

Attachment 1, Volume 9, Rev. 0, Page 10 of 415 DISCUSSION OF CHANGES ITS 3.4.1, RCS PRESSURE, TEMPERATURE, AND FLOW DEPARTURE FROM NUCLEATE BOILING (DNB) LIMITS the core will be too small to provide valid results. Furthermore, during this additional time period the RCS total flow is still required to be monitored by ITS SR 3.4.1.3, and the instrumentation used to perform this verification has been previously calibrated by the last performance of ITS SR 3.4.1.4. This change is designated as less restrictive because Surveillances can be performed less frequently under the ITS than in the CTS.

Davis-Besse Page 4 of 4 Attachment 1, Volume 9, Rev. 0, Page 10 of 415

Attachment 1, Volume 9, Rev. 0, Page 11 of 415 Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs)

Attachment 1, Volume 9, Rev. 0, Page 11 of 415

Attachment 1, Volume 9, Rev. 0, Page 12 of 415 CTS RCS Pressure, Temperature, and Flow DNB Limits 3.4.1 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.1 RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB)

Limits 3.2.5 LCO 3. 4.1 RCS DNB parameters for loop pressure, hot leg temperature, and RCS total flow rate shall be within the limits -specified below.

Table 3.2-2

a. With four reactor coolant DumDs (RCPs) operatina:

RCS loop pressure shall be > 2 1.6 psig, RCS hot leg

~00 gpm temperature shall be <

__[139.7/E6]Ibh and F, and RCS total flow rate shall be 6 10

b. With three RCPs operating:

2060.8 RCS loop pressure shall be _ 2 7.2 psig, RCS hot leg temperature shall be 6M4.6 0 F. and RCS total flow rate shall be1 M1O1D4,4E6] lb/h 61.

APPLICABILITY: MODE 1.

NOTI--

Table 3.2-2 RCS loop pressure limit does not apply during:

0 Note (2)

a. THERMAL POWER ramp > 5% RTP per minute or

b. THERMAL POWER step> 10% RTP.

0 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME 3.2.5 Action A. One or more RCS DNB A.1 Restore RCS DNB 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months parameters not within parameter(s) to within limit.

limits.

3.2.5 B. Required Action and B.1 Be in MODE 2. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months Action associated Completion Time not met.

BWOG STS 3.4.1-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 0, Page 12 of 415

Attachment 1, Volume 9, Rev. 0, Page 13 of 415 CTS RCS Pressure, Temperature, and Flow DNB Limits 3.4.1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.1.1 -. .-----------

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

Table 3.2-2 With three RCPs operating, the limits are applied to Note (1) the loop with two RCPs in operation.

4.2.5.1 Verify RCS loop pressure >_Ž 2 ".psig 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months 0 0

four RCPs operating or _ [ . psig with three RCPs operating.

Table 3.2-2 Note (1)

SR 3.4.1.2 TE...-...

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

With three RCPs operating, the limits are applied to the loop with two RCPs in operation.

0 4.2.5.1 Verify RCS hot leg temperature _ 46- F. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months 0 4.2.5.1 SR 3.4.1.3 Verify RCS total flow_> r[139.7/E6] Ib/hý with four 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months RCPs operating or >104.!ýE6] Ib/h with three 0D RCPý operating.

0 4.2.5.2 SR 3.4.1.4 ------ ------- NOTE -----------------

Only requir d to be perfor ed when sta le thermal conditions are established/in the higher rower range 0

of MODE 1. / /

Verify RCS total flow rate is within limit by [1l8months measurement.

Not required to be performed until 7 days after stable thermal conditions are established at _ 70% RTP.

I BWOG STS 3.4.1-2 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 0, Page 13 of 415

Attachment 1, Volume 9, Rev. 0, Page 14 of 415 JUSTIFICATION FOR DEVIATIONS ITS 3.4.1, RCS PRESSURE, TEMPERATURE, AND FLOW DEPARTURE FROM NUCLEATE BOILING (DNB) LIMITS

1. Brackets have been removed and the proper plant specific information/value has been provided.

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

3. Typographical error corrected.

4. The ISTS SR 3.4.1.4 Note currently requires performance of the SR immediately upon establishing stable conditions in the higher power range. The proposed change removes the ambiguity of "higher power range" by using a specific power level requirement. Also, as described in ISTS Section 1.4, Example 1.4-5, the wording of the Note regarding stable thermal conditions means that it must be completed when stable conditions are established. No time is provided after the establishment of stable conditions. The Note has been revised to allow some time after the "stable thermal conditions are established in the higher power range of MODE 1" to actually perform the measurement. Therefore, the Note is revised to allow 7 days after stable thermal conditions are established at _Ž 70% RTP. This is consistent with the current manner in which Davis-Besse performs the Surveillance, since it provides the necessary time to allow test procedure completion and calculation verifications.

Davis-Besse Page 1 of 1 Attachment 1, Volume 9, Rev. 0, Page 14 of 415

Attachment 1, Volume 9, Rev. 0, Page 15 of 415 Improved Standard Technical Specifications (ISTS) Bases Markup and Justification for Deviations (JFDs)

Attachment 1, Volume 9, Rev. 0, Page 15 of 415

Attachment 1, Volume 9, Rev. 0, Page 16 of 415 I

All changes are 0 unless otherwise noted 9 RCS .Pressure, Temperature, and Flow DNB Limits B 3.4.1 B 3.4 REACTOR COOLANT SYSTEM (RCS)

B 3.4.1 RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits BASES BACKGROUND These Bases address requirements for maintaining RCS pressure, temperature, and flow rate within limits assumed in the safety analyses.

The safety analyses (Ref. 1) of normal operating conditions and anticipated operational occurrences assume initial conditions within the normal steady state envelope. The limits placed on DNB related parameters ensure that these parameters will not be less conservative than were assumed in the analyses and thereby provide assurance that the minimum departure from nucleate boiling ratio (DNBR) will meet the required criteria for each of the transients analyzed.

The LCO for minimum RCS pressure is consistent with operation within the nominal operating envelope andlis above-that used asIthe initial corresponds to pressure in the analyses. A pressure greater than the minimum specified will produce a higher minimum DNBR. A pressure lower than the minimum specified will cause the plant to approach the DNB limit.

TThe LCO for maximum RCS coolant hot leg temperature is consistent with full power operation within the nominal operating envelope and[p ow tanthe initial hot leg temperature in the analyses. A hot leg temperature lower than that specified will produce a higher minimum DNBR. A temperature higher than that specified will cause the plant to approach the DNB limit.

The RCS flow rate is not expected to vary during operation with all pumps running. The LCO for the minimum RCS flow rate corresponds to that assumed for the DNBR analyses. A higher RCS flow rate will produce a higher DNBR. A lower RCS flow will cause the plant to approach the DN B limit.

APPLICABLE The requirements of LCO 3.4.1 represent the initial conditions for DNB SAFETY limited transients analyzed in the plant safety analyses (Ref. 1). The ANALYSES safety analyses have shown that transients initiated from the limits of this for the current reload cycle (Ref. 2) 1 LCO will meet the DNBR criterion *of1.3] This is the acceptance limit for the RCS DNBR parameters. Changes to the facility that could impact these parameters must be assessed for their impact on the DNBR 0

BWOG STS B 3.4.1-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 0, Page 16 of 415

Attachment 1, Volume 9, Rev. 0, Page 17 of 415 All changes are 1 unless otherwise noted 9 RCS Pressure, Temperature, and Flow DNB Limits B 3.4.1 BASES APPLICABLE SAFETY ANALYSES (continued) criterion. The transients analyzed for include loss of coolant flow events and dropped or stuck control rod events. A key assumption for the analysis of these events is that the core power distribution is within the limits of LCO 3.2.1, "RegulatingRod Inserion Limits," LCO 3.2.3, "AXIAL POWER IMBALANCE C L fPI ,"and LCO 3.2.4, "QUADRANT POWER TILTFRT 0Q inal The core outlet pressure assumed in the safety anal ses ism[ ps-ia.O The minimum pressure specified in LCO 3.4.1 is the Ii it alue in the reactor coolant loop as measured at the hot leg pressure tap. cor The safety analyses are performed with an assumed RCS coolant average temperature oLj*F j(579*F plus 2"F wance for culational to limit the range of allowable, steady) Func rtainty . The ý2rre :on din hot leg temperature of6

/state operation, consistent with the _,calculated by a!ýSuming an RCS core outlet pressure of 2;135 psia and an linitial conditions assumed inthse IRCS flow rate Of 374,880 gpm . The -maximumn temperature specified is N-related accident analyse the imt value at the hot leg resistance temperature detector.

The safety analyses are performed with an assumed RCS flow rate of

-380,o000--4*- gpm. The minimum flow rate specified in LCO 3.4.1 is the minimum mass flow rate. including a 2.5% uncertainty euvlent Analyses have been performed to establish the pressure, temperature, and flow rate requirements for three pump and four pump operation. The flow limits for three pump operation are substantially lower than for four pump operation. To meet the DNBR criterion, a corresponding maximum power limit is required (see Bases for LCO 3.4.4, "RCS Loops - MODES 1 and 2").

The RCS DNB limits satisfy Criterion 2 of 10 CFR 50.36(c)(2)(ii).

LCO This LCO specifies limits on the monitored process variables: RCS loop (hot leg) pressure, RCS hot leg temperature, and RCS total flow rate to ensure that the core operates within the limits assumed for the plant safety analyses. Operating within these limits will result in meeting DNBR criteria in the event of a DNB limited transient.

The pressure and temperature limits are to be applied to the loop with two reactor coolant pumps (RCPs) running for the three RCPs operating condition.

BVVOG STS B 3.4.1-2 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 0, Page 17 of 415

Attachment 1, Volume 9, Rev. 0, Page 18 of 415 All changes are unless otherwise noted RCS Pressure, Temperature, and Flow DNB Limits B 3.4.1 BASES LCO (continued) measured values and are The LCO numerical values for pressure, temperature, and flow rate are "

given for the measurement location bu ave not been adjuste/for instrumenterror. Plant specific limit"f instrument error are /stabli Zby.the plant stff to meet the operafional requirements of t LCO.

APPLICABILITY In MODE 1, the limits on RCS pressure, RCS hot leg temperature, and RCS flow rate must be maintained during steady state with four pump or three pump operation in order to ensure that DNBR criteria will be met in the event of an unplanned loss of forced coolant flow or other DNB limited transient. In all other MODES the power level is low enough so that DNB is not a concern.

The Note indicates the limit on RCS pressure may be exceeded during short term operational transients such as a THERMAL POWER ramp increase > 5% RTP per minute or a THERMAL POWER step increase

> 10% RTP. These conditions represent short term perturbations where actions to control pressure variations might be counterproductive. Also, since they represent transients initiated from power levels < 100% RTP, increased DNBR margin exists to offset the temporary pressure variations.

Another set of limits on DNBR related parameters is provided in Safety Limit (SL) 2.1.1, "Reactor Core SLs." Those limits are less restrictive than the limits of LCO 3.4.1, but violation of an SL merits a stricter, more severe Required Action. Should a violation of LCO 3.4.1 occur, the operator must check whether an SL may have been exceeded.

ACTIONS A.1 Loop pressure and hot leg coolant temperature are controllable and measurable parameters. With one or both of these parameters not within the LCO limits, action must be taken to restore the parameters. RCS flow rate is not a controllable parameter and is not expected to vary during steady state four pump or three pump operation. However, if the flow rate is below the LCO limit, the parameter must be restored to within limits or power must be reduced as required in Required Action B.1, to restore DNBR margin and eliminate the potential for violation of the accident analysis bounds.

The 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months Completion Time for restoration of the parameters provides sufficient time to adjust plant parameters, determine the cause for the off normal condition, and restore the readings within limits. The Completion Time is based on plant operating experience.

BWOG STS B 3.4.1-3 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 0, Page 18 of 415

Attachment 1, Volume 9, Rev. 0, Page 19 of 415 All changes are Q"]

unless otherwise noted J RCS Pressure, Temperature, and Flow DNB Limits B 3-4.1 BASES ACTIONS (continued)

B.1 If the Required Action A-1 is not met within the Completion Time, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 2 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . In MODE 2, the reduced power condition eliminates the potential for violation of the accident analysis bounds.

The 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months Completion Time is reasonable, based on operating experience, to reduce power in an orderly manner in conjunction with even control of steam generator heat removal.

SURVEILLANCE SR 3.4.1.1 REQUIREMENTS Since RequiredAction A.1 allows a Completion Time of 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months to restore parameters that are not within limits, the 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Surveillance Frequency for loop (hot leg) pressure is sufficient to ensure that the pressure can be restored to a normal operation, steady state condition following load changes and other expected transient operations. The RCS pressure value specified is dependent on the number of pumps in operation and has been adjusted to account for the pressure loss difference between 2200 the core exit and the measurement location. The value used in the plant Ssafety analysis i235 psi. The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months interval has been shown by (nominal) operating practice to be sufficient to regularly-assess potential degradation and to verify operation is within safety analysis assumptions.

A Note has been added to indicate the pressure limits are to be applied to the loop with two pumps in operation for the three pump operating condition.

SR 3.4.1.2 Since Required Action A.1 allows a Completion Time of 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months to restore parameters that are not within limits, the 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Surveillance Frequency for hot leg temperature is sufficient to ensure that the RCS coolant temperature can be restored to a normal operation, steady state condition following load changes and other expected transient operations. The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months interval has been shown by operating practice to be sufficient to regularly assess potential degradation and to verify that operation is within safety analysis assumptions.

BVVOG STS B 3.4.1-4 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 0, Page 19 of 415

Attachment 1, Volume 9, Rev. 0, Page 20 of 415 I All changes are 1 u

unless otherwise noted

,RCS Pressure, Temperature, and Flow DNB Limits B 3.4.1 BASES SURVEILLANCE REQUIREMENTS (continued)

A Note has been added to indicate the temperature limits are to be applied to the loop with two pumps in operation for the three pump operating condition.

SR 3.4.1.3 The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Surveillance Frequency for RCS total flow rate is performed using the installed flow instrumentation. The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months interval has been shown by operating practice to be sufficient to regularly assess potential degradation and to verify that operation is within safety analysis assumptions.

SR 3.4.1.4 Measurement of RCS total flow rate by performance of a precision calorimetric heat balance once every.V81months allows the installed RCS flow instrumentation to be calibrated and verifies that the actual RCS flow is greater than or equal to the minimum required RCS flow rate. I INSERT 2 The Frequency of 18 months refle s the importance of v ifying flow after a refueli routage when the 96re has been altered oy RCS flow characteristW5 may have been rndified, which may havgycaused chang 0

10 l w INSERT 3J The Surveillance is modified by a Note that indicates the SR~do=s not rýC need to b/e/erformed until stable tll~lrmal conditions are established at* -

lhigher~ower levels, [The Note is necessary to allow measurement of the flow rate at normal operating conditions at power in MODE 1. The Surveillance cannot be performed at low power or in MODE 2 or below because at low power the AT across the core will be too small to provide valid results.

REFERENCES ID --. FSAR, S 00

2. UFSAR, Appendix 4B 0 BWOG STS B 3.4.1-5 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 0, Page 20 of 415

Attachment 1, Volume 9, Rev. 0, Page 21 of 415 B 3.4.1 0 INSERT I These minimum required measured flows include a flow rate uncertainty of 2.5%.

0 INSERT 2 is considered adequate for ensuring accurate RCS flow measurement instrumentation and has been shown by operating experience to be acceptable.

0 INSERT 3 is not required to be performed until 7 days after stable thermal conditions are established at > 70% RTP.

Insert Page B 3.4.1-5 Attachment 1, Volume 9, Rev. 0, Page 21 of 415

Attachment 1, Volume 9, Rev. 0, Page 22 of 415 JUSTIFICATION FOR DEVIATIONS ITS 3.4.1 BASES, RCS PRESSURE, TEMPERATURE, AND FLOW DEPARTURE FROM NUCLEATE BOILING (DNB) LIMITS

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. Typographical error corrected.

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

5. Editorial change made for clarity.

Davis-Besse Page 1 of 1 Attachment 1, Volume 9, Rev. 0, Page 22 of 415

Attachment 1, Volume 9, Rev. 0, Page 23 of 415 Specific No Significant Hazards Considerations (NSHCs)

Attachment 1, Volume 9, Rev. 0, Page 23 of 415

Attachment 1, Volume 9, Rev. 0, Page 24 of 415 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.4.1, RCS PRESSURE, TEMPERATURE, AND FLOW DEPARTURE FROM NUCLEATE BOILING (DNB) LIMITS There are no specific NSHC discussions for this Specification.

Davis-Besse Page 1 of 1 Attachment 1, Volume 9, Rev. 0, Page 24 of 415

Attachment 1, Volume 9, Rev. 0, Page 25 of 415 ATTACHMENT 2 ITS 3.4.2, RCS MINIMUM TEMPERATURE FOR CRITICALITY Attachment 1, Volume 9, Rev. 0, Page 25 of 415

, Volume 9, Rev. 0, Page 26 of 415 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) , Volume 9, Rev. 0, Page 26 of 415

Attachment 1, Volume 9, Rev. 0, Page 27 of 415 ITS 3.4.2 ITS 1jREACTIVITY CONTROL SYSTEMS MINIMUM TEMPERATURE FOR CRITICALITY LIMITING CONDITION FOR OPERATION LCO 3.4.2 3.1.1.4 The Reactor Coolant System lowest loop temperature (T avg) shall be : 525F.av APPLICABILITY: MODES I and 2*.

ACTION:

With a Reactor Coolant System loop 1:iperaeure ACTION A IT ithin its mit witin minutes orl be TA T within MODE 2 with keff <1.0 A03 SURVEILLANCE REOUIREMENTS SR 3.4.2.1 4.1.1.4 The RCS temperature (Tavg ) shall be determined to be > 525 0 F:

a. Within 15 min gs prior to achieving actor criticality, a d

b. At the least onc olant Reactor/ per t30System minutesTvg whenIst esstan5 e reactor is °.critical / and La

-- every" 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Applicability With Keff >_1.0.

DAVIS-BESSE, UNIT I 3/4 1-5 1 of 1 Attachment 1, Volume 9, Rev. 0, Page 27 of 415

Attachment 1, Volume 9, Rev. 0, Page 28 of 415 DISCUSSION OF CHANGES ITS 3.4.2, RCS MINIMUM TEMPERATURE FOR CRITICALITY 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 The CTS 3.1.1.4 Action states that with a Reactor Coolant System (RCS) operating loop temperature (Tavg) < 525°F, to "restore Tavg to within its limit within 15 minutes or be in HOT STANDBY within the next 15 minutes." ITS 3.4.2 ACTION A states that with Tavg in one or more RCS loops not within limit, be in MODE 2 with keff < 1.0 within 30 minutes. This changes the CTS by eliminating the redundant and unnecessary requirement to restore Tavg to within its limit within 15 minutes. The change associated with entering MODE 2 with keff < 1.0 instead of HOT STANDBY is discussed in DOC A03.

This change is acceptable because it results in no technical change to the Technical Specifications. Although the CTS 3.1.1.4 Action allows only 15 minutes to restore the parameter to within the limit, it actually allows the entire 30 minutes to either restore the parameter or to be in HOT STANDBY (essentially outside the Applicability of CTS 3.1.1.4). In addition, the CTS 3.1.1.4 Action only requires actual steps to begin reducing reactor power at the beginning of the last 15 minutes of the 30-minute time period. However, CTS 3.0.2 states that "In the event the Limiting Condition for Operation is restored prior to expiration of the specified time interval, completion of the ACTION Statement is not required."

Therefore, for this specific case, if the parameter is restored between 15 minutes and 30 minutes after the Limiting Condition for Operation (LCO) parameter is not met, completion of the CTS 3.1.1.4 Action to be in HOT STANDBY is not required. Thus, 30 minutes is essentially allowed for either the parameter to be restored to within limit or the unit to be in HOT STANDBY (i.e., only one of the two CTS Actions must be met within 30 minutes). The CTS 3.0.2 requirement is retained in ITS LCO 3.0.2. Therefore, this change does not expand the total time interval allowed to restore the parameter, as a 30-minute time period is already essentially allowed by the CTS. This change is designated as administrative as it results in no technical change to the CTS.

A03 The CTS 3.1.1.4 Action states that with a Reactor Coolant System operating loop temperature (Tavg) < 525°F, to restore Tavg to within its limit within 15 minutes or be in "HOT STANDBY" within the next 15 minutes. ITS 3.4.2 ACTION A states that with Tavg in one or more RCS loops not within limit, be in "MODE 2 with keff < 1.0" within 30 minutes. This changes the CTS by requiring entry into MODE 2 with ke, < 1.0 instead of entry into HOT STANDBY (MODE 3). The change associated with the time to be in HOT STANDBY is discussed in DOC A02.

Davis-Besse Page 1 of 3 Attachment 1, Volume 9, Rev. 0, Page 28 of 415

Attachment 1, Volume 9, Rev. 0, Page 29 of 415 DISCUSSION OF CHANGES ITS 3.4.2, RCS MINIMUM TEMPERATURE FOR CRITICALITY This change is acceptable because it results in no technical change to the Technical Specifications. CTS 3.1.1.4 is applicable in MODE 1 and MODE 2 with kef - 1.0. CTS 3.0.1 (and ITS LCO 3.0.1) states that Actions are applicable during the MODES or other conditions specified for the Specification. Therefore, the CTS 3.1.1.4 Action to enter HOT STANDBY (MODE 3) ceases to be applicable once the unit enters MODE 2 with kff < 1.0. As a result, changing the ACTION to "be in MODE 2 with kf < 1.0" results in no operational difference from the CTS Action. This change is designated as administrative as it results in no technical change to the CTS.

MORE RESTRICTIVE CHANGES None RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES None LESS RESTRICTIVE CHANGES L01 (Category 7- Relaxation Of Surveillance Frequency, Non-24 Month Type Change) CTS 4.1.1.4 states that the RCS Tavg shall be determined to be > 5251F within 15 minutes prior to achieving reactor criticality, and every 30 minutes when the reactor is critical and the RCS T2 vg < 5300 F. ITS SR 3.4.2.1 requires RCS Tavg in each loop to be verified > 5251F every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . This changes the CTS by deleting the within 15 minutes prior to achieving criticality Frequency and the Surveillance Frequencies based on the condition of the reactor (critical) and reactor coolant temperature (< 530*F), and replacing them with a periodic 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Frequency.

The purpose of CTS 4.1.1.4 is to ensure RCS Tavg is within limit when the reactor is critical. The requirement is that RCS Tavg be > 5251F, and it is required to be met when the unit is operating in MODE 2 with keff > 1.0 and MODE 1. Based on ITS SR 3.0.4, this would require the SR to be met within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months prior to entry into MODE 2 with keff > 1.0 (i.e., before the reactor is critical). This change is acceptable because the new Surveillance Frequency provides an acceptable level of assurance that the RCS Tavg is within limit. The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Frequency is considered frequent enough to prevent inadvertent violation of the LCO. In the approach to criticality, with the required reactor coolant pumps running, the RCS is at normal operating pressure, so the conditions before and after criticality are similar. The approach to criticality is a carefully controlled evolution during which RCS temperature is closely monitored. Therefore, 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is frequent enough Davis-Besse Page 2 of 3 Attachment 1, Volume 9, Rev. 0, Page 29 of 415

Attachment 1, Volume 9, Rev. 0, Page 30 of 415 DISCUSSION OF CHANGES ITS 3.4.2, RCS MINIMUM TEMPERATURE FOR CRITICALITY for the Technical Specifications to require recording of Tavg prior to criticality given that it is being routinely monitored. This change is designated as less restrictive because Surveillances will be performed less frequently under the ITS than under the CTS.

Davis-Besse Page 3 of 3 Attachment 1, Volume 9, Rev. 0, Page 30 of 415

Attachment 1, Volume 9, Rev. 0, Page 31 of 415 Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs)

Attachment 1, Volume 9, Rev. 0, Page 31 of 415

Attachment 1, Volume 9, Rev. 0, Page 32 of 415

. CTS RCS Minimum Temperature for Criticality 3.4.2 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.2 RCS Minimum Temperature for Criticality LCO 3.1.1.4 LCO 3.4.2 Each RCS loop average temperature (Tavg) shall be _>525 0 F.

APPLICABILITY: MODE 1, MODE 2 with ke, ->1.0.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME 3.1.1.4 Action A. Tavg in one or more RCS loops not within limit.

A.1 Be in MODE 2 with[i

< 1.0.

30 minutes 0

SURVEILLANCE REQUIREMENTS SURVEILLANCE r FREQUENCY 4.1.1.4 SR 3.4.2.1 Verify RCS Ta,, in each loop >_5250F. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months BVOG STS 3.4.2-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 0, Page 32 of 415

Attachment 1, Volume 9, Rev. 0, Page 33 of 415 JUSTIFICATION FOR DEVIATIONS ITS 3.4.2, RCS MINIMUM TEMPERATURE FOR CRITICALITY

1. Typographical error corrected.

Davis-Besse Page 1 of 1 Attachment 1, Volume 9, Rev. 0, Page 33 of 415

Attachment 1, Volume 9, Rev. 0, Page 34 of 415 Improved Standard Technical Specifications (ISTS) Bases Markup and Justification for Deviations (JFDs)

Attachment 1, Volume 9, Rev. 0, Page 34 of 415

Attachment 1, Volum 9, Rev. 0, Page 35 of 415 S Alcanges are

0 K unless otherwise noted J RCS Minimum Temperature for Criticality B 3.4.2 B 3.4 REACTOR COOLANT SYSTEM (RCS)

B 3.4.2 RCS Minimum Temperature for Criticality BASES BACKGROUND Establishing the value for the minimum temperature for reactor criticality is based upon considerations for:

a. Operation within the existing instrumentation ranges and accuracies.._-

and Li 0

b. Operation with reactor vessel above its minimum nil ductility reference temperature when the reactor is critical.

The reactor coolant moderator temperature coefficient used in core operating and accident analysis is typically defined for the normal 5operating temperature range (532 F toLVA7F). The Reactor Protection System (RPS) receives inputs from the narrow range hot leg temperature detectors, which have a range of 520°F to 620'F. The integrated control system controls average temperature (Tavg) using inputs of the same range. Nominal TaLg for making the reactor critical is 532°F. Safety and joperating analyses for lower temperatures have not been made.

APPLICABLE There are no accident analyses that dictate the minimum SAFETY for criticality b Utll low power safety ar lyses assume ini ANALYSES temperatur near the 525 F limit (Rex 1+ 41 heRCS minimum temperature for criticality satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii).

Lmuch 7 LCO The purpose of the LCOi pre ent criticality outside theOnormal operating regime (532°F and to prevent operation in an unanalyzed condition.

The LCO limit of 5250 F has been selected to be within the instrument indicating range (520*F to 620*F). The limit is also set slightly below the lowest power range operating temperature (5320 F).

APPLICABILITY The reactor has been designed and analyzed to be critical in MODES 1 and 2 only and in accordance with this Specification. Criticality is not permitted in any other MODE. Therefore, this LCO is applicable in MODE 1 and MODE 2 when kef ý 1.0.

BWOG STS B 3.4.2-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 0, Page 35 of 415

Attachment 1, Volume 9, Rev. 0, Page 36 of 415 B 3.4.2 0 INSERT 1 Compliance with the LCO ensures that the reactor will not be made or maintained critical at a temperature significantly less than the hot zero power (HZP) temperature, which is assumed in the safety analysis (Ref. 1). Failure to meet the requirements of this LCO may produce initial conditions inconsistent with the initial conditions assumed in the safety analysis.

Insert Page B 3.4.2-1 Attachment 1, Volume 9, Rev. 0, Page 36 of 415

Attachment 1, Volume 9, Rev. 0, Page 37 of 415 RCS Minimum Temperature for Criticality B 3.4.2 BASES ACTIONS A.1 With Tvg below 5250 F, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to MODE 2 with *, < 1.0 in 30 minutes. Rapid reactor shutdown can be readily and practically achieved in a 30 minute period. The Completion Time reflects the ability to perform this 4ction and maintain the plnt 0

within the analyzed range. IfTv9 can be restored within the 30 minute time period, shutdown is not required.

SURVEILLANCE SR 3.4.2.1 REQUIREMENTS RCS loop average temperature is required to be verified at or above 525°F every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . The SR to verify RCS loop average temperatures every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months takes into account indications and alarms that are continuously available to the operator in the control room and is consistent with other routine Surveillances which are typically performed once per shift. In addition, operators are trained to be sensitive to RCS temperature during approach to criticality and will ensure that the minimum temperature for criticality is met as criticality is approached.

REFERENCES E31.'FSAR, r[ Sectin 15.2.1 ] 00 BWOG STS B 3.4.2-2 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 0, Page 37 of 415

Attachment 1, Volume 9, Rev. 0, Page 38 of 415 JUSTIFICATION FOR DEVIATIONS ITS 3.4.2 BASES, RCS MINIMUM TEMPERATURE FOR CRITICALITY

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

4. Typographical error corrected.

Davis-Besse Page 1 of 1 Attachment 1, Volume 9, Rev. 0, Page 38 of 415

Attachment 1, Volume 9, Rev. 0, Page 39 of 415 Specific No Significant Hazards Considerations (NSHCs)

Attachment 1, Volume 9, Rev. 0, Page 39 of 415

Attachment 1, Volume 9, Rev. 0, Page 40 of 415 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.4.2, RCS MINIMUM TEMPERATURE FOR CRITICALITY There are no specific NSHC discussions for this Specification.

Davis-Besse Page 1 of 1 Attachment 1, Volume 9, Rev. 0, Page 40 of 415

Attachment 1, Volume 9, Rev. 0, Page 41 of 415 ATTACHMENT 3 ITS 3.4.3, RCS PRESSURE AND TEMPERATURE (P/T) LIMITS Attachment 1, Volume 9, Rev. 0, Page 41 of 415

, Volume 9, Rev. 0, Page 42 of 415 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) , Volume 9, Rev. 0, Page 42 of 415

Attachment 1, Volume 9, Rev. 0, Page 43 of 415 ITS 0 ITS 3.4.3 REACTOR COOLANT SYSTEM 314.4.9 PRESSURE/TEMPERATURE LIMITS REACTOR COOLANT SYSTEM LIMITING CONDITION FOR OPERATION LCO 3.4.3

a. A maximum heat' of 50 0 F in any one hour Seriod, and

b. A maximum cooalown of 100*F in any one h/r period with cold leg temperatur > 270 0 F and a maximum co Idown of 50*F in any one hour perio with cold leg temperatur <2700F.

APPLICABILITY: At all times. A03 ACTION: Add proposed Conditions Aand C Notes With any of the above limits exceeded, restore the temperature and/or ACTIONS A pressure to within the limits within L30 minute ,perzorm n eng$neering SevaIuaton ofto tb/e atermne Reactor the efects System; Coolant o Laedetermine ou-o-imt the Reacto on the tht condition AcinA2an .

Aio.anC.

0 and C llintegrity operat:ion or be in at Completion Times/

oolant:

C System'remains acceptable fo-rcontinued and be in COýLD ýSH=UTDOWN within B* lIea&st. ROT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ACTION -the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

_EURIETLAE SUVI SR 3.4.3. 4.4.9.1.1 The Reactor Coolant System temperature and pressure shall be determined to be within the limits at least once per 30 minutes during system heatup, cooldown, and inservice leak and hydrostatic testing operations.

4.4.9.1.2 The reactor vessel material irradiation ýurveillance specimens representative of the essel materials shall be re ved and examined, to determine changes in terial properties, at the i tervals defined in BAW 1543A. The results oa these examinations shall be/used to update Figures I 3.4-2, 3.4-3 and 3/.4 DAVIS-BESSE, UNIT I 3/4 4-24 Amendment No. 8$, 116 Page 1 of 5 Attachment 1, Volume 9, Rev. 0, Page 43 of 415

Attachment 1, Volume 9, Rev. 0, Page 44 of 415 LO

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Attachment 1, Volume 9, Rev. 0, Page 47 of 415 ITS 3.4.3 Table 4.4-5 Reactor Vessel Mat rial Irradiation Surveillan e Scbedule D3LETED DAVIS-BESSE, UNIT 1 3/4 4-28 Amendment No. $1, 116 Page 5 of 5 Attachment 1, Volume 9, Rev. 0, Page 47 of 415

Attachment 1, Volume 9, Rev. 0, Page 48 of 415 DISCUSSION OF CHANGES ITS 3.4.3, RCS PRESSURE AND TEMPERATURE (PIT) LIMITS 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 3.4.9.1 states that the RCS temperature and pressure shall be limited "during heatup, cooldown, criticality, and inservice leak and hydrostatic testing."

CTS 3.4.9.1 is applicable at all times. ITS 3.4.3 states that the RCS pressure, RCS temperature, and RCS heatup and cooldown rates shall be maintained.

ITS 3.4.3 is applicable at all times. This changes the CTS by eliminating the LCO requirements that the limits must be met only during heatup, cooldown, criticality, and inservice leak and hydrostatic testing.

This change is acceptable because the CTS and ITS limits, including heatup, cooldown, criticality, and inservice leak and hydrostatic testing, are applicable at all times. Stating that the limits are applicable during heatup, cooldown, criticality, and inservice leak and hydrostatic testing in the LCO presents an apparent conflict with the Applicability which states that the limits apply at all times. This change is designated as administrative as it is an editorial change to eliminate an apparent conflict in the CTS.

A03 CTS 3.4.9.1 Action states that with any of the P/T limits exceeded, restore the temperature and/or pressure to within the limit within 30 minutes, perform an engineering evaluation to determine the effects of the out-of-limit condition on the integrity of the Reactor Coolant System, and determine that the Reactor Coolant System remains acceptable for continued operation. ITS 3.4.3 Conditions A and C are modified by a Note that requires the determination that the RCS is acceptable for continued operation be performed whenever the Condition is entered. This changes the CTS by explicitly stating that a determination that the RCS is acceptable for continued operation must be performed whenever the Condition is entered.

This change is acceptable because it is the current understanding and application of the CTS Action. The CTS 3.4.9.1 Action is currently interpreted as requiring a determination that the RCS is acceptable for continued operation whenever the LCO is not met. This change is designated as editorial as it clarifies the current understanding of the CTS requirement.

A04 CTS 3.4.9.1 Action states, in part, that with any of the P/T limits exceeded, restore the temperature and/or pressure to within the limit within 30 minutes.

ITS 3.4.3 ACTION C states that with the requirements of the LCO not met any time other than MODE 1, 2, 3, or 4, to immediately initiate action to restore the parameter(s) to within limits. This changes the CTS by requiring immediate action to restore P/T limits and continuing the action until complete, when the unit is in other than MODE 1, 2, 3, or 4.

Davis-Besse Page 1 of 4 Attachment 1, Volume 9, Rev. 0, Page 48 of 415

Attachment 1, Volume 9, Rev. 0, Page 49 of 415 DISCUSSION OF CHANGES ITS 3.4.3, RCS PRESSURE AND TEMPERATURE (P/T) LIMITS This change is acceptable because this change reflects an enhanced presentation of the existing intent. The CTS 3.4.9.1 Action to "restore... within 30 minutes" is proposed to be revised to "initiate action to restore... Immediately" for conditions other than MODES 1, 2, 3, and 4. This existing Action would appear to provide a half hour in which pressure and temperature requirements could exceed the limits, even it capable of being returned to within limits. Also, if the parameters are incapable of being restored within the limits within 30 minutes, the existing Action would appear to result in the requirement of a Licensee Event Report, since no additional Actions apply (the unit is already in MODE 5 or below). The intent of the Action is believed to be more appropriately presented in ITS 3.4.3 Required Action C.1. This interpretation of the intent is supported by the Babcock and Wilcox Standard Technical Specifications, NUREG-1430, Rev 3.1. This change is designated as administrative as it reflects an enhanced presentation of the existing intent.

A05 CTS 4.4.9.1.2 states that the reactor vessel material irradiation surveillance specimens representative of the vessel materials shall be removed and examined to determine changes in material properties, at the intervals defined in BAW 1543A. The results of these examinations shall be used to update Figures 3.4-1, 3.4-3, and 3.4-4. ITS 3.4.3 does not contain this Surveillance nor the Table. This changes the CTS by deleting the reactor vessel material irradiation Surveillance Requirement.

This change is acceptable because the Surveillance is unnecessary and repetitive. The unit is required by applicable regulations to remove material irradiation surveillance specimens and generate P/T curves in accordance with 10 CFR 50, Appendix H. Therefore, the Surveillance serves no purpose and is removed. This change is designated as administrative as it eliminates a requirement that is duplicative of a regulatory requirement in the CFR.

MORE RESTRICTIVE CHANGES M01 CTS 3.4.9.1 Action states that if the P/T limits are exceeded, an analysis must be performed and a determination made that the RCS remains acceptable for continued operation. No time limit is given for the performance of this analysis and determination. ITS 3.4.3 Required Action A.2 states that when the LCO is not met in MODES 1, 2, 3, or 4, determination is required that the RCS is acceptable for continued operation within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . ITS 3.4.3 Required Action C.2 states that when the LCO is not met any time other than in MODES 1, 2, 3, or 4, determination is required that the RCS is acceptable for continued operation prior to entering MODE 4. This changes the CTS by specifying a finite time to perform the determination.

This change is acceptable because it provides adequate time to evaluate exceeding the LCO requirements. The Completion Time of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months is considered reasonable for operation in MODES 1, 2, 3, and 4 because P/T limits are based on very conservative flaw assumptions and large factors of safety.

The Completion Time of "prior to entering MODE 4" during operations other than MODE 1, 2, 3, or 4 is considered reasonable since it would prevent entry into the operating MODES, and is consistent with LCO. 3.0.4. This change is designated Davis-Besse Page 2 of 4 Attachment 1, Volume 9, Rev. 0, Page 49 of 415

Attachment 1, Volume 9, Rev. 0, Page 50 of 415 DISCUSSION OF CHANGES ITS 3.4.3, RCS PRESSURE AND TEMPERATURE (PIT) LIMITS as more restrictive as it provides a limited time to perform an action for which the CTS provides not time limit.

RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES LA01 (Type I - Removing Details of System Design and System Description, Including Design Limits) CTS 3.4.9.1 states that the RCS (except the pressurizer) temperature and pressure shall be limited. The LCO also contains limits on RCS heatup and cooldown rates. ITS 3.4.3 states that the RCS pressure, RCS temperature, and RCS heatup and cooldown rates shall be maintained within limits. This changes the CTS by moving the exclusion of the pressurizer from the LCO 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 P/T limits on the RCS.

Neither the CTS or the ITS P/T limits apply to the pressurizer. It is the ITS convention to state this detail of the LCO in the ITS Bases. This detail of the LCO is not required to be in the Technical Specifications in order to provide adequate protection of the public health and safety. Also this change is acceptable because the removed information will be adequately controlled in the ITS Bases. 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.

LA02 (Type 4 - Removal of LCO, SR, or other TS requirement to the TRM, UFSAR, ODCM, QAPM, IST Program,PTLR, or liP) CTS 3.4.9.1 states, in part, that the Reactor Coolant system temperature and pressure shall be limited in accordance with the limits lines shown on Figures 3.4-2, 3.4-3, and 3.4-4. Additionally, CTS 3.4.9.1 .a and 3.4.9.1 .b specify the maximum heatup rate and the maximum cooldown rates, respectively. ITS 3.4.3 states that the RCS pressure, RCS temperature, and RCS heatup and cooldown rate shall be maintained within the limits specified in the PTLR. This changes the CTS by relocating the Figures and the maximum heatup and maximum cooldown rates to the PTLR.

The removal of these figures, heatup rate, and cooldown rate from the Technical Specification to the PTLR is acceptable because the PTLR is developed and utilized under NRC-approved methodologies, which will ensure that the RCS pressure and temperature limits for heatup, cooldown, low temperature operation, criticality, and hydrostatic testing as well as heatup and cooldown rates are met. This type of information is not necessary to be included in the Technical Specification to provide adequate protection of public health and safety. The ITS still retains the RCS P/T Limit requirements. The methodologies used to develop the parameters in the PTLR have obtained prior approval by the Davis-Besse Page 3 of 4 Attachment 1, Volume 9, Rev. 0, Page 50 of 415

Attachment 1, Volume 9, Rev. 0, Page 51 of 415 DISCUSSION OF CHANGES ITS 3.4.3, RCS PRESSURE AND TEMPERATURE (P/T) LIMITS NRC. Also, this change is acceptable because the removed information will be adequately controlled in the PTLR under the requirements provided in ITS 5.6.4, "Reactor Coolant System (RCS) PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)." ITS 5.6.4 ensures that the applicable RCS pressure and temperature limits are met. This change is designated as a less restrictive removal of detail change because the detailed P/T limits are being removed from the Technical Specifications.

LA03 (Type 3 - Removing ProceduralDetails for Meeting TS Requirements and Related Reporting Problems) CTS 3.4.9.1 Action states that with any P/T limits exceeded, to perform an engineering evaluation to determine the effects of the out-of-limit condition on the integrity of the RCS. ITS 3.4.3 ACTIONS A and C, in part, state that with the requirements of the LCO not met, to determine the RCS is acceptable for continued operation. The specific requirement to perform an engineering evaluation is not included in ITS 3.4.3. This changes the CTS by moving the requirement to "perform an engineering evaluation" to determine the effects of the out-of-limit condition on the integrity of the RCS to the Bases.

The removal of these details for performing actions 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 determine that the RCS remains acceptable for continued operation and this detail of how the determination is made is not required to be in the Technical Specifications in order to provide adequate protection of the public health and safety. The requirement to perform an engineering evaluation to determine the effects of the out-of-limit condition on the integrity of the RCS is a step in determining that the RCS remains acceptable for continued operation. Therefore, this detail on how the determination is made is moved to the Bases, which provides additional detail on how to the determination should be made. Also, this change is acceptable because these types of procedural details will be adequately controlled in the ITS Bases. This change is designated a less restrictive removal of detail change because procedural details for meeting Technical Specification requirements are being removed from the Technical Specifications.

LESS RESTRICTIVE CHANGES None Davis-Besse Page 4 of 4 Attachment 1, Volume 9, Rev. 0, Page 51 of 415

Attachment 1, Volume 9, Rev. 0, Page 52 of 415 Markup Improved Standard Technical Specifications (ISTS) and Justification for Deviations (JFDs)

Attachment 1, Volume 9, Rev. 0, Page 52 of 415

Attachment 1, Volume 9, Rev. 0, Page 53 of 415 CTS RCS PIT Limits 3.4.3 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.3 RCS Pressure and Temperature (P/T) Limits 3.4.9.1 LCO 3.4.3 RCS pressure, RCS temperature, and RCS heatup and cooldown rates shall be maintained within the limits specified in the PTLR.

APPLICABILITY: At all times.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME Action A. ----- NOTE--..... A.1 Restore parameter(s) to 30 minutes Required Action A.2 within limits.

shall be completed whenever this Condition AND is entered.

---- A.2 Determine RCS is 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months acceptable for continued Requirements of operation.

LCO not met in MODE 1,2, 3, or 4.

Action B. Required Action and B.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time of Condition A not AND met.

B.2 Be in MODE 5. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months Action C. - ----------- NOTE---------- C.1 Initiate action to restore Immediately Required Action C.2 parameter(s) to within limit.

shall be completed whenever this Condition AND is entered.

C.2 Determine RCS is Prior to entering acceptable for continued MODE 4 Requirements of operation.

LCO not met in other than MODE 1, 2, 3, or 4.

BVWOG STS 3.4.3-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 0, Page 53 of 415

Attachment 1, Volume 9, Rev. 0, Page 54 of 415 CTS RCS PIT Limits 3.4.3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY 4.4.9.1.1 SR 3.4.3.1 ------- NOTE ---.......----- - .-------

Only required to be performed during RCS heatup and cooldown operations and RCS inservice leak and hydrostatic testing.

Verify RCS pressure, RCS temperature, and RCS 30 minutes heatup and cooldown rates are within the limits specified in the PTLR.

BWOG STS 3.4.3-2 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 0, Page 54 of 415

Attachment 1, Volume 9, Rev. 0, Page 55 of 415 JUSTIFICATION FOR DEVIATIONS ITS 3.4.3, RCS PRESSURE AND TEMPERATURE (P/T) LIMITS None Davis-Besse Page 1 of 1 Attachment 1, Volume 9, Rev. 0, Page 55 of 415

Attachment 1, Volume 9, Rev. 0, Page 56 of 415 Improved Standard Technical Specifications (ISTS) Bases Markup and Justification for Deviations (JFDs)

Attachment 1, Volume 9, Rev. 0, Page 56 of 415

Attachment 1, Volume 9, Rev. 0, Page 57 of 415 RCS PIT Limits B 3.4.3 B 3.4 REACTOR COOLANT SYSTEM (RCS)

B 3.4.3 RCS Pressure and Temperature (P/T) Limits BASES BACKGROUND All components of the RCS are designed to withstand effects of cyclic loads due to system pressure and temperature changes. These loads are introduced by startup (heatup) and shutdown (cooldown) operations, power transients, and reactor trips. This LCO limits the pressure and temperature changes during RCS heatup and cooldown, within the design assumptions and the stress limits for cyclic operation.

The PTLR contains P/T limit curves for heatup, cooldown,4and inservice crtica-ity,0 leak and hydrostatic (ISLH) testing, and data for the maximum rate of change of reactor coolant temperature (Ref. 1).

Each PIT limit curve defines an acceptable region for normal operation.

The usual use of the curves is operational guidance during heatup or cooldown maneuvering, when pressure and temperature indications are monitored and compared to the applicable curve to determine that operation is within the allowable region.

The LCO establishes operating limits that provide a margin to brittle failure of the reactor vessel and piping of the reactor coolant pressure boundary (RCPB). The vessel is the component most subject to brittle failure, and the LCO limits apply mainly to the vessel. The limits do not apply to the pressurizer, which has different design characteristics and operating functions.

10 CFR 50, Appendix G (Ref. 2), requires the establishment of PIT limits for material fracture toughness requirements of the RCPB materials.

Reference 2 requires an adequate margin to brittle failure during normal operation, anticipated operational occurrences, and system hydrostatic tests. It mandates the use of the American Society of Mechanical Engineers (ASME), Boiler and Pressure Vessel Code,Section III, Appendix G (Ref. 3).

Linear elastic fracture mechanics (LEFM) methodology is used to determine the stresses and material toughness at locations within the RCPB. The LEFM methodology follows the guidance given by 10 CFR 50, Appendix G; ASME Code,Section III, Appendix G; and Regulatory Guide 1.99 (Ref. 4).

BWOG STS B 3.4.3-1 Rev. 3.0, 03/31104 Attachment 1, Volume 9, Rev. 0, Page 57 of 415

Attachment 1, Volume 9, Rev. 0, Page 58 of 415 RCS PIT Limits B 3.4.3 BASES BACKGROUND '(continued)

Material toughness properties of the ferritic materials of the -reactor vessel are determined in accordance with the NRC Standard Review Plan

,(Ref. 5), ASTM E 185 (Ref. 6), and additional reactor vessel requirements. These properties are then evaluated in-accordance with Reference 3.

The actual shift in the nil ductility reference temperature (RTNDT) of the vessel material will be established periodically by removing and evaluating the irradiated reactor vessel material specimens, in accordance with ASTM E 185 (Ref. 6) and Appendix H of 10 CFR 50 (Ref. 7). The operating P/T limit curves will be adjusted, as necessary, based on the evaluation findings and the recommendations of Reference 3.

The PIT limit curves are composite curves established by superimposing limits derived from stress analyses of those portions of the reactor vessel and head that are the most restrictive. At any specific pressure, temperature, and temperature rate of change, one location within the reactor vessel will dictate the most restrictive limit. Across the span of the P/T limit curves, different locations are more restrictive, and, thus, the curves are composites of the most restrictive regions.

The heatup curve represents a different set of restrictions than the cooldown curve because the directions of the thermal gradients through the vessel wall are reversed. The thermal gradient reversal alters the location of the tensile stress between the outer and inner walls.

The calculation to generate the ISLH testing curve uses different safety factors (per Ref. 3) than the heatup and cooldown curves. The ISLH testing curve also extends to the RCS design pressure of 2500f (.

The PIT limit curves and associated temperature rate of change limits are developed in conjunction with stress analyses for large numbers of operating cycles and provide conservative margins to nonductile failure.

Although created to provide limits for these specific normal operations, the curves also can be used to determine if an evaluation is necessary for an abnormal transient.

BWOG STS B 3.4.3-2 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 0, Page 58 of 415

Attachment 1, Volume 9, Rev. 0, Page 59 of 415 RCS P/T Limits B 3.4.3 BASES BACKGROUND (continued)

The consequence of violating the LCO limits is that the RCS has been operated under conditions that can result in brittle failure of the RCPB, possibly leading to a nonisolable leak or loss of coolant accident. In the event these limits are exceeded, an evaluation must be performed to determine the effect on the structural integrity of the RCPB components.

The ASME Code, Section Xl, Appendix E (Ref. 8) provides a recommended methodology for evaluating an operating event that causes an excursion outside the limits.

APPLICABLE The P/T limits are not derived from Design Basis Accident (DBA)

SAFETY analyses. They are prescribed during normal operation to avoid ANALYSES encountering pressure, temperature, and temperature rate of change conditions that might cause undetected flaws to propagate and cause nonductile failure of the RCPB, an unanalyzed condition. Reference 1 establishes the methodology for determining the P/T limits. Since the P/T limits are not derived from any DBA analysis, there are no acceptance limits related to the P/T limits. Rather, the P/T limits are acceptance limits themselves since they preclude operation in an unanalyzed condition.

RCS PIT limits satisfy Criterion 2 of 10 CFR 50.36(c)(2)(ii).

LCO The two elements of this LCO are:

a. The limit curves for heatup, cooldownand ISLH testing and

b. Limits on the rate of change of temperature.

The LCO limits apply to all components of the RCS, except the pressurizer. These limits define allowable operating regions and permit a large number of operating cycles while providing a wide margin to nonductile failure.

The limits for the rate of change of temperature control the thermal gradient through the vessel wall and are used as inputs for calculating the heatup, cooldown, and ISLH P/T limit curves. Thus, the LCO for the rate of change of temperature restricts stresses caused by thermal gradients and also ensures the validity of the PIT limit curves.

Violating the LCO limits places the reactor vessel outside of the bounds of the stress analyses and can increase stresses in other RCPB components. The consequences depend on several factors, as follows:

BVWOG STS B 3.4.3-3 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 0, Page 59 of 415

Attachment 1, Volume 9, Rev. 0, Page 60 of 415 RCS P/T Limits B 3.4.3 BASES LCO (continued)

a. The severity of the departure fromthe allowable operating P/T regime or the severity of the rate of change of temperature* . -

b. The length of time the limits were violated (longer violations allow the 0

temperature gradient in the thick vessel walls to become more pronounced d 0

c. The existences, sizes, and orientations of flaws in the vessel material.

APPLICABILITY The RCS P/T limits Specification provides a definition of acceptable operation for prevention of nonductile failure in accordance with 10 CFR 50, Appendix G (Ref. 2). Although the P/T limits were developed to provide guidance for operation during heatup or cooldown (MODES 3, 4, and 5) or ISLH testing, their applicability is at all times in keeping with the concern for nonductile failure. The limits do not apply to the pressurizer.

During MODES 1 and 2, other Technical Specifications provide limits for operation that can be more restrictive than or can supplement these PIT limits. LCO 3.4.1, "RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits," LCO 3.4.2, "RCS Minimum Temperature for Criticality," and Safety Limit (SL) 2.1, "SLs," also provide operational restrictions for pressure and temperature and maximum pressure. MODES 1 and 2 are above the temperature range of concern for nonductile failure, and stress analyses have been performed for normal maneuvering profiles, such as power ascension or descent.

ACTIONS A.1 and A.2 Operation outside the P/T limits during MODE 1, 2, 3, or 4 must be corrected so that the RCPB is returned to a condition that has been verified by stress analyses.

The 30 minute Completion Time reflects the urgency of restoring the parameters to within the analyzed range. Most violations will not be severe, and the activity can be accomplished in this time in a controlled manner.

BVWOG STS B 3.4.3-4 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 0, Page 60 of 415

Attachment 1, Volume 9, Rev. 0, Page 61 of 415 RCS P/T Limits B 3.4.3 BASES ACTIONS (continued) egei 0 Besides restoring operation to within limits, an valuation is required to determine if RCS operation can continue. The evaluation must verify the RCPB integrity remains~acceptable and must be completed e re 1continuirna*operation* Several methods may be used, including within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Q, comparison with pre-analyzed transients in the stress analyses, new analyses, or inspection of the components. The evaluation must be completed, documented, and approved in accordance with established plant procedures and administrative controls.

ASME Code, Section Xl, Appendix E (Ref. 8) may be used to support the evaluation. However, its use is restricted to evaluation of the vessel beltline. The evaluation must extend to all components of the RCPB.

may De The 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Completion ime is reas a to accomplish the evaluation.

The evaluation for a mild violatio-it] possible within this time, but more 0 severe violations may require special, event specific stress analyses or inspections. JA favorable evalpation must be complet,&d before continuingI 0 Condition A is modified by a Note requiring Required Action A.2 to be completed whenever the Condition is entered. The Note emphasizes the need to perform the evaluation of the effects of the excursion outside the allowable limits. Restoration alone per Required Action A.1 is insufficient because higher than analyzed stresses may have occurred and may have affected the RCPB integrity.

B.1 and B.2 2

ED--i lRequired Action and associated Completion Time of Condition A@

0 not met, the plant must be brought to a lower MODE because: (a) the RCS remained in an unacceptable pressure and temperature region for an extended period of increased stress, or (b) a sufficiently severe event caused entry into an unacceptable region. Either possibility indicates a need for more careful examination of the event, best accomplished with the RCS at reduced pressure and temperature. With reduced pressure and temperature conditions, the possibility of propagation of undetected flaws is decreased.

If the required restoration activity cannot be accomplished within 30 minutes, Required Action B.1 and Required Action B.2 must be implemented to reduce pressure and temperature.

BWVOG STS B 3.4.3-5 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 0, Page 61 of 415

Attachment 1, Volume 9, Rev. 0, Page 62 of 415 RCS PIT Limits B 3.4.3 BASES ACTIONS (continued)

If the required evaluation for continued operation cannot be accomplished within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , or the results are indeterminate or unfavorable, action must proceed to reduce pressure and temperature as specified in Required Actions B.1 and B.2. A favorable evaluation must be completed and documented before returning to operating pressure and temperature conditions. However, if the favorable evaluation is accomplished while reducing pressure and temperature conditions, a return to power - S operation may be considered without completing Required Actior*B 2.2 t n d Pressure and temperature are reduced by bringing the plant to MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and to MODE 5 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . The allowed Completion Times are reasonable, based on operating experience, to reach the required MODE from full power conditions in an orderly manner and without challenging plant systems.

C.1 and C.2 Actions must be initiated immediately to correct operation outside of the P/T limits at times other than MODE 1, 2, 3, or 4, so that the RCPB is returned to a condition that has been verified acceptable by stress analysis.

The immediate Completion Time reflects the urgency of initiating action to restore the parameters to within the analyzed range. Most violations will not be severe, and the activity can be accomplished within this time in a controlled manner.

In addition to restoring operation to within limits, an evaluation is required to determine if RCS operation can continue. The evaluation must verify that the RCPB integrity remains acceptable and must be completed prior to entry into MODE 4. Several methods may be used, including comparison with pre-analyzed transients in the stress analysis, or inspection of the components.

ASME Code,Section XI, Appendix E (Ref. 8), may also be used to support the evaluation. However, its use is restricted to evaluation of the vessel beltline.

BWAOG STS B 3.4.3-6 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 0, Page 62 of 415

Attachment 1, Volume 9, Rev. 0, Page 63 of 415 RCS PIT Limits B 3.4.3 BASES ACTIONS (continued)

Condition C is modified by a Note requiring Required Action C.2 to be completed whenever the Condition is entered. The Note emphasizes the need to perform the evaluation of the effects of the excursion outside the allowable limits. Restoration alone, per Required Action C.1, is insufficient because higher than analyzed stresses may have occurred and may have affected RCPB integrity.

SURVEILLANCE SR 3.4.3.1 REQUIREMENTS Verification that operation is within the PTLR limits is required every 30 minutes when RCS pressure and temperature conditions are undergoing planned changes.

This Frequency is considered reasonable in view of the control room indication available to monitor RCS status. Also, since temperature rate of change limits are specified in hourly increments, 30 minutes permits assessment and correction for minor deviations within a reasonable time.

Surveillance for heatup, cooldown, or ISLH testing may be discontinued when the definition given in the relevant plant procedure for ending the activity is satisfied.

This SR is modified by a Note that requires this SR to be performed only during system heatup, cooldown, and ISLH testing.

REFERENCES 1. BAW-10046A, Rev. 1, July 1977.

2. 10 CFR 50, Appendix G.

3. ASME, Boiler and Pressure Vessel Code,Section III, Appendix G.

4. Regulatory Guide 1.99, Revision 2, May 1988.

5. NUREG-0800, Section 5.3.1, Rev. 1, July 1981.

6. ASTM E 185-82, July 1982.

7. 10 CFR 50, Appendix H.

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

BVVOG STS B 3.4.3-7 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 0, Page 63 of 415

Attachment 1, Volume 9, Rev. 0, Page 64 of 415 JUSTIFICATION FOR DEVIATIONS ITS 3.4.3 BASES, RCS PRESSURE AND TEMPERATURE (P/T) LIMITS

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. 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.

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

4. Editorial change.

Davis-Besse Page 1 of 1 Attachment 1, Volume 9, Rev. 0, Page 64 of 415

Attachment 1, Volume 9, Rev. 0, Page 65 of 415 Specific No Significant Hazards Considerations (NSHCs)

Attachment 1, Volume 9, Rev. 0, Page 65 of 415

Attachment 1, Volume 9, Rev. 0, Page 66 of 415 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.4.3, RCS PRESSURE AND TEMPERATURE (PIT) LIMITS There are no specific NSHC discussions for this Specification.

Davis-Besse Page 1 of 1 Attachment 1, Volume 9, Rev. 0, Page 66 of 415

, Volume 9, Rev. 0, Page 67 of 415 ATTACHMENT 4 ITS 3.4.4, RCS LOOPS - MODES 1 AND 2 , Volume 9, Rev. 0, Page 67 of 415

, Volume 9, Rev. 0, Page 68 of 415 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) , Volume 9, Rev. 0, Page 68 of 415

Attachment 1, Volume 9, Rev. 0, Page 69 of 415 ITS 3.4.4 ITS 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 COOLANT LOOPS AND COOLANT CIRCULATION STARTUP AND POWER OPERATION LIMITING CONDITION FOR OPERATION LCO 3.4.4.a 3.4.1.1 Both reactor coolant loops and both reactor coolant pumps in each loop shall be in operation.

APPLICABILITY: MODES 1 and 2Q S0 ACTION:

LCO 3.4.4.b a.- Fwith*-one reactor O1PERATION may be coolant pump not in operation, STARTUP and POWER initiated and may proceed provided THEMPA POWER I.ls rotrieted to less than 80.6%of RATED THEMRA POWER Fa-nd-wiffhln]

10 horl e setpoints for the following trips have been reduced in EL01 accor ance with Specification 2.2.1 for operation with three reactor ACTION A coolant pumps operating:

1. High Flux LCO 3.4.4.b

2. Flux-AFlux-Flow Add proposed ACTION B ]M0 SURVEILLANCE REQUIREMENTS SR 3.4.4.1 4.4.1.1.1 The above required reactor coolant loops shall be verified to be in operation Land circu/ating reactot coolantlat least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

ACTION A LO1 Ireactor coolant p~umps 2operating e~jtther:F-ACTION A

a. Within [Iursafter switching to a three pump combination if the switch is made while operating, or

b. Prior to reactor criticality if the sitch s made while shut- A0 1

down.

. 7 i sic/s1 hl ht I See Spgeial Test Excep on 3.10.3. T A0S DAVISIBESSE, UNIT I 3/4 4-1 Amendment No. UJ,)*, -0,,

14;.135 Page 1 of 1 Attachment 1, Volume 9, Rev. 0, Page 69 of 415

Attachment 1, Volume 9, Rev. 0, Page 70 of 415 DISCUSSION OF CHANGES ITS 3.4.4, RCS LOOPS - MODES 1 AND 2 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 4.4.1.1.2.b requires a verification that the three reactor coolant pumps (RCPs) operating Reactor Protection System (RPS) trip setpoints for the High Flux and Flux-AFlux-Flow Functions are properly set prior to reactor criticality if the switch to three RCPs was made while not within the Applicability of CTS 3.4.1.1. This specific Surveillance is not maintained in the ITS. This changes the CTS by deleting the prior to criticality Surveillance.

The purpose of CTS 4.4.1.1.2.b is to ensure the three RCPs operating RPS trip setpoints are properly set prior to reactor criticality ifthe switch to three RCPs was made while not within the Applicability of CTS 3.4.1.1. This requirement however, is already enforced by other ITS requirements. ITS 3.4.4 requires the setpoints to be adjusted properly for operation with three RCPs. Thus, prior to entering the Applicability of ITS LCO 3.4.4 (MODES 1 and 2), the LCO must be met as required by ITS LCO 3.0.4. Furthermore, ITS LCO 3.3.1 provides the RPS setpoints for operation with three RCPs, and ITS LCO 3.0.4 would also require the setpoint requirement to be met prior to entering the two RPS Functions' (ITS Table 3.3.1-1 Functions 1.a and 8) Applicability (which includes MODES 1 and 2). Therefore, this current requirement is unnecessary and has been deleted. This change is designated as administrative and is acceptable since it does not result in any technical change to the CTS.

A03 The CTS 3.4.1.1 includes a footnote stating "See Special Test Exception 3.10.3."

ITS 3.4.4 Applicability does not contain the footnote or a reference to the Special Test Exception.

The purpose of the footnote is to alert the user that a Special Test Exception exists that may modify the Applicability of the Specification. However, CTS 3.10.3 has not been adopted into the ITS (see CTS 3/4.10.3 DOC M01 in Section 3.1), therefore the cross-reference is not needed. Furthermore, it is an ITS convention to not include these types of footnotes or cross-references even if the CTS LCO were maintained in the ITS. This change is designated as administrative as it incorporates an ITS convention with no technical change.

MORE RESTRICTIVE CHANGES M01 CTS 3.4.1.1 does not specify a default Action if more than one reactor coolant pump is not in operation or if the trips are not reduced in the 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months time period required by the CTS 3.4.1.1 Action. Thus, CTS 3.0.3 would be entered requiring entry into HOT STANDBY (MODE 3) within 7 hours8.101852e-5 days 0.00194 hours 1.157407e-5 weeks 2.6635e-6 months . ITS 3.4.4 ACTION A Davis-Besse Page 1 of 3 Attachment 1, Volume 9, Rev. 0, Page 70 of 415

Attachment 1, Volume 9, Rev. 0, Page 71 of 415 DISCUSSION OF CHANGES ITS 3.4.4, RCS LOOPS - MODES 1 AND 2 requires the plant to be in MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months under the same conditions.

This changes the CTS by providing one less hour for entry into MODE 3.

The purpose of requiring a shutdown when under the above conditions is to bring the unit to a subcritical condition since the unit is not within the accident analysis assumptions. This change is acceptable because it provides an adequate period of time to be in a MODE in which the LCO does not apply. The Completion Time of 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months is reasonable, based on operating experience, for reaching MODE 3 from full power in an orderly manner and without challenging unit systems.

RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES LA01 (Type 3 - Removing ProceduralDetails for Meeting TS Requirements or Reporting Requirements) CTS 4.4.1.1.1 states that the required reactor coolant loops shall be verified to be in operation and circulating reactor coolant at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . ITS SR 3.4.4.1 states that each RCS loop shall be verified to be in operation every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . This changes the CTS by moving the Surveillance Requirement detail to verify that the reactor coolant loops are circulating reactor coolant to the Bases.

The removal of this detail for performing Surveillance Requirements from the Technical Specifications is acceptable because this type of information is not necessary to be in the Technical Specifications in order to provide adequate protection of the public health and safety. The ITS retains the requirement that an RCS loop be in operation. This will require recirculation of reactor coolant since the ITS Bases specify that verification that a reactor coolant loop is in operation includes flow rate, temperature, or pump status monitoring, which helps ensure that forced flow is providing heat removal. Also, this change is acceptable because these types of procedural details will be adequately controlled in the ITS Bases. Changes to the Bases are controlled by the Technical Specification Bases Control Program in Chapter 5. This program provides for the evaluation of changes to ensure the Bases are properly controlled. This change is designated as a less restrictive removal of detail change because procedural details for meeting Technical Specification requirements are being removed from the Technical Specifications.

LESS RESTRICTIVE CHANGES L01 (Category3- Relaxation of Completion Time) CTS 3.4.1.1 Action a, which applies when shifting from four RCPs operating to three RCPs operating, requires a reduction of the High Flux trip setpoint from the four RCPs operating to three RCPs operating trip setpoint within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . Under the same conditions, ITS 3.4.4 ACTION A requires the reduction in the trip setpoints within 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months .

Davis-Besse Page 2 of 3 Attachment 1, Volume 9, Rev. 0, Page 71 of 415

Attachment 1, Volume 9, Rev. 0, Page 72 of 415 DISCUSSION OF CHANGES ITS 3.4.4, RCS LOOPS - MODES 1 AND 2 This changes the CTS by extending the Completion Time to reduce the trip setpoints from "4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months " to "10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months ."

The purpose of CTS 3.4.1.1 Action a is to ensure the proper trips setpoints for the new RCP configuration are set into the RPS High Flux Function. This change is acceptable because the Completion Time is consistent with safe operation under the specified Condition, considering the OPERABLE status of the redundant systems or features. This includes a reasonable time for repairs or replacement, and the low probability of a DBA occurring during the allowed Completion Time. The required Completion Time of 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months is reasonable based on the low probability of an accident occurring while operating outside the three RCPs operating trip setpoints, the automatic protection provided by the RPS Flux-AFlux-Flow Function (which is automatically reset), and the number of steps required to complete the Required Action, and the THERMAL POWER restriction provided in the LCO (i.e., 80.6% RTP). This proposed time is also consistent with the time allowed to reset the High Flux trip setpoints in ITS 3.2.4 and ITS 3.2.5, when QPT or a power peaking factor parameter is not within the required limits. Under these conditions, similar actions are required by plant personnel to reset the High Flux trip setpoints. This change is designated as less restrictive because additional time is allowed to reduce the trip setpoints.

L02 (Category 5 - Deletion of Surveillance Requirement) CTS 4.4.1.1.2 requires verification that the RPS trip setpoints for the High Flux and Flux-AFlux-Flow Functions are properly set after shifting from four RCPs operating to three RCPs operating. The ITS does not include this additional Surveillance as part of ITS 3.4.4 ACTION A for the Flux-AFlux-Flow Function. This changes the CTS by not including this conditional Surveillance for the Flux-AFlux-Flow Function.

The purpose of CTS 4.4.1.1.2 is to ensure the three RCPs operating RPS trip setpoints are properly set following a shift from four RCPs operating to three RCPs operating. However, the Flux-AFlux-Flow Function automatically changes its trip setpoint based on the number of operating RCPs. Thus, when one RCP trips, the three RCPs operating Flux-AFlux-Flow trip setpoint is automatically enabled - no manual setpoint adjustment is necessary. Thus the only function of this Surveillance is to ensure the automatic adjustment feature of the instrumentation functioned properly. This change is acceptable since the CHANNEL CALIBRATION testing required by ITS 3.3.1, "Reactor Protection System (RPS) Instrumentation," (ITS SR 3.3.1.3) already ensures that the instrumentation can automatically adjust the trip setpoints based on the number of operating RCPs. Therefore, this specific Surveillance is redundant to the normal, routine CHANNEL CALIBRATION Surveillances in the RPS Specification and is not needed. This change is designated as less restrictive because a Surveillance which is required in the CTS will not be required in the ITS.

Davis-Besse Page 3 of 3 Attachment 1, Volume 9, Rev. 0, Page 72 of 415

Attachment 1, Volume 9, Rev. 0, Page 73 of 415 Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs)

Attachment 1, Volume 9, Rev. 0, Page 73 of 415

Attachment 1, Volume 9, Rev. 0, Page 74 of 415 CTS RCS Loops - MODES 1 and 2 3.4.4 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4A4 RCS Loops - MODES 1 and 2 LCO 3.4.1.1 LCO 3.4.4 Two RCS Loops shall be in operation, with:

a. Four reactor coolant pumps (RCPs) operating or D Action 2 APPLICABILITY:

b. Three RCPs operating an MODES 1 and 2.

1[7g2INSERT HER POWER stricted to 1

}0 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME DOC M01 Requirements of LCO not met.

T 1 Be in MODE 3.

L_{for reasons other than Condition AJ SURVEILLANCE REQUIREMENTS 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months

}0 SURVEILLANCE FREQUENCY 4.4.1.1.1 SR 3.4.4.1 Verify required RCS loops are in operation. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months BWOG STS 3.4.4-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 0, Page 74 of 415

Attachment 1, Volume 9, Rev. 0, Page 75 of 415 CTS 3.4.4 2 INSERT 1 Action a 1. THERMAL POWER is < 80.6% RTP; Action a.1 2. LCO 3.3.1, "Reactor Protection System (RPS) Instrumentation," Function l.a (High Flux - High Setpoint), Allowable Value of Table 3.3.1-1 is reset for three RCPs operating; and Action a.2 3. LCO 3.3.1, Function 8 (Flux-AFlux-Flow), Allowable Value of Table 3.3.1-1 is reset for three RCPs operating.

2 INSERT 2 Action a A. Requirements of A.1 Satisfy the 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months LCO 3.4.4.b.2 not requirements of met. LCO 3.4.4.b.2.

Insert Page 3.4.4-1 Attachment 1, Volume 9, Rev. 0, Page 75 of 415

Attachment 1, Volume 9, Rev. 0, Page 76 of 415 JUSTIFICATION FOR DEVIATIONS ITS 3.4.4, RCS LOOPS - MODES I AND 2

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. ISTS LCO 3.4.4 is written for a plant whose design includes an automatic setdown feature for the nuclear overpower trip setpoint. That is, when shifting from four reactor coolant pump (RCP) operation to three RCP operation, the trip setpoints for the Reactor Protection System (RPS) instrumentation automatically adjust based on RCP configuration. This is described in the ISTS Bases, Background section, last paragraph. The Davis-Besse design does not include this automatic setdown feature for the High Flux trip setpoints - the setpoints must be manually adjusted. The current licensing basis provides for time to make a manual adjustment after shifting from four RCPs operating to three RCPs operating (CTS 3.4.1.1 Action a). ITS 3.4.4 has been written to allow the same two options as ISTS LCO 3.4.4: four RCPs must be operating (ITS LCO 3.4.4.a or three RCPs must be operating with a maximum power level restriction (ITS LCO 3.4.4.b and LCO 3.4.4.b.1). ITS LCO 3.4.4 also requires the trip setpoints of the High Flux and Flux-AFlux-Flow Functions to be set within the three RCP operating limits when operating with only three RCPs (ITS LCO 3.4.4.b.2 and LCO 3.4.4.b.3). Furthermore, a new ACTION has been added that provide 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months to manually reset the High Flux trip setpoints to within the Allowable Value for three RCP operation. While the current licensing basis only provides 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months to reset the trip setpoints (CTS 3.4.1.1 Action a), the 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months provided in ITS 3.4.4 ACTION A is consistent with the time provided in ISTS 3.2.4 and ISTS 3.2.5 to reset the High Flux trip setpoints when a QPT or power peaking factor limit is not met. Due to this change, ISTS 3.4.4 ACTION A has been renumbered as ACTION B and its associated Condition modified to only apply for reasons other than that provided in ITS 3.4.4 Condition A. In addition, the format of ITS 3.4.4 is also consistent with the format of NUREG-1433, ISTS 3.4.1, which has a similar requirement to manually change a trip setpoint when a recirculation pump (the BWR equivalent to an RCP) is not in operation.

Davis-Besse Page 1 of 1 Attachment 1, Volume 9, Rev. 0, Page 76 of 415

Attachment 1, Volume 9, Rev. 0, Page 77 of 415 Improved Standard Technical Specifications (ISTS) Bases Markup and Justification for Deviations (JFDs)

Attachment 1, Volume 9, Rev. 0, Page 77 of 415

Attachment 1, Volume 9, Rev. 0, Page 78 of 415 All changes are unless otherwise noted 9 RCS Loops - MODES 1 and 2

-B 3.4.4 B 3.4 REACTOR COOLANT SYSTEM (RCS)

B 3.4.4 RCS Loops - MODES 1 and 2 BASES BACKGROUND The primary function of the RCS is removal of the heat generated in the fuel due to the fission process, and transfer of this heat, via the steam generators (SGs), to the secondary plant.

The secondary functions of the RCS include:

a- Moderating the neutron energy level to the thermal state, to increase the probability of fission,* o

b. Improving the neutron economy by acting as a reflectorLl

c. Carrying the soluble neutron poison, boric acid 0-.__(

d. Providing a second barrier against fission product release to the environmen a

e. Removing the heat generated in the fuel due to fission product decay following a unit shutdown.

The RCS configuration for heat transport uses two RCS loops. Each RCS loop contains an SG and two reactor coolant pumps (RCPs). An RCP is located in each of the two SG cold legs. The pump flow rate has been sized to provide core heat removal with appropriate margin to departure from nucleate boiling (DNB) during power operation and for anticipated transients originating from power operation. This Specification requires two RCS loops with either three or four pumps to be in operation. With three pumps in operation the reactor power level is 2 restricted t0"-*/b RTP to preserve the core power to flow relationship, thus maintaining the margin to DN B. The intent of the Specification is to require core heat removal with forced flow during power operation.

Specifying the minimum number of pumps is an effective technique for designating the proper forced flow rate for heat transport, and specifying two loops provides for the needed amount of heat removal capability for the allowed power levels. Specifying two RCS loops also provides the minimum necessary paths (two SGs) for heat removal.

Flux - £TFlux - Flow (Table 3.3.1-1 Function 8)

The Reactor Protection System (RPS)l nulear 9werpowe trip setpoint is automatically reduced when one pump is taken out of service; manual resetting is not necessary.,

However, the RPS High Flux - High Setpoint (Table 3.3.1-1

[Function 1.a) trip setpoint must be manually reset.

BVVOG STS B 3.4.4-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 0, Page 78 of 415

Attachment 1, Volume 9, Rev. 0, Page 79 of 415 RCS Loops - MODES 1 and 2 B 3.4.4 BASES APPLICABLE Safety analyses contain various assumptions for the Design Bases SAFETY Accident (DBA) initial conditions including: RCS pressure, RCS ANALYSES temperature, reactor power level, core parameters, and safety system setpoints. The important aspect for this LCO is the reactor coolant forced flow rate, which is represented by the numberof pumps in service.

Both transient and steady state analyses have been performed to establish the effect of flow on DNB. The transient or accident analysis for the plant has been performed assuming either three or four pumps are in operation. The majority of the plant safety analysis is based on initial conditions at high core power or zero power. The accident analyses that are of most importance to RCP operation are the four pump coastdown, single pump locked rotor, and single pump (broken shaft or coastdown)

(Ref. 1).

Steady state DNB analysis has been performed for four, three, and two pump combinations. For four pump operation, the steady state DNB analysis, which generates the pressure and temperature SL (i.e., the 110.2% of 2817 MWt departure from nucleate boiling ratio (DNBR) limit), assumes a maximum power level ot[112] RTP. This is the design overpower condition for four pump operation. The V value is the accident analysis setpoint of the nuclear overpower (high flux) trip and is based on an analysis assumption that bounds possible instrumentation errors. The DNBR limit defines a locus of pressure and temperature points that result in a minimum DNBR greater than or equal to the critical heat flux correlation limit.

The three pump pressure temperature limit is tied to the steady state DNB analysis, which is evaluated each cycle. The flow used is the minimum allowed for three pump operation. The actual RCS flow rate will exceed the assumed flow rate. With three pumps operating, overpower protection is automatically provided by the Fpo~we-rto flow1 ati~oof -theRPS ISR Inuclear overpower I;(ased on RCS flow and AXIAL POWFER IMBALAFNýCE r setpoint. The maximum power level for three pump operation is

% RTP and is based on the three pump flow as a fraction of the four pump flow at full power.

Although the Specification limits operation to a minimum of three pumps total, existing design analyses show that operation with one pump in each loop (two pumps total) is acceptable when core THERMAL POWER is restricted to be proportionate to the flow. However, continued power operation with two RCPs removed from service is not allowed by this Specification.

RCS Loops - MODES 1 and 2 satisfy Criterion 2 of 10 CFR 50.36(c)(2)(ii).

BWOG STS B 3.4.4-2 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 0, Page 79 of 415

Attachment 1, Volume 9, Rev. 0, Page 80 of 415 B 3.4.4 O INSERT 1 RPS Flux - AFlux - Flow Function. Overpower protection is also provided by the High Flux - High Setpoint Function, which must be manually reset for three pump operation.

Insert Page B 3.4.4-2 Attachment 1, Volume 9, Rev. 0, Page 80 of 415

Attachment 1, Volume 9, Rev. 0, Page 81 of 415 RCS Loops - MODES 1 and 2 B 3.4.4 BASES LCO The purpose of this LCO is to require adequate forced flow for core heat removal. Flow is represented by the number of RCPs in operation in both RCS loops for removal of heat bY the two SGs. To meet safety analysis and certain acceptance criteria for DNB, four pumps are required at rated power; if mRPS setpoints only three pumps are available, power must be reduced.*" must be reset APPLICABILITY In MODES 1 and 2, the reactor is critical and has the potential to produce maximum THERMAL POWER. To ensure that the assumptions of the accident analyses remain valid, all RCS loops are required to be OPERABLE and in operation in these MODES to prevent DNB and core damage.

The decay heat production rate is much lower than the full power heat rate. As such, the forced circulation flow and heat sink requirements are reduced for lower, noncritical MODES as indicated by the LCOs for MODES 3, 4, and 5.

Operation in other MODES is covered by:

LCO 3.4.5, LCO 3.4.6, "RCS Loops - MODE "RCS Loops - MODE 0

LCO 3.4.7, "RCS Loops - MODE 5, Loops Fille LCO 3.4.8, "RCS Loops - MODE 5, Loops Not Filleý"

LCO 3.9.4, "Decay Heat Removal (DHR) and Coolant Circulation -

High Water Level" ; and LCO 3.9.5, "Decay Heat Removal (DHR) and Coolant Circulation -

Low Water Level; )

0 ACTIONS E 81 ISRT2 for reasons other than Condition A If the requirements of the LCO are not me the Required Action is to reduce power and bring the plant to MODE 3. This lowers power level and thus reduces the core heat removal needs and minimizes the possibility of violating DNB limits.

The Completion Time of 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months is reasonable, based on operating experience, to reach MODE 3 from full power conditions in an orderly manner and without challenging safety systems.

BVWOG STS B 3.4.4-3 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 0, Page 81 of 415

Attachment 1, Volume 9, Rev. 0, Page 82 of 415 B 3.4.4 (0 INSERT 2 A.1 If only three RCPs are in operation and the RPS High Flux - High Setpoint trip setpoints have not been reset to within the Allowable Value provided in Table 3.3.1-1 Function 1.a for three RCP operation, the trip setpoints must be reset within 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months .

This ensures the proper automatic overpower protection is provided by the RPS. The 10 hour1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months Completion Time is reasonable based on the low probability of an accident occurring while operating outside the three RCP limit, the automatic protection provided by the RPS Flux - AFlux - Flow Function (which is automatically reset), and the number of steps required to complete the Required Action.

Insert Page B 3.4.4-3 Attachment 1, Volume 9, Rev. 0, Page 82 of 415

Attachment 1, Volume 9, Rev. 0, Page 83 of 415 RCS Loops - MODES 1 and 2 B 3.4.4 BASES SURVEILLANCE SR 3.4.4.1 REQUIREMENTS This SR requires verification every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months of the required number of loops in operation. Verification includes flow rate, temperature, or pump status monitoring, which help ensure that forced -flow is providing heat removal while maintaining the margin to DNB. The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months interval has been shown byoperating practice to be sufficientto regularly assess degradation and verify operation within safety analyses assumptions. In addition, control room-indication and alarms will normally indicate loop status.

REFERENCES ET1F.SAR, 00 BWAOG STS B 3.4.4-4 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 0, Page 83 of 415

Attachment 1, Volume 9, Rev. 0, Page 84 of 415 JUSTIFICATION FOR DEVIATIONS ITS 3.4.4 BASES, RCS LOOPS - MODES 1 AND 2

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. 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 made to be consistent with the Specification.

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

Davis-Besse Page 1 of 1 Attachment 1, Volume 9, Rev. 0, Page 84 of 415

Attachment 1, Volume 9, Rev. 0, Page 85 of 415 Specific No Significant Hazards Considerations (NSHCs)

Attachment 1, Volume 9, Rev. 0, Page 85 of 415

Attachment 1, Volume 9, Rev. 0, Page 86 of 415 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.4.4, RCS LOOPS - MODES I AND 2 There are no specific NSHC discussions for this Specification.

Davis-Besse Page 1 of 1 Attachment 1, Volume 9, Rev. 0, Page 86 of 415

, Volume 9, Rev. 0, Page 87 of 415 ATTACHMENT 5 ITS 3.4.5, RCS LOOPS - MODE 3 , Volume 9, Rev. 0, Page 87 of 415

, Volume 9, Rev. 0, Page 88 of 415 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) , Volume 9, Rev. 0, Page 88 of 415

Attachment 1, Volume 9, Rev. 0, Page 89 of 415 IITS 3.4.5 ITSS 3/4.4, P.LCTOR COOLA.ST STST-(

SEUTt*V.-N A.'.* 1ROT STA.YDBT LTITTNC CON'DITION FOR OPERATION

a. .1.eAt st coolaot lops o shall be LCO 3.4.5 OPERAZLZ:

1. Re-actor 'COoIlaLopI and Its eassociat steasi

2. Reactor lat Loop 2 and Its assoc* cd Pteam 3.ewlmDcay 1"PI**See "C ITS 3.4.6, ITS 3.4.7, and

.seI2 ITS 3.4.8

b. At least o-*o a mcoolant loops shall be.,,

op erat£ion.

C. lNot mzire than 06 dGeCA heat rTiw p p m=, b opera ISe T vith the Dole suction path through VI-11 and DE-12 unless See ITS 3.4.7 the control pover has bean removed twu th DII-11 * ' . - and ITS 3.4.8 12 valve operator. or maua~l vAlue DI-21 and IM-23 a'o opewad.

d. The p lowamo Specifk~tions 3.0.3 0n~.4~ are not A02s

// See ITS 3.4.7 A.P?*LCAMiTT: HfODESq appl

[, W

'ble.

and ITS 3.4.89 A*,xA-uT MOE 3,4an *,* :See ITS 3.4.6 ACTION A, a. at

than the above requird Coolant lops O?'.,WLZ.

Required Action C.2 1I t[.° r.' " gv; t 0"C=% thi 72hours A03 ACTIONI Bo eLl= :t os

b. With none of the abovs required coola*t loops in operaticin, ACTION C suspend a2U2 operatiow ns voslving a reduction In boron Concentration of the Reactor Coolant Ssta aqd i=xd*A.te.Ly/

Initiate corrective action tor* et= t required coolamt or two required RCS loops inoperable See ITS 3.4.7 loop to opertdam. * *and ITS 3 4 8J I*The normal or emergency power source may be inoperable in MODE 5.-y- is I opmynoY be select.ed in mOuut 3 unles/te primary sidle l*emper~a~ur an~d --- A04 within the decay heat r=mv~ steSdsgnoditio#./

4pesur'e

'The decay heat removal pumps may be de-energized for up to I hour provided (1) no operations are permitted that would cause dilution of SeeITS3.4.7 and ITS 3.4.8 the reactor coolant system boron concentration, and (2) core outlet tPmnerature ismaintained at least 100F below saturation temperature.

DAVIS-BESSE UNIT 1 3/4 4-2 Amendment No. 4, A, *,

8, 92 Page 1 of 2 Attachment 1, Volume 9, Rev. 0, Page 89 of 415

Attachment 1, Volume 9, Rev. 0, Page 90 of 415 ITS 3.4.5 ITS 314.4 REACTOR COOLANT SYST2'f SURVEILLANCE REOUIRENENTS See ITS 3.4.6, ITS

.4.1.2.1 The required decay heat removal loop(s) shall be determined OPERABLEI 3.4.7, and ITS 3.4.8 ]

e Soecifcation fr 4.0.5.

SR 3.4.5.2 4.4.1.2.2 The required steam generator(s) shall be determined OPERABLE by verifying secondary side level to be greater than or equal to (a) IS inches above the lover tube sheet once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months if an associated reactor coolant pump is operating, or, (b) 35 inches above the lover tube sheet once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months if no reactor coolant pumps are operating.

SR 3.4.5.1 4.4.1.2.3 At least one coolant loop shall be verified to be in operation F[i6

[circuXating rp'actor Okolant lat least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . I Add proposed SR 3.4.5.3 MO01 DAVIS-BESSE, UNIT 1 3/4 4-2a Amendment No. ýf,135 Page 2 of 2 Attachment 1, Volume 9, Rev. 0, Page 90 of 415

Attachment 1, Volume 9, Rev. 0, Page 91 of 415 DISCUSSION OF CHANGES ITS 3.4.5, RCS LOOPS - MODE 3 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 3.4.1.2.d states that the provisions of Specifications 3.0.3 and 3.0.4 are not applicable. ITS 3.4.5 does not include this exception. This changes the CTS by deleting the specific exception to Specifications 3.0.3 and 3.0.4.

This change is acceptable because it results in no technical change to the Technical Specifications. CTS 3.0.3 (and ITS 3.0.3) provides actions for when an Action is not provided in the CTS for the given unit conditions. Furthermore, it only requires a shutdown to COLD SHUTDOWN (MODE 5). Since the Applicability of CTS 3.4.1.2 includes MODE 5, this exception is needed to ensure the unit does not enter CTS 3.0.3 if an Action of CTS 3.4.1.2 was not completed.

It essentially requires the Actions of CTS 3.4.1.2 to be met and not to default to the Actions of CTS 3.0.3. In the ITS, the CTS requirements have been divided up into MODE specific Specifications. Since ITS 3.4.5 covers only MODE 3, the specific exception to ITS 3.0.3 is not needed. CTS 3.0.4 provides requirements to preclude changing MODES with inoperable equipment. However, ITS LCO 3.0.4 has been modified to allow MODE changes under certain circumstances. This is justified in the Discussion of Changes for ITS Section 3.0.

Therefore, this specific exception to CTS 3.0.4 is not needed in the ITS. This change is designated as administrative because it does not result in a technical change to the CTS.

A03 CTS 3.4.1.2 Action a states that when less than the required reactor coolant loops are OPERABLE, action must be immediately initiated to restore the required loops. CTS 3.4.1.2 Action b states that when no coolant loops are in operation, all operations involving a reduction in boron concentration of the RCS must be suspended and action must be immediately initiated to return the required loop to operation. ITS 3.4.5 ACTION A specifies the Required Action for one required RCS loop inoperable. The Required Action is to restore the RCS loop to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . ITS 3.4.5 ACTION C specifies the Required Actions for two required RCS loops inoperable and for no required RCS loop in operation. The Required Actions are to immediately suspend operations that would cause introduction of coolant into the RCS with boron concentration less than required to meet the requirements of LCO 3.1.1, and to immediately initiate action to restore one RCS loop to OPERABLE status and operation. This changes the CTS by revising the Actions to immediately suspend operations that would cause introduction of coolant into the RCS with boron concentration less than required to meet the requirements of LCO 3.1.1 when two RCS loops are inoperable, and breaking up the Actions for one and two inoperable RCS loops into two separate Actions. The change to when one RCS loop is inoperable Davis-Besse Page 1 of 4 Attachment 1, Volume 9, Rev. 0, Page 91 of 415

Attachment 1, Volume 9, Rev. 0, Page 92 of 415 DISCUSSION OF CHANGES ITS 3.4.5, RCS LOOPS - MODE 3 (change in time from immediately to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months ) is justified in Discussion of Change L01.

This change is acceptable because it results in no technical changes to the CTS.

When both required RCS loops are inoperable, in all likelihood no RCS loops will be in operation. With no RCS loops in operation at the same time as both required RCS loops are inoperable, the same ITS ACTION (ACTION C) would be required. Therefore, since ITS 3.4.5 ACTION C would also require entry when no RCS loops are in operation, the identical actions would be required (i.e.,

immediately suspend operations that would cause introduction of coolant into the RCS with boron concentration less than required to meet the requirements of LCO 3.1.1). This change is designated as administrative because it does not result in any technical changes to the CTS.

A04 CTS LCO 3.4.1.2 Applicability Note

states that decay heat removal loops may not be used in MODE 3 to meet the LCO requirements, unless the primary side temperature and pressure are within the Decay Heat Removal System's design conditions. This Note is not included in the ITS. This changes the CTS by deleting the Applicability Note describing when decay heat removal loops can be used to meet the LCO requirements.

The purpose of the Note in CTS was to ensure DHR cooling is placed in service only if the required design parameters for DHR are met. As described in the ITS 3.4.5 Bases, LCO section, only the RCS loops are allowed to be used to meet the LCO requirements. The decay heat removal pumps are not described as an acceptable means for meeting the LCO. Therefore, the Applicability Note

is not needed for this MODE 3 Specification. This change is designated as administrative because no technical changes are being made to the CTS.

MORE RESTRICTIVE CHANGES M01 ITS SR 3.4.5.3 requires verification that correct breaker alignment and indicated power are available to each required pump. A Note further explains that the Surveillance 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. This Surveillance is not required by the CTS. This changes the CTS by requiring verification of correct breaker alignment and indicated power availability on required reactor coolant pumps that are not in operation.

The purpose of the ITS SR 3.4.5.3 is to ensure a standby pump is available to provide RCS cooling should the operating pump fail. This change is acceptable because the verification of proper breaker alignment and power availability ensures that an additional reactor coolant pump can be placed in operation, if needed, to maintain decay heat removal and reactor coolant circulation. This change is designated as more restrictive because it requires performance of the Surveillance on the non-operating reactor coolant pump.

RELOCATED SPECIFICATIONS None Davis-Besse Page 2 of 4 Attachment 1, Volume 9, Rev. 0, Page 92 of 415

Attachment 1, Volume 9, Rev. 0, Page 93 of 415 DISCUSSION OF CHANGES ITS 3.4.5, RCS LOOPS - MODE 3 REMOVED DETAIL CHANGES LA01 (Type I - Removing Details of System Design and System Description, Including Design Limits) CTS 3.4.1.2.a.1 and 3.4.1.2.a.2 contain a description of what constitutes an OPERABLE coolant loop. ITS 3.4.5 does not include this description of what constitutes an OPERABLE coolant loop. This changes the CTS by moving the details of what constitutes an OPERABLE coolant loop 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 a requirement for the RCS 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.

LA02 (Type 3 - Removing ProceduralDetails for Meeting TS Requirements or Reporting Requirements) CTS 4.4.1.2.3 states that the required coolant loops shall be verified to be in operation and circulating reactor coolant at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . ITS SR 3.4.5.1 states that one RCS loop shall be verified to be in operation every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . This changes the CTS by moving the Surveillance Requirement details, to verify that the coolant loops are circulating reactor coolant to the Bases.

The removal of this detail for performing Surveillance Requirements from the Technical Specifications is acceptable because this type of information is not necessary to be in the Technical Specifications in order to provide adequate protection of the public health and safety. The ITS retains the requirement that an RCS loop be in operation. This will require recirculation of reactor coolant since the ITS Bases specify that verification that a reactor coolant loop is in operation includes flow rate, temperature, or pump status monitoring, which helps ensure that forced or natural circulation flow is providing heat removal.

Also, this change is acceptable because these types of procedural details will be adequately controlled in the ITS Bases. Changes to the Bases are controlled by the Technical Specification Bases Control Program in Chapter 5. This program provides for the evaluation of changes to ensure the Bases are properly controlled. This change is designated as a less restrictive removal of detail change because procedural details for meeting Technical Specification requirements are being removed from the Technical Specifications.

LESS RESTRICTIVE CHANGES L01 (Category 1 - Relaxation of LCO Requirements) CTS 3.4.1.2 Action a, which applies when one or both required coolant loops are inoperable, states immediately initiate corrective action to return the required coolant loops to Davis-Besse Page 3 of 4 Attachment 1, Volume 9, Rev. 0, Page 93 of 415

Attachment 1, Volume 9, Rev. 0, Page 94 of 415 DISCUSSION OF CHANGES ITS 3.4.5, RCS LOOPS - MODE 3 OPERABLE status as soon as possible, or be in COLD SHUTDOWN within 20 hours2.314815e-4 days 0.00556 hours 3.306878e-5 weeks 7.61e-6 months . ITS 3.4.5 ACTION A, which applies when one RCS loop is inoperable, requires restoration of the RCS loop to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . If not restored, ITS 3.4.5 ACTION B requires the unit to be in MODE 4 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

This changes the CTS by allowing 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months to restore one inoperable RCS loop in lieu of requiring immediate action to be taken to restore the RCS loop, and allowing 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months to reach MODE 4 in lieu of 20 hours2.314815e-4 days 0.00556 hours 3.306878e-5 weeks 7.61e-6 months to reach MODE 5. Once in MODE 4, ITS 3.4.6 would become applicable.

The purpose of CTS 3.4.1.2 Action a is to provide appropriate compensatory measures when an RCS loop is inoperable. This change is acceptable since another RCS loop remains OPERABLE and capable of removing the decay heat.

In addition, this remaining RCS loop is still required to be in operation with a heat transfer capability greater than that needed to remove the decay heat produced in the reactor core. The proposed 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Completion Time is reasonable, considering the low probability of an event resulting in loss of the remaining RCS loop. Furthermore, the Applicability of ITS 3.4.5 is MODE 3. Therefore, the requirement to only require placing the unit in MODE 4 in lieu of MODE 5 (COLD SHUTDOWN) is acceptable because being in MODE 4 exits the Applicability.

The proposed Completion Time of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is reasonable, based on operating experience, to achieve cooldown and depressurization from MODE 3 without challenging plant systems. This change is designated as less restrictive since a longer Completion Time is being provided in the ITS than in the CTS.

Davis-Besse Page 4 of 4 Attachment 1, Volume 9, Rev. 0, Page 94 of 415

Attachment 1, Volume 9, Rev. 0, Page 95 of 415 Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs)

Attachment 1, Volume 9, Rev. 0, Page 95 of 415

Attachment 1, Volume 9, Rev. 0, Page 96 of 415 CTS RCS Loops - MODE 3 3.4.5 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.5 RCS Loops - MODE 3 3.4.1.2.a, LCO 3.4.5 Two RCS loops shall be OPERABLE and one RCS loop shall be in 3.4.1.2.b operation.

.NOT E --------

All reactor coola t pumps (RCPs) ma be removed from o eration for

_<8 hours per 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months period for the tr nsition to or from t Decay Heat Removal Syste , and all RCPs may e de-energized 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 f r any other reason, pr vided:

a. No opera ions are permitted th t would cause intro uction of coolant into the CS with boron conce tration less than re uired to meet the SDM of LCO 3.1.1 and 0

b. Core ou let temperature is mai tained at least [10]'F below saturati n temperature.

APPLICABILITY: MODE 3.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME Action a A. One RCS loop A.1 Restore RCS loop to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months inoperable. OPERABLE status.

Action a B. Required Action and B.1 Be in MODE 4. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months associated Completion Time of Condition A not met.

BWOG STS 3.4.5-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 0, Page 96 of 415

Attachment 1, Volume 9, Rev. 0, Page 97 of 415 CTS RCS Loops - MODE 3 3.4.5 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME Action a, C. Two RCS loops C.1 Suspend operations that Immediately Action b inoperable, would cause introduction of coolant into the RCS with OR boron concentration less than required to meet SDM Required RCS loop not of LCO 3.1.1.

in operation.

AND C.2 Initiate action to restore Immediately one RCS loop to OPERABLE status and operation.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY 4.4.1.2.3 SR 3.4.5.1 Verify one RCS loop is in operation. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months DOC M01 SR ------- NOTE-----------

Not required to be performed until 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after a 0

required pump is not in operation.

Verify correct breaker alignment and indicated 7 days power available to each required pump.

4.4.1.2.2 SR 3.4.5.2 Verify, for each required RCS loop. SG secondary side water level is: 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months 0 a) - 18 inches above the lower tube sheet if associated reactor coolant pump is operating: or b) > 35 inches above the lower tube sheet if reactor coolant pumps are not operating.

BVOG STS 3.4.5-2 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 0, Page 97 of 415

Attachment 1, Volume 9, Rev. 0, Page 98 of 415 JUSTIFICATION FOR DEVIATIONS ITS 3.4.5, RCS LOOPS - MODE 3

1. This LCO Note allowance has been deleted since it is not required. Davis-Besse is allowed to credit natural circulation flow to meet the LCO requirements. This was approved by the NRC as documented in the NRC Safety Evaluation for Amendment 38. Furthermore, ITS SR 3.4.5.2 has been added to ensure adequate SG water level, consistent with current licensing basis.

Davis-Besse Page 1 of 1 Attachment 1, Volume 9, Rev. 0, Page 98 of 415

Attachment 1, Volume 9, Rev. 0, Page 99 of 415 Improved Standard Technical Specifications (ISTS) Bases Markup and Justification for Deviations (JFDs)

Attachment 1, Volume 9, Rev. 0, Page 99 of 415

Attachment 1, Volume 9, Rev. 0, Page 100 of 415 RCS Loops - MODE 3 B 3.4.5 B 3.4 REACTOR COOLANT SYSTEM (RCS)

B 3.4.5 RCS Loops - MODE 3 BASES BACKGROUND The primary function of the reactor coolant in MODE 3 is removal of decay heat and transfer of this heat, via the steam generators (SGs), to the secondary plant fluid. The secondary function of the reactor coolant is to act as a carrier for soluble neutron poison, boric acid.

In MODE 3, reactor coolant pumps (RCPs) are used to provide forced circulation for heat removal during heatup and cooldown. The number of f:IfIce flw is usedro eete RCPs in operation will vary depending on operational needs, and inte o this LCOis o rovide forced flowvfrom at least one RCP for corespi ip rv e 0 Dl_ eat removal and transport. The flow provided by one RCP is adequate for heat removal and for boron mixing. However, two RCS loops are required to be OPERABLE to provide redundant paths for heat removal.

Reactor coolant natural circulation is not normally used; however, the natural circulation flow rate is sufficient for core cooling. If entry into ( and boron mixing 0

natural circulation is required, the reactor coolant at the highest elevation of the hot leg must be maintained subcooled for single phase circulation.

When in natural circulation, it is preferable to remove heat using both SGs to avoid idle loop stagnation that might occur if only one SG were in service. One generator will provide adequate heat removal.ron reduction ' natural circulation is, rohibited because mixing obtain a 0 homoge eous concentration inAll portions of the RCS can ot be ensur APPLICABLE No safety analysespare performed with initial conditions in MODE 3.

SAFETY ANALYSES Failure to provide heat removal may result in challenges to a fission rlop related to loss of R-CS oops product functionsbarrier. The RCS loops are part of the primary success path that or actuates to prevent or mitigate a Design Basis Accident or transient that either assumes the failure of, or presents a challenge to, the integrity of a fission product barrier.

RCS Loops - MODE 3 satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii).

BWOG STS B 3.4.5-1 Rev. 3.1, 12/01/05 Attachment 1, Volume 9, Rev. 0, Page 100 of 415

Attachment 1, Volume 9, Rev. 0, Page 101 of 415 RCS Loops - MODE 3 B 3.4.5 BASES LCO The purpose of this LCO is to require two loops to be available for heat removal thus providing redundancy. The LCO requires the two loops to be OPERABLE with the intent of requiring both SGs to be capable of and can be transferring heat from the reactor coolant at a controlled rate. Forced used to meet (0

n reactor coolant flow is ther way to transport heat, although natural rethe LCO quirements when forced flow is being used to", circulation flow provides adequate remova. A minimum of one running meet the LCO requirements. / RCP meets the LCO requirement for one loop in operationw Furthermore, the requirements for aI lop in operation are also met whenI natural circulation is established. The Note permi a limited period of o eration without RC s. All RCPs may be remove from operation for _ hours per 24 hou period for the transition to or f om the Decay Heat emoval (DHR) Syst m, and otherwise may e de-energized for _ hour per 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months p riod. This means that nat ral circulation has be n established. Wh n in natural circulation, bor n reduction with cool nt at boron concen rations less than 0

required to ass re the SDM of LCO .1.1, is prohibited b cause an even concentration istribution throughout the RCS cannot be ensured. Core outlet tempera re is to be maintain at least [1OOF bel w the saturation temperature s that no vapor bubble may form and pos bly cause a natural circula on flow obstruction.

In MODES 3, , and 5, it is someti s necessary to sto all RCP or DHR pump forced rculation (e.g., chan operation from on DHR train to the other, to perf m surveillance or sta up testing, to perf rm the transition to and from D R System cooling, o to avoid operation elow the RCP minimum net ositive suction head imit). The time peri d is acceptable because nat I circulation is adeq ate for heat remov I, or the reactor If forced flow is used, IST1 coolant temp rature can be mainta ned subcooled and boron stratification affecting rea tivity control is not ex ected.

n OPERABLE RCS loop consists of at least one OPERABLE RCP and an SG that is OPERABLE. An RCP is OPERABLE if it is capable 0

of being powered and is able to provide forced flow if required. . .FiNET2 APPLICABILIT'Y In MODE 3, the heat load is lower than at power; therefore, one RCS loop in operation is adequate for transport and heat removal. A second RCS loop is required to be OPERABLE but not in operation for redundant heat removal capability.

Operation in other MODES is covered by:

LCO 3.4.4, "RCS Loops - MODES 1 and LCO 3.4.6, LCO 3.4.7, "RCS Loops - MODE . M--I "RCS Loops - MODE 5, Loops Filled" 0 BWOG STS B 3.4.5-2 Rev. 3.1, 12/01/05 Attachment 1, Volume 9, Rev. 0, Page 101 of 415

Attachment 1, Volume 9, Rev. 0, Page 102 of 415 B 3.4.5 0 INSERT 1 Alternately, if natural circulation is used, an OPERABLE RCS loop consists of an SG that is OPERABLE.

0 INSERT 2 For forced flow, an OPERABLE steam generator requires > 18 inches of secondary side water level above the lower tube sheet. For natural circulation flow, an OPERABLE steam generator requires > 35 inches of secondary side water level above the lower tube sheet. In both cases, the steam generator maximum level must be maintained low enough such that the steam generator remains capable of decay heat removal by maintaining a steam flow path (i.e., < 625 inches full range level).

Insert Page B 3.4.5-2 Attachment 1, Volume 9, Rev. 0, Page 102 of 415

Attachment 1, Volume 9, Rev. 0, Page 103 of 415 RCS Loops - MODE 3 B 3.4.5 BASES APPLICABILITY (continued)

LCO 3-4.8, "RCS Loops - MODE 5, Loops Not Filled F-71 0 LCO 3.9.4, "Decay Heat Removal (DHR) and Cop nt Circulation - 00 LCO 3.9.5, High Water Level" and "Decay Heat Removal (DHR) and Coolant Circulation - Low 00 ACTIONS A.1 Water Level%

[l 0

Ior natural circulation If one RCS loop is inoperable, redundancy for forcedýflow heat removal is lost- The Required Action is restoration of the RCS loop to OPERABLE status within a Completion Time of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . This time allowance is a justified period to be without the redundant nonoperating loop because a single loop in operation has a heat transfer capability greater than that needed to remove the decay heat produced in the reactor core.

0 B.1 Required Actio.n J If restoration of an RCS loop as required in!A.1 is not possible within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , the unit must be brought to MODE 4. In MODE 4, the plant may be placed on the DHR System. The allowed Completion Time of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is reasonable, based on operating experience, to achieve cooldown and depressurization from the existing plant conditions and without challenging plant systems. 0 C.1 and C.2 If two RCS loops are inoperable or a required RCS loop is not in operation, lexcept as 4rovided in th5YNote in the CO section, all operations involving introduction of coolant into the RCS with boron concentration less than required to meet the minimum SDM of LCO 3.1.1 must be immediately suspended. Action to restore one RCS loop to operation shall be immediately initiated and continued until one RCS loop 0 is restored to OPERABLE status and to operation. Suspending the introduction of coolant into the RCS of coolant with boron concentration less than required to meet the minimum SDM of LCO 3.1.1 is required to assure continued safe operation. With coolant added without -oc-7ed circulation, unmixed coolant could be introduced to the core, however coolant added with boron concentration meeting the minimum SDM maintains acceptable margin to subcritical operations. The immediate Completion Time reflects the importance of maintaining operation for decay heat removal.

BVWG STS B 3.4.5-3 Rev. 3.1, 12/01/05 Attachment 1, Volume 9, Rev. 0, Page 103 of 415

Attachment 1, Volume 9, Rev. 0, Page 104 of 415 RCS Loops - MODE 3 B 3.4.5 BASES

'SURVEILLANCE SR 3.4.5.1 REQUIREMENTS This SR requires verification every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months that the required nurer of 0 loo anpurn is in operation. Verification includes flow rate, temperature, or pump status monitoring, which help ensure that forced o.r aturl circulation 0

flow is providing heat removal. The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months interval has been shown by operating practice to be sufficient to regularly assess RCS loop status. In addition, control room indication and alarms will normally indicate loop status.

SR 3.4.5.2 -E 0 Verification that each required RCP is OPERABLE ensures that the single failure criterion is met and that an additional RCS loop 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 availability to each required pump. Alternatively, verification that a pump is in operation also verifies proper breaker alignment and power availability. The Frequency of 7 days is considered reasonable in view of other administrative controls available and has been shown to be acceptable by operating experience.

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.

REFERENCES None.

/SR 3.4.5.2 SR 3.4.5.2 requires verification of SG OPERABILITY. SG OPERABILITY is verified by ensuring that the secondary side water level is either > 18 inches above the lower tube sheet when the associated reactor coolant pump is operating (forced flow) or_> 35 inches above the lower tube sheet if reactor coolant pumps are not operating (natural circulation flow). If the SG water level is 0

not within the associated limit, it may not be capable of providing the heat sink necessary for removal of decay heat. The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Frequency is considered adequate in view of other indications available in the control room to alert the operator to the loss of SG level.

BVVOG STS B 3.4.5-4 Rev. 3.1, 12/01/05 Attachment 1, Volume 9, Rev. 0, Page 104 of 415

Attachment 1, Volume 9, Rev. 0, Page 105 of 415 JUSTIFICATION FOR DEVIATIONS ITS 3.4.5 BASES, RCS LOOPS - MODE 3

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. 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 made to be consistent with the Specification.

5. Changes have been made to allow natural circulation flow to meet the LCO requirements. In addition, due to this change the LCO Note was deleted, thus the Note description in the Bases has been deleted.

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

Davis-Besse Page 1 of 1 Attachment 1, Volume 9, Rev. 0, Page 105 of 415

Attachment 1, Volume 9, Rev. 0, Page 106 of 415 Specific No Significant Hazards Considerations (NSHCs)

Attachment 1, Volume 9, Rev. 0, Page 106 of 415

Attachment 1, Volume 9, Rev. 0, Page 107 of 415 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.4.5, RCS LOOPS - MODE 3 There are no specific NSHC discussions for this Specification.

Davis-Besse Page 1 of 1 Attachment 1, Volume 9, Rev. 0, Page 107 of 415

, Volume 9, Rev. 0, Page 108 of 415 ATTACHMENT 6 ITS 3.4.6, RCS LOOPS - MODE 4 , Volume 9, Rev. 0, Page 108 of 415

, Volume 9, Rev. 0, Page 109 of 415 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) , Volume 9, Rev. 0, Page 109 of 415

Attachment 1, Volume 9, Rev. 0, Page 110 of 415 ITS 3.4.6 iTS 3/..PzCtlC- RCOOLA.'VT SYSITD(

SRUTDO~W HO ARD I(7 TANTIN?

LI1rVTINC CO? DIT!ON FOR OFERAT ION LCO 3.4.6 ...

2 a. At least two C C002AUt 100PS t shaL :.be OnAELZ:

I

1. Reactor Coolan LOOP I and its associat st4m gemeratore

2. Reactor Coo ant Loop 2 mill Its assocla ptan generator,

3. Decay E Removal Lo" I,*

A. . DecAy z at Rmoval IMM 2.0'

b. At least One lof/ths/ above coolant loops shell be in C Not noris t are decsy beat raw one any Operac vith the 9 0 10 suc t I* path through Is suction d f 11 the and DR-11 VS-12-and=lax*

cant 1 pover th:coor the pover has been ranow 'T 12 Val 'a operator, or man"I valves DR-2121 :

end M-23 are a ad r_'ý Ant1cA3r LITT: H~ODES 3, & andlsj ITS----a 3.4.1 ACTTO!:

ACTION A, A. WiVth ass than the above required Coolant Ioops O7na3Z.

Required Action B.2 +/-=ad Lately Initiate corrective action to returnu the required coolant loops to OPDJ3LI statu am %*on as L01 S. 24 M01 possibla, ýý be In COLD SEMDOM within 20 ACTION B Lunitiate correctle actlot to ratu= the equ.*Ld coola.-t loop to opearat4mz.

r See 1 or, two required RCS loops inoperable TS 3.4.7 and L1Thp npnmaL oreMeroency power source my be inoopperableeýn Id) . /ThiS - ITS 3.4.8 uera e and - See ITS loop may not be selected in MODE 3 unless the primary ioe emp pressure'are within the decay heat removal system's design conditions.

"The decay heat removal pumps may be de-energized for up to 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months provided (1) no operations are permitted that would cause dilution of See ITS 3.4.7and the reactor coolant system boron concentration, and (2) core outlet ITS 3.4.8 tpmperature is maintained at least IO*F below saturation temperature.

DAVIS-BESSE UNIT 1 3/4 4-2 Amendment No.

AS. 92 Page 1 of 2 Attachment 1, Volume 9, Rev. 0, Page 110 of 415

Attachment 1, Volume 9, Rev. 0, Page 111 of 415 ITS 3.4.6 ITS 3/4.4 REACTOR COOLANT SYSTEM SURVEILLANCE REOUIREMENTS The.2.1 Th: ired decay heat rm llop(s) shall bejlteýrmined OPERABLE A04 SR 3.4.6.2 4.4.1.2.2 The required stem generator(s) shall be determined OPERABLE by verifying secondary side level to be greater than or equal to (a) 18 inches above the lover tube sheet once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months if an associated reactor coolant pump is operating, or, (b) 35 inches above the lover tube sheet once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months if no reactor coolant pumps are operating.

SR 3.4.6.1 4.4.1.2.3 At least one coolant loop shall be verified to be in operationf Icircu ating r actor 0olant a east once per 1Z hours.

~Add proposed SR 3.4.6.3 DAVIS-BESSE, UNIT 1 3/4 4-2a Amendment No. M135 Page 2 of 2 Attachment 1, Volume 9, Rev. 0, Page 111 of 415

Attachment 1, Volume 9, Rev. 0, Page 112 of 415 DISCUSSION OF CHANGES ITS 3.4.6, RCS LOOPS - MODE 4 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 3.4.1.2.d states that the provisions of Specifications 3.0.3 and 3.0.4 are not applicable. ITS 3.4.6 does not include this exception. This changes the CTS by deleting the specific exception to Specifications 3.0.3 and 3.0.4.

This change is acceptable because it results in no technical change to the Technical Specifications. CTS 3.0.3 (and ITS 3.0.3) provides actions for when an Action is not provided in the CTS for the given unit conditions. Furthermore, it only requires a shutdown to COLD SHUTDOWN (MODE 5). Since the Applicability of CTS 3.4.1.2 includes MODE 5, this exception is needed to ensure the unit does not enter CTS 3.0.3 if an Action of CTS 3.4.1.2 was not completed.

It essentially requires the Actions of CTS 3.4.1.2 to be met and not to default to the Actions of CTS 3.0.3. In the ITS, the CTS requirements have been divided up into MODE specific Specifications. Since ITS 3.4.6 covers only MODE 4, the specific exception to ITS 3.0.3 is not needed. CTS 3.0.4 provides requirements to preclude changing MODES with inoperable equipment. However, ITS LCO 3.0.4 has been modified to allow MODE changes under certain circumstances. This is justified in the Discussion of Changes for ITS Section 3.0.

Therefore, this specific exception to CTS 3.0.4 is not needed in the ITS. This change is designated as administrative because it does not result in a technical change to the CTS.

A03 CTS 3.4.1.2 Action a states that when less than the required reactor coolant loops are OPERABLE, action must be immediately initiated to restore the required loops. CTS 3.4.1.2 Action b states that when no coolant loops are in operation, all operations involving a reduction in boron concentration of the RCS must be suspended and action must be immediately initiated to return the required loop to operation. ITS 3.4.6 ACTION A specifies the Required Action for one required RCS loop inoperable. The Required Action is to immediately initiate action to restore the second RCS loop to OPERABLE status. ITS 3.4.6 ACTION B specifies the Required Actions for two required RCS loops inoperable and for no required RCS loop in operation. The Required Actions are to immediately suspend operations that would cause introduction of coolant into the RCS with boron concentration less than required to meet the requirements of LCO 3.1.1, and to immediately initiate action to restore one RCS loop to OPERABLE status and operation. This changes the CTS by revising the Actions to immediately suspend operations that would cause introduction of coolant into the RCS with boron concentration less than required to meet the requirements of LCO 3.1.1 when two RCS loops are inoperable and to break up the Actions for one and two inoperable RCS loops into two separate Actions.

Davis-Besse Page 1 of 5 Attachment 1, Volume 9, Rev. 0, Page 112 of 415

Attachment 1, Volume 9, Rev. 0, Page 113 of 415 DISCUSSION OF CHANGES ITS 3.4.6, RCS LOOPS - MODE 4 This change is acceptable because it results in no technical changes to the CTS.

When both required RCS loops are inoperable, in all likelihood no RCS loops will be in operation. With no RCS loops in operation at the same time as both required RCS loops are inoperable, the same ITS ACTION (ACTION B) would be required. Therefore, since ITS 3.4.6 ACTION B would also require entry when no RCS loops are in operation, the identical actions would be required (i.e.,

immediately suspend operations that would cause introduction of coolant into the RCS with boron concentration less than required to meet the requirements of LCO 3.1.1). This change is designated as administrative because it does not result in any technical changes to the CTS.

A04 CTS 4.4.1.2.1 states that the required decay heat removal loop(s) shall be determined OPERABLE per Specification 4.0.5, the inservice testing Surveillance Requirements for ASME Code Class 1, 2, and 3 components. ITS 3.4.6 does not contain this explicit Surveillance Requirement. This changes the CTS by deleting the explicit requirement to perform the inservice testing Surveillance Requirements for ASME Code Class 1, 2, and 3 components.

The purpose of CTS 4.4.1.2.1 is to ensure the appropriate inservice testing Surveillance Requirements for ASME Code Class 1, 2, and 3 components are performed for the required decay heat removal loops. The inservice testing requirements of CTS 4.0.5 are retained in ITS 5.5.7, "Inservice Testing Program."

See the Discussion of Changes for ITS 5.5 for any changes to the requirements of CTS 4.0.5. The explicit cross reference is not necessary because when the system is determined to be inoperable when tested in accordance with the inservice testing program, the plant procedures will require the Decay Heat Removal System to be declared inoperable and the appropriate ITS 3.4.6 ACTIONS will be entered when applicable. This change is designated as administrative because it does not result in technical changes to the CTS.

MORE RESTRICTIVE CHANGES M01 When one RCS loop is inoperable, CTS 3.4.1.2 Action a requires a unit cooldown to COLD SHUTDOWN (MODE 5) only if immediate action is not initiated to restore the inoperable RCS loop as soon as possible. As long as action is being taken to restore the loop, entry into MODE 5 is not required. Under the same conditions, ITS 3.4.6 ACTION A will require both of the CTS Actions to be taken -

immediately initiating action to restore the inoperable RCS loop and a cooldown to MODE 5. This changes the CTS by requiring a unit cooldown to MODE 5 anytime one RCS loop is inoperable.

The purpose of CTS 3.4.1.2 Action a is to provide compensatory measures when an RCS loop is inoperable. The change is acceptable because placing the unit in MODE 5 is a conservative action with regard to decay heat removal. When a single RCS loop is inoperable, the other RCS loop is still capable of removing decay heat. This change is designated more restrictive because a cooldown to MODE 5 that is not required in the CTS will be required in the ITS.

M02 ITS SR 3.4.6.3 requires verification that correct breaker alignment and indicated power are available to each required pump. A Note further explains that the Davis-Besse Page 2 of 5 Attachment 1, Volume 9, Rev. 0, Page 113 of 415

Attachment 1, Volume 9, Rev. 0, Page 114 of 415 DISCUSSION OF CHANGES ITS 3.4.6, RCS LOOPS - MODE 4 Surveillance 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. This Surveillance is not required by the CTS. This changes the CTS by requiring verification of correct breaker alignment and indicated power availability on required pumps that are not in operation.

The purpose of ITS SR 3.4.6.3 is to ensure a standby pump is available to provide RCS cooling should the operating pump fail. This change is acceptable because the verification of proper breaker alignment and power availability ensures that an additional reactor coolant pump or DHR pump can be placed in operation, if needed, to maintain decay heat removal and reactor coolant circulation. This change is designated as more restrictive because it requires performance of the Surveillance on the non-operating pump.

RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES LA01 (Type I - Removing Details of System Design and System Description, Including Design Limits) CTS 3.4.1.2.a and 3.4.1.2.c contain a description of what constitutes an OPERABLE coolant loop. ITS 3.4.6 does not include this description of what constitutes an OPERABLE coolant loop. This changes the CTS by moving the details of what constitutes an OPERABLE coolant loop 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 a requirement for the RCS 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.

LA02 (Type 3 - Removing ProceduralDetails for Meeting TS Requirements or Reporting Requirements) CTS 4.4.1.2.3 states that the required coolant loops shall be verified to be in operation and circulating reactor coolant at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . ITS SR 3.4.6.1 states that the required DHR or RCS loop shall be verified to be in operation every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . This changes the CTS by moving the Surveillance Requirement detail to verify that the coolant loops are circulating reactor coolant to the Bases.

The removal of this detail for performing Surveillance Requirements from the Technical Specifications is acceptable because this type of information is not necessary to be in the Technical Specifications in order to provide adequate Davis-Besse Page 3 of 5 Attachment 1, Volume 9, Rev. 0, Page 114 of 415

Attachment 1, Volume 9, Rev. 0, Page 115 of 415 DISCUSSION OF CHANGES ITS 3.4.6, RCS LOOPS - MODE 4 protection of the public health and safety. The ITS retains the requirement that a DHR or RCS loop be in operation. This will require recirculation of reactor coolant since the ITS Bases specify that verification that a reactor coolant loop is in operation includes flow rate, temperature, or pump status monitoring, which helps ensure that forced or natural circulation flow is providing heat removal.

Also, this change is acceptable because these types of procedural details will be adequately controlled in the ITS Bases. Changes to the Bases are controlled by the Technical Specification Bases Control Program in Chapter 5. This program provides for the evaluation of changes to ensure the Bases are properly controlled. This change is designated as a less restrictive removal of detail change because procedural details for meeting Technical Specification requirements are being removed from the Technical Specifications.

LESS RESTRICTIVE CHANGES L01 (Category 3 - Relaxation of Completion Time) CTS 3.4.1.2 Action a requires a cooldown to COLD SHUTDOWN (MODE 5) within 20 hours2.314815e-4 days 0.00556 hours 3.306878e-5 weeks 7.61e-6 months under certain conditions. When a cooldown to MODE 5 is required in ITS 3.4.6 ACTION A, 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months are provided to be in MODE 5. This changes the CTS by extending the time allowed to reach MODE 5 from 20 hours2.314815e-4 days 0.00556 hours 3.306878e-5 weeks 7.61e-6 months to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

The purpose of the CTS 3.4.1.2 Action a time limit to reach MODE 5 is to provide an appropriate amount of time for the unit to be cooled down to MODE 5 conditions in a controlled manner. This change is acceptable because the proposed time is still limited, and provides additional time to reach MODE 5 in an orderly manner and without challenging plant systems. During this additional 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , another RCS loop is still OPERABLE, thus capable of removing the decay heat. Furthermore, the proposed time is consistent with the time normally provided to reach MODE 5 from MODE 4 in other CTS Specifications, such as CTS 3.0.3. This change is designated as less restrictive since more time is provided in the ITS to reach MODE 5 than is provided in the CTS.

L02 (Category 4 - Relaxation of RequiredAction) CTS 3.4.1.2 Action b states that when no coolant loops are in operation, all operations involving a reduction in boron concentration of the RCS must be suspended. ITS 3.4.6 Required Action B.1 states that operations that would cause introduction of coolant into the RCS with boron concentration less than required to meet the requirements of LCO 3.1.1, "SHUTDOWN MARGIN (SDM)," must be suspended. This relaxes the CTS Action by revising the action from suspending reductions in boron concentration to suspending introduction of coolant into the RCS with a boron concentration less than required to meet LCO 3.1.1.

The purpose of CTS 3.4.1.2 Action b is to ensure that "pockets" of coolant with boron concentration less than that required to maintain the SDM are not created when there is no forced or natural circulation flow through the reactor. 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 and the low probability of a DBA Davis-Besse Page 4 of 5 Attachment 1, Volume 9, Rev. 0, Page 115 of 415

Attachment 1, Volume 9, Rev. 0, Page 116 of 415 DISCUSSION OF CHANGES ITS 3.4.6, RCS LOOPS - MODE 4 occurring during the repair period. As long as coolant with boron concentration less than that required to meet the SDM requirement in LCO 3.1.1 is not introduced into the RCS, there is no possibility of creating "pockets" of coolant with less than the required boron concentration. 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 5 of 5 Attachment 1, Volume 9, Rev. 0, Page 116 of 415

Attachment 1, Volume 9, Rev. 0, Page 117 of 415 Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs)

Attachment 1, Volume 9, Rev. 0, Page 117 of 415

Attachment 1, Volume 9, Rev. 0, Page 118 of 415 CTS RCS Loops - MODE 4 3.4.6 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.6 RCS Loops - MODE 4 3.4.1.2 LCO 3.4.6 Two loops consisting of any combination of RCS loops and decay heat removal (DHR) loops shall be OPERABLE and one loop shall be in operation.

.-- - - - - - - - NO TE-All reactor coolant pu ps (RCPs) may be removed from peration for

_<8 hours per 24 ho period for the transition to or from he DHR System, and all RCPs and D R pumps may be de-energized for/ s 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months per provided:

8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months period for a y other reason,

a. No operatio s are permitted that would cause in oduction of coolant Q into the RC with boron concentration less tha required to meet the SDM o LCO 3.1.1 and f

b. Core outl t temperature is maintained at leas 10WF below saturation temperat re.

APPLICABILITY: MODE 4.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME Action a A. One required loop A.1 Initiate action to restore a Immediately inoperable, second loop to OPERABLE status.

AND A.2 -------------- NOTE -.......----

Only required if one DHR loop is OPERABLE.

Be in MODE 5. 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months BWOG STS 3.4.6-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 0, Page 118 of 415

Attachment 1, Volume 9, Rev. 0, Page 119 of 415 CTS RCS Loops - MODE 4 3.4.6 ACTIONS (continued)_

CONDITION REQUIRED ACTION COMPLETION TIME Actions a and b B. Two required loops B-1 Suspend operations that Immediately inoperable, would cause introduction of coolant into the RCS with OR boron concentration less than required to meet SDM Required loop not in operation.

of LCO 3.1.1 0

AND "SHUTDOWN MARGIN (SOM)

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

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY 4.4.1.2.3 SR 3.4.6.1 Verify required DHR or RCS loop is in operation. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months DOC M02 SR--- ---- ---------------- NOTE -----------

0 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.

Verify correct breaker alignment and indicated 7 days power available to each required pump.

SR 3.4.6.2 Verify, for each required RCS loop, SG secondary side water level is: 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months 4.4.1.2.2 a) > 18 inches above the lower tube sheet if associated reactor coolant pump is operating: or 0 b) - 35 inches above the lower tube sheet if reactor coolant pumps are not operating.

BWOG STS 3.4.6-2 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 0, Page 119 of 415

Attachment 1, Volume 9, Rev. 0, Page 120 of 415 JUSTIFICATION FOR DEVIATIONS ITS 3.4.6, RCS LOOPS - MODE 4

1. The title of the LCO has been provided since this is the first reference to the LCO.

2. This LCO Note allowance has been deleted since it is not required. Davis-Besse is allowed to credit natural circulation flow to meet the LCO requirements. This was approved by the NRC as documented in the NRC Safety Evaluation for Amendment 38. Furthermore, ITS SR 3.4.6.2 has been added to ensure adequate SG water level, consistent with current licensing basis.

Davis-Besse Page 1 of 1 Attachment 1, Volume 9, Rev. 0, Page 120 of 415

Attachment 1, Volume 9, Rev. 0, Page 121 of 415 Improved Standard Technical Specifications (ISTS) Bases Markup and Justification for Deviations (JFDs)

Attachment 1, Volume 9, Rev. 0, Page 121 of 415

Attachment 1, Volume 9, Rev. 0, Page 122 of 415 RCS Loops - MODE 4

B 3.4.6 B3.4 REACTOR COOLANT SYSTEM (RCS)

B 3.4.6 RCS Loops - MODE 4 BASES BACKGROUND In MODE 4, the primary function of the reactor coolant is the removal of decay heat and transfer of this heat to the steam generators (SGs) or decay heat removal (DHR) heat exchangers. The secondary function of the reactor coolant is to act as a carrier for soluble neutron poison, boric acid.

forcd in MODE 4, either reactor coolant pumps (RCPs) or DHR pumps can be ifforced flow is used fo oolant circulation. The number of pumps in operation can vary used to meet the suit the operational needs. [The in nt of this LCO- Jitrvdeforced vded flovfrom at least one RCP or one DHR pump for decay neat removal and transport. The flow provided by one RCP or one DHR pump is adequate 0 for heat removal. The other intent of this LCO is to require that two paths INSERT 1 (loops) be available to provide redundancy for heat removal.

related to loss]

of RCS loops APPLICABLE SAFETY No safety analyses'are performed with initial condition in MODE 4.

0 ANALYSES RCS Loops - MODE 4 satisfies Criterion 4 of 10 CFR 50.36(c)(2)(ii).

LCO The purpose of this LCO is to require that two loops, RCS or DHR, be OPERABLE in MODE 4 and one of these loops be in operation. The LCO allows the two loops that are required to be OPERABLE to consist of any combination of RCS or DHR System loops. Any one loop in operation provides enough flow to remove the decay heat from the core or natural with forced circulation. The second loop that is required to be o OPERABLE provides redundant pathg for heat removal.

The Note permi a limited period of oeration without R Ps. All RCPs may be remove from operation for </8 hours per 24 hou= period for the transition to or om the DHR Systemeand otherwise may/be de-energized 00 for 5 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months pe 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months period. This L- ans that natural irculation has been establish d using the SGs. Th Note prohibits bor n dilution with coolant at bor n concentration s than required to asqure the SDM of LCO 3.1.1 is intained when for flow is stopped bqcause an even concentration distribution cannot b ensured. Core out t temperature is to be maintai ed at least 10*F belo saturation temper ture so that no vapor bubble may form and possib cause a natural ci culation flow obstruction. i BVWOG STS B 3.4.6-1 Rev. 3.1, 12/01/05 Attachment 1, Volume 9, Rev. 0, Page 122 of 415

Attachment 1, Volume 9, Rev. 0, Page 123 of 415 B 3.4.6 5 INSERT 1 Reactor coolant natural circulation is not normally used; however, the natural circulation flow rate is sufficient for core cooling and boron mixing. If entry into natural circulation is required, the reactor coolant at the highest elevation of the hot leg must be maintained subcooled for single phase circulation. When in natural circulation, it is preferable to remove heat using both SGs to avoid idle loop stagnation that might occur if only one SG were in service. One generator will provide adequate heat removal.

Insert Page B 3.4.6-1 Attachment 1, Volume 9, Rev. 0, Page 123 of 415

Attachment 1, Volume 9, Rev. 0, Page 124 of 415 RCS Loops - MODE 4 B 3.4.6 BASES LCO (continued)

The Note also p rmits the DHR pump to be stopped for 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months er hour period. en the DHR pumps are stopped, no alt rnate heat removal path e ists, unless the RCS nd SGs have been laced in service in force or natural circulation The response oftie RCS without the DHR Syste depends on the cor decay heat load ard the length of time that the D R pumps are stoppe .As decay heat ddi inihshes, the o effects on RCS temperature and pre sure diminish. With ut cooling by DHR, higher h at loads will cause th reactor coolant te perature and pressure to inc ease at a rate propo ional to the decay h at load.

Because pres ure can increase, the pplicable system p essure limits (pressure and emperature (P/T) or I w temperature ove pressure protection (LT P) limits) must be ob erved and forced HR flow or heat 0 removal via th SGs must be re-est blished prior to rea hing the pressure limit. The circumstances f r stopping both DH trains are to be limited to situ tions where:

a. Pressure and pressure and te perature increases n be maintained well with n the allowable press re (PIT and LTOP) nd 1O°F subcooli g limits or

b. An alter ate heat removal pat through the SG is i operation.

Ifforced flow is used, )--rn OPERABLE RCS loop consists of at least one OPERABLE RCP and an SG that is OPERABLE 0!

0!

Similarly for the DHR System, an OPERABLE DHR loop is comprised of tthe OPERABLE DHR pump(s) capable of providing forced flow to the Hcaners). DHR pumps are OPERABLE if they are capable iNSERT 3 0!

of being powered and are able to provide flow if required.

APPLICABILITY In MODE 4, this LCO applies because it is possible to remove core decay heat and to provide proper boron mixing with either the RCS loops and SGs or the DHR System.

Operation in other MODES is covered by:

LCO 3.4.4, "RCS Loops - MODES 1 and LCO 3.4.5, "RCS Loops - MODE "

LCO LCO 3.4.7, 3.4.8, "RCS Loops - MODE 5, Loops Fille "RCS Loops - MODE 5, Loops Not Fille 0 LCO 3.9.4, "Decay Heat Removal DHR) and Coolant Circulation -

High Water Level" FM DE 6 and 0!

LCO 3.9.5, "Decay Heat Removal (DHR) and Coolant Circulation -

Low Water L (M E 6). 0 BWOG STS B 3.4.6-2 Rev. 3.1, 12/01/05 Attachment 1, Volume 9, Rev. 0, Page 124 of 415

Attachment 1, Volume 9, Rev. 0, Page 125 of 415 B 3.4.6 0 INSERT 2 Alternately, if natural circulation is used, an OPERABLE RCS loop consists of an SG that is OPERABLE. For forced flow, an OPERABLE steam generator requires - 18 inches of secondary water level above the lower tube sheet. For natural circulation flow, an OPERABLE steam generator requires __35 inches of secondary water level above lower tube sheet. In both cases, the steam generator maximum level must be maintained low enough such that the steam generator remains capable of decay heat removal by maintaining a steam flow path (i.e., -<625 inches full range level).

2 INSERT 3 Furthermore, the two DHR loops share the same suction path through DH-1 1 and DH-1 2.

Therefore, when both DHR loops are being used to meet the LCO requirements, control power is required to be removed from DH-1 1 and DH-12 valve operators, or manual valves DH-21 and DH-23 are required to be open. 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 3.4.6-2 Attachment 1, Volume 9, Rev. 0, Page 125 of 415

Attachment 1, Volume 9, Rev. 0, Page 126 of 415 RCS Loops - MODE 4

B 3.4.6 BASES ACTIONS A.1 If only one required RCS loop or DHR loop is OPERABLE and in operation, redundancy for heat removal is lost. Action must be initiated to restore a second loop to OPERABLE status. The immediate Completion Time reflects the importance of maintaining the availability of two .paths for heat removal.

A.2 If restoration is not accomplished and a DHR loop is OPERABLE, the unit must be brought to MODE 5 within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . Bringing the unit to MODE 5 is a conservative action with regard to decay heat removal. With only one DHR loop OPERABLE, redundancy for decay heat removal is lost and, in the event of a loss of the remaining DHR loop, it would be safer to initiate that loss from MODE 5 rather than MODE 4.

The Completion Time of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months is reasonable, based on operating experience, to reach MODE 5 in an orderly manner and without challenging plant systems.

This Required Action is modified by a Note which indicates that the unit must be placed in MODE 5 only if a DHR loop is OPERABLE. With no DHR loop OPERABLE, the unit is in a condition with only limited cooldown capabilities. Therefore, the actions are to be concentrated on the restoration of a DHR loop, rather than a cooldown of extended duration.

B.1 and B.2 If two required RCS or DHR loops are inoperable or a required loop is not in operation, lexcept durpg conditions perpitfted by the Nxte- in the LCOI Fse ion, a*l operations involving introduction of coolant into the RCS with boron concentration less than required to meet the minimum SDM of LCO 3.1.1 must be suspended and action to restore one RCS or DHR loop to OPERABLE status and operation must be initiated. The required margin to criticality must not be reduced in this type of operation.

Suspending the introduction of coolant, into the RCS, with boron concentration less than required to meet the minimum SDM of LCO 3.1.1 is required to ensure continued safe operation. With coolant added without fo-ced circulation, unmixed coolant could be introduced to the core, however, coolant added with boron concentration meeting the 0

minimum SDM maintains acceptable margin to subcritical operations.

The immediate Completion Times reflect the importance of maintaining operation for decay heat removal. The action to restore must continue until one loop is restored to operation.

BWVOG STS B 3.4.6-3 Rev. 3.1, 12/01/05 Attachment 1, Volume 9, Rev. 0, Page 126 of 415

Attachment 1, Volume 9, Rev. 0, Page 127 of 415 RCS Loops - MODE 4 B 3.4.6 BASES SURVEILLANCE SR 3.4.6.1 REQUIREMENTS This Surveillance requires verification every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months of the required DH R or natural or RCS loop in operation to ensure forcediflow is providing decay heat removal. Verification includes flow rate, temperature, or pump status circulation 0 monitoring. The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months interval has been shown'by operating practice to be sufficient to regularly assess RCS loop:status. In addition, control room indication and alarms will normally indicate loop status. 0 INSERT 4 SR 3A.4.6.*

0 Verification that each required pump is OPERABLE ensures that an additional RCS or DHR loop 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 each required pump. Alternatively, verification that a pump is in operation also verifies proper breaker alignment and power availability. The Frequency of 7 days is considered reasonable in view of other administrative controls and has been shown to be acceptable by operating experience.

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.

REFERENCES None.

B'AOG STS B 3.4.6-4 Rev. 3.1, 12/01/05 Attachment 1, Volume 9, Rev. 0, Page 127 of 415

Attachment 1, Volume 9, Rev. 0, Page 128 of 415 O* INSERT 4 SR 3.4.6.2 SR 3.4.6.2 requires verification of SG OPERABILITY. SG OPERABILITY is verified by ensuring that the secondary side water level is either -- 18 inches above the lower tube sheet when the associated reactor coolant pump is operating (forced flow) or

> 35 inches above the lower tube sheet if reactor coolant pumps are not operating (natural circulation flow). If the SG water level is not within the associated limit, it may not be capable of providing the heat sink necessary for removal of decay heat. The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Frequency is considered adequate in view of other indications available in the control room to alert the operator to the loss of SG level Attachment 1, Volume 9, Rev. 0, Page 128 of 415

Attachment 1, Volume 9, Rev. 0, Page 129 of 415 JUSTIFICATION FOR DEVIATIONS ITS 3.4.6 BASES, RCS LOOPS - MODE 4

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. Changes are made (additions, deletions, and/or changes) to the ISTS Bases which reflect the plant specific nomenclature, number, reference, system description, analysis, or licensing basis description.

3. Changes made to be consistent with the Specification.

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

5. Changes have been made to allow natural circulation flow to meet the LCO requirements. In addition, due to these changes, the LCO Note was deleted; thus the Note description in the Bases has also been deleted.

Davis-Besse Page 1 of 1 Attachment 1, Volume 9, Rev. 0, Page 129 of 415

Attachment 1, Volume 9, Rev. 0, Page 130 of 415 Specific No Significant Hazards Considerations (NSHCs)

Attachment 1, Volume 9, Rev. 0, Page 130 of 415

Attachment 1, Volume 9, Rev. 0, Page 131 of 415 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.4.6, RCS LOOPS - MODE 4 There are no specific NSHC discussions for this Specification.

Davis-Besse Page 1 of 1 Attachment 1, Volume 9, Rev. 0, Page 131 of 415

Attachment 1, Volume 9, Rev. 0, Page 132 of 415 ATTACHMENT 7 ITS 3.4.7, RCS LOOPS - MODE 5, LOOPS FILLED Attachment 1, Volume 9, Rev. 0, Page 132 of 415

, Volume 9, Rev. 0, Page 133 of 415 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) , Volume 9, Rev. 0, Page 133 of 415

Attachment 1, Volume 9, Rev. 0, Page 134 of 415 ITS 3.4.7 ITS I3/4-.G ISH TWUMMt.rATOR JL%

A'MY1TCOOL&.-%7 STSM1 STAN*DSY fLTI~TTC CODIT10O4 FOR OPrR.4 T0N LAO1 3.4.1.2 a. At least two iE ccoolant loops 1ou hall be \

LCO 3.4.7 OPERABLE:

2. Reactor Coolan Loop I and Its associated stm-e -

SeueratoT Reactor Coo *an Loop 2 and its assoc/at Pte=n LAO 1 2.

generator.

3. Decay 1e.s/ Remov al 1" o

.. Decay B t :Rmal Loop 2.'

b. Atlacctl/zw SAr least one w- coolant loops shall be in Io/r/a~i!ý operation.**

C. Not more n: one decy nbsa r Val pup ma7 go zparare unlessLA01 v.th the le suction path /hroulsh D-11 and D9-the cos*tl pver has beyraoved from the Dt- 'and 03-3 12 valv/ opera1to:r, or u&4valves DU-Z1 and JJ-2 are open".d

d. 7&qe p-=7,1ona of Seacifthtions 3.0.3 ande e O "See ITS applJtcjrla.

AP?LICA31 -TY: MODES 3- 4anId 5O7 3.4.6 See ITS I

3.4.5 ACTZO!:

ACTION A, a. with qag than the above required coolamt loops OF="lZ, Id-ately initiate corrective action to return the See ITS A0 Required Action B.2 3.4.5 required coolat 1ooi8 to P?.A3LZ status as soon as and possible, lor be In COLD S2UZUO N in 0ithh .2 ITS 3.4.6

[ or tw o re ouired D )HRIo ~ in omerable p era t

b. With none of the abovt requ:Led, coolaIar pe-t ACTION B L01 suspend alU operations Involving a Treunutizo 19 DO concentration of the Reactor Coolant Stat a* = ttly

Ln&t2Ate COtrectZ* e ac0t o0 .un requST dIcoo.ant loop to operati"n. A04 I-The normal*or emergency power" $ouromay be inoerae in KID S iSSee ITS

[lop may not be selected in M00t 3 unless Mhe primar111y sid L*,coU ' e and---- 3.4.5 lpressure'°re within the decay heat removal system's design conditions. "

"The decay heat remo al pumps may be de-energized f up to 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months provided (1) no oper tions are permitted that would ause dilution of the reactor coolant ystem boron concentration, and (2) core outlet t~mperature is mainn aned at least 106F below satur tion temperature.

DAVIS-BESSE UNIT 1 3/4 4-2 Amendment No.

Ia, 92 A, h, t, Page 1 of 2 Attachment 1, Volume 9, Rev. 0, Page 134 of 415

Attachment 1, Volume 9, Rev. 0, Page 135 of 415 ITS 3.4.7 ITS 3/4.4 REACTOR COOLANT SYST-E.

SURVEILLANCE REOUIREMENTS 4.12.1 The r uired decay heat rem2ov1 loop(s) shall be d terained OPERABLE A0 4.0.5.

SR 3.4.7.2 4.4.1.2.2 The required steam generator(s) shall be determined OPERABLE by See ITS 3.4.5 and verifying secondary side level to be greater than or equal to (a) 18 inches ITS 3.4.6 above the lover tube sheet once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months if an associated reactor cool*Int.

RUMP Is opera ti ng, or{(b) 35 inches above the lover tube sheet once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ýf no repitor coolan pumps are perating.I A06 SR 3.4.7.1 4.4.1.2.3 At least one coolant loop shall be verified to be in operation and circu Yatin ridactor o iattat least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ,

* ~Add proposed SR 3.4.7.3 DAVIS-BESSE, UNIT 1 3/4 4-2a Amendment No. Y1,135 Page 2 of 2 Attachment 1, Volume 9, Rev. 0, Page 135 of 415

Attachment 1, Volume 9, Rev. 0, Page 136 of 415 DISCUSSION OF CHANGES ITS 3.4.7, RCS LOOPS - MODE 5, LOOPS FILLED 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 3.4.1.2.d states that the provisions of Specifications 3.0.3 and 3.0.4 are not applicable. ITS 3.4.7 does not include this exception. This changes the CTS by deleting the specific exception to Specifications 3.0.3 and 3.0.4.

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 5, which is the Applicability of ITS 3.4.7. Therefore, this exception to CTS 3.0.3 is redundant and unnecessary.

CTS 3.0.4 provides requirements to preclude changing MODES with inoperable equipment. However, ITS LCO 3.0.4 has been modified to allow MODE changes under certain circumstances. This is justified in the Discussion of Changes for ITS Section 3.0. Therefore, this specific exception to CTS 3.0.4 is not needed in the ITS. This change is designated as administrative because it does not result in a technical change to the CTS.

A03 CTS 3.4.1.2 Action a states that when less than the required reactor coolant loops are OPERABLE, action must be immediately initiated to restore the required loops. CTS 3.4.1.2 Action b states that when no coolant loops are in operation, all operations involving a reduction in boron concentration of the RCS must be suspended and action must be immediately initiated to return the required loop to operation. ITS 3.4.7 ACTION A specifies the Required Actions when one of the two required loops is inoperable. Required Action A.1 is to immediately initiate action to restore the second loop to OPERABLE status.

ITS 3.4.7 ACTION B specifies the Required Actions when two required loops are inoperable and when no required loop is in operation. The Required Actions are to immediately suspend operations that would cause introduction of coolant into the RCS with boron concentration less than required to meet the requirements of LCO 3.1.1, and to immediately initiate action to restore one loop to OPERABLE status and operation. This changes the CTS by revising the Actions to immediately suspend operations that would cause introduction of coolant into the RCS with boron concentration less than required to meet the requirements of LCO 3.1.1 when two required loops are inoperable and to break up the Actions for one and two inoperable required loops into two separate Actions.

This change is acceptable because it results in no technical changes to the CTS.

When both required loops are inoperable, in all likelihood no loops will be in operation. With no loops in operation at the same time as both required loops are inoperable, the same ITS ACTION (ACTION B) would be required.

Therefore, since ITS 3.4.7 ACTION B would also require entry when no loops are in operation, the identical actions would be required (i.e., immediately suspend Davis-Besse Page 1 of 5 Attachment 1, Volume 9, Rev. 0, Page 136 of 415

Attachment 1, Volume 9, Rev. 0, Page 137 of 415 DISCUSSION OF CHANGES ITS 3.4.7, RCS LOOPS - MODE 5, LOOPS FILLED operations that would cause introduction of coolant into the RCS with boron concentration less than required to meet the requirements of LCO 3.1.1). This change is designated as administrative because it does not result in any technical changes to the CTS.

A04 CTS 3.4.1.2 footnote

states the decay heat removal (DHR) loops normal or emergency power may be inoperable in MODE 5. ITS 3.4.7 has not retained this specific footnote allowance. This changes the CTS by deleting a specific footnote allowance concerning power supplies.

This change is acceptable because the ITS definition of OPERABLE -

OPERABILITY requires an OPERABLE component to have only a normal or an emergency power source. This change to the CTS definition of OPERABLE -

OPERABILITY is discussed in the ITS Section 1.0 Discussion of Changes.

Given this change to the definition of OPERABLE - OPERABILITY, a specific allowance for the DHR loops is not required. This change is designated as an administrative change since it does not result in a technical change to the CTS.

A05 CTS 4.4.1.2.1 states that the required decay heat removal loop(s) shall be determined OPERABLE per Specification 4.0.5, the inservice testing Surveillance Requirements for ASME Code Class 1, 2, and 3 components. ITS 3.4.7 does not contain this explicit Surveillance Requirement. This changes the CTS by deleting the explicit requirement to perform the inservice testing Surveillance Requirements for ASME Code Class 1, 2, and 3 components.

The purpose of CTS 4.4.1.2.1 is to ensure the appropriate inservice testing Surveillance Requirements for ASME Code Class 1, 2, and 3 components are performed for the required decay heat removal loops. The inservice testing requirements of CTS 4.0.5 are retained in ITS 5.5.7, "Inservice Testing Program."

See the Discussion of Changes for ITS 5.5 for any changes to the requirements of CTS 4.0.5. The explicit cross reference is not necessary because when the system is determined to be inoperable when tested in accordance with the inservice testing program, the plant procedures will require the Decay Heat Removal System to be declared inoperable and the appropriate ITS 3.4.7 ACTIONS will be entered when applicable. This change is designated as administrative because it does not result in technical changes to the CTS.

A06 CTS 4.4.1.2.2, in part, specifies the steam generator water level requirements for when the reactor coolant pumps (RCPs) are not operating. ITS LCO 3.4.7 and SR 3.4.7.2 provide the same steam generator water level requirements, but do not state that this level is for when the RCPs are not operating. This changes the CTS by deleting the amplifying information that the RCPs are not operating.

The change is acceptable since the unit is in MODE 5 and the RCPs are not routinely operated in MODE 5, and the ITS 3.4.7 Bases, LCO section, clearly defines the required loop does not include an RCP, only the steam generators.

This change is designated as administrative because it does not result in any technical changes to the CTS.

A07 CTS 3.4.1.2 includes all MODE 5 coolant loop requirements in one Specification.

ITS 3.4.7 includes only the MODE 5, Loops Filled requirements. The MODE 5, Davis-Besse Page 2 of 5 Attachment 1, Volume 9, Rev. 0, Page 137 of 415

Attachment 1, Volume 9, Rev. 0, Page 138 of 415 DISCUSSION OF CHANGES ITS 3.4.7, RCS LOOPS - MODE 5, LOOPS FILLED Loops Not Filled requirements are included in ITS 3.4.8. This changes the CTS by splitting the MODE 5 requirements into two Specifications.

This change is acceptable since all facets of MODE 5 operation are covered in the two ITS Specifications. This change is designated as administrative because it does not result in any technical changes.

MORE RESTRICTIVE CHANGES M01 CTS 3.4.1.2 states the number of coolant loops that shall be OPERABLE, and states that at least one loop must be in operation. This requirement is modified by footnote **that states that the DHR pumps may be de-energized for up to 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , provided certain requirements are met. ITS 3.4.7 does not include this allowance.

The purpose of the 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months allowance is to allow the DHR pump to be removed from operation to perform various activities, such as to place the other DHR pump in service. However, this allowance is not necessary since the CTS already allows natural circulation flow to be used as a means to meet the LCO 3.4.1.2.b requirement that a loop be in operation. This CTS footnote essentially allows all DHR pumps to be de-energized and natural circulation to not be occurring for up to 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . This change is acceptable since it will ensure either a DHR pump is in operation or one RCS loop will be in natural circulation at all times; otherwise ACTIONS to restore a loop to operation will be required.

This change is designated as more restrictive because it will ensure at least one loop is in operation (either a DHR loop with forced flow or an RCS loop with natural circulation flow).

M02 ITS SR 3.4.7.3 requires verification that correct breaker alignment and indicated power are available to each required pump. A Note further explains that the Surveillance 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. This Surveillance is not required by the CTS. This changes the CTS by requiring verification of correct breaker alignment and indicated power availability on required DHR pumps that are not in operation.

The purpose of ITS SR 3.4.7.3 is to ensure a standby pump is available to provide RCS cooling should the operating pump fail. This change is acceptable because the verification of proper breaker alignment and power availability ensures that an additional DHR pump can be placed in operation, if needed, to maintain decay heat removal and reactor coolant circulation. This change is designated as more restrictive because it requires performance of the Surveillance on the non-operating pump.

RELOCATED SPECIFICATIONS None Davis-Besse Page 3 of 5 Attachment 1, Volume 9, Rev. 0, Page 138 of 415

Attachment 1, Volume 9, Rev. 0, Page 139 of 415 DISCUSSION OF CHANGES ITS 3.4.7, RCS LOOPS - MODE 5, LOOPS FILLED REMOVED DETAIL CHANGES LA01 (Type 1 - Removing Details of System Design and System Description, Including Design Limits) CTS 3.4.1.2.a and 3.4.1.2.c contain a description of what constitutes an OPERABLE coolant loop. ITS 3.4.7 does not include this description of what constitutes an OPERABLE coolant loop. This changes the CTS by moving the details of what constitutes an OPERABLE coolant loop 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 a requirement for the RCS loops or decay heat removal 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.

LA02 (Type 3 - Removing ProceduralDetails for Meeting TS Requirements or Reporting Requirements) CTS 4.4.1.2.3 states that the required coolant loops shall be verified to be in operation and circulating reactor coolant at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . ITS SR 3.4.7.1 states that the required DHR loop shall be verified to be in operation every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . This changes the CTS by moving the Surveillance Requirement to verify that the coolant loops are circulating reactor coolant to the Bases.

The removal of this detail for performing Surveillance Requirements from the Technical Specifications is acceptable because this type of information is not necessary to be in the Technical Specifications in order to provide adequate protection of the public health and safety. The ITS retains the requirement that a DHR loop be in operation. This will require recirculation of reactor coolant since the ITS Bases specify that verification that a DHR loop is in operation includes flow rate, temperature, or pump status monitoring, which helps ensure that forced flow is providing heat removal. Also, this change is acceptable because these types of procedural details will be adequately controlled in the ITS Bases.

Changes to the Bases are controlled by the Technical Specification Bases Control Program in Chapter 5. This program provides for the evaluation of changes to ensure the Bases are properly controlled. This change is designated as a less restrictive removal of detail change because procedural details for meeting Technical Specification requirements are being removed from the Technical Specifications.

LESS RESTRICTIVE CHANGES L01 (Category4 - Relaxation of Required Action) CTS 3.4.1.2 Action b states that when no coolant loops are in operation, all operations involving a reduction in boron concentration of the RCS must be suspended. ITS 3.4.7 Required Davis-Besse Page 4 of 5 Attachment 1, Volume 9, Rev. 0, Page 139 of 415

Attachment 1, Volume 9, Rev. 0, Page 140 of 415 DISCUSSION OF CHANGES ITS 3.4.7, RCS LOOPS - MODE 5, LOOPS FILLED Action B.1 states that operations that would cause introduction of coolant into the RCS with boron concentration less than required to meet the requirements of LCO 3.1.1, "SHUTDOWN MARGIN (SDM)," must be suspended. This relaxes the CTS Action by revising the action from suspending reductions in boron concentration to suspending introduction of coolant into the RCS with a boron concentration less than required to meet LCO 3.1.1.

The purpose of CTS 3.4.1.2 Action b is to ensure that "pockets" of coolant with boron concentration less than that required to maintain the SDM are not created when there is no forced or natural circulation flow through the reactor. 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 and the low probability of a DBA occurring during the repair period. As long as coolant with boron concentration less than that required to meet the SDM requirement in LCO 3.1.1 is not introduced into the RCS, there is no possibility of creating "pockets" of coolant with less than the required boron concentration. 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 5 of 5 Attachment 1, Volume 9, Rev. 0, Page 140 of 415

Attachment 1, Volume 9, Rev. 0, Page 141 of 415 Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs)

Attachment 1, Volume 9, Rev. 0, Page 141 of 415

Attachment 1, Volume 9, Rev. 0, Page 142 of 415 CTS RCS Loops - MODE 5, Loops Filled 3.4.7 34 REACTOR COOLANT SYSTEM (RCS) Two loops consisting of any combination of RCS loops and decay heat removal (DHR) loops shall be OPERABLE and one loop shall be in operation.

3.4.7 RCS Loops - MODE 5, Loops Filled 3.4.1.2 LCO 3.4.7

/

One decay hea removal (DHR) loop hall be OPERABL and in operation, and ither:/

a. One adcional DHR loop shall be OPERABLE or 0

b. The seq'ondary side water lev I of each steam ge erator (SG) shall be Ž [5p]%.

1 The DHR ump of the loop in op ration may be re ved from operation or 5 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months per 8 hou period provided:

a. No perations are permitt d that would cause introduction of cool nt into the RCS with oron concentratio less than req ired to meet the SDM of LCO 3.1.1 and

b. Cor outlet temperature i maintained at leas 10°F below sat ration temperature.

0

2. One req ired DHR loop may b inoperable for up t 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for surveilla ce testing provided t at the other DHR Io p is OPERABLE and in o eration.

3. All DH loops may be not in peration during pla ned heatup to MODE when at least one R S loop is in operati n.

APPLICABILITY: MODE 5 with RCS loops filled.

BVVOG STS 3.4.7-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 0, Page 142 of 415

Attachment 1, Volume 9, Rev. 0, Page 143 of 415 CTS All changes are a unless otherwise noted 9 RCS Loops - MODE 5, Loops Filled 3.4.7 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME Action a A. One required FDR loop A.1 Initiate action to restore a Immediately inoperable, second D-TR loop to OPERABLE status.

loop One JHR OPE ABLE. A.2 Initiate actio to restore mine required SC - secondary side water/l evels to within limits. /

B. One or more r quired B.1 Init ate action to restore a Immediately SGs with sec ndary side se nd DHR loop to water level no within 0 ERABLE status.

limit.

OR AND B.2 I itiate action to restore Immediately One DHR I op quired SGs seconda OPERABLE. ide water level to within imit.

Action a. LFK .- required FD-RI loop, 1 Suspend operations that Immediately Action b T IF EABL B would cause introduction of coolant into the RCS with OR [inoperable boron concentration less than required to meet SDM Required PE loop not of LCO 3.1.1*

in operation.

AND "SHUTDOWN MARGIN (SDM).'"

79. Initiate action to restore one Immediately DR loop to OPERABLE status and operation.

BVWOG STS 3.4.7-2 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 0, Page 143 of 415

Attachment 1, Volume 9, Rev. 0, Page 144 of 415 CTS RCS Loops - MODE 5, Loops Filled 3.4.7 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY 4.4.1.2.3 SR 3.4.7.1 Verify required DHR loop is in operation. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months 0

,for each required. RCS loop, 0 4.4.1.2.2 SR 3.4.7.2 Veri

Ž ý5 re ired SG secondary side water level5e[

=35inches above the lower tube sheet E 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months 0

DOC M02 SR 3.4.7.3 ......................- NOTE. . ..---------.

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.

Verify correct breaker alignment and indicated 7 days power available to each required DHR pump.

BWOG STS 3.4.7-3 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 0, Page 144 of 415

Attachment 1, Volume 9, Rev. 0, Page 145 of 415 JUSTIFICATION FOR DEVIATIONS ITS 3.4.7, RCS LOOPS - MODE 5, LOOPS FILLED

1. The Specification has been modified to allow credit for natural circulation flow to meet the LCO requirements. Thus, any combination of DHR and RCS loops can be used to meet both the OPERABLE and in operation requirements, similar to the ITS 3.4.6 requirements. This was approved by the NRC as documented in the Safety Evaluation for Amendment 38. Furthermore, due to this change, the NOTES have been deleted and the ACTIONS have been modified to reflect the natural circulation option. The proposed ACTIONS are consistent with the ACTIONS of ITS 3.4.6, which has similar LCO requirements. In addition, ITS SR 3.4.7.1 and SR 3.4.7.2 have been modified to reflect the natural circulation allowances.

2. The title of the LCO has been provided since this is the first reference to the LCO.

3. Removed brackets and provided plant specific information.

Davis-Besse Page 1 of 1 Attachment 1, Volume 9, Rev. 0, Page 145 of 415

Attachment 1, Volume 9, Rev. 0, Page 146 of 415 Improved Standard Technical Specifications (ISTS) Bases Markup and Justification for Deviations (JFDs)

Attachment 1, Volume 9, Rev. 0, Page 146 of 415

Attachment 1, Volume 9, Rev. 0, Page 147 of 415 RCS Loops - MODE 5, Loops Filled B 3.4.7 B 3.4 REACTOR COOLANT SYSTEM (RCS)

B 3.4.7 RCS Loops - MODE 5, Loops Filled BASES BACKGROUND In MODE 5with RCS loops filled, the primary function of the reactor coolant is the removal of decay heat and transfer of this heat either to the steam generator (SG) secondary side coolant or the component cooling water via the decay heat removal (DHR) heat e.(changer*. While the principal means for decay heat removal is via the DHR System, the SGs Q are specified as a backup means for redundancy. Although the SGs cannot remove heat unless steaming occurs Kwhich i. not possible inI IM-OE 51, they are available as a temporary heat sink and can be used 0

by allowing the RCS to heat up into the temperature region of MODE 4 where steaming can be effective for heat removal. The secondary function of the reactor coolant is to act as a carrier for soluble neutron poison, boric acid.

In MODE 5 with RCS loops filled, DHR loops are the principal means for IIf forced flow Is used to meet the heat removal. The number of loops in operation can vary to suit the operational needs.:[the irfnt of thisILCOs rovielforced flo from at is provided 0

least one DHR loop for decay heat remova an ransport. The flow provided by one DHR loop is adequate for decay heat removal. The other intent of this LCO is to require that a second path be available to INSERT 1 provide redundancy for heat removal.

0 The LCO provides for either SG heat removal or DHR System heat removal. In this MODE, reactor coolant pump (RCP) operation may be restricted because of net positive suction head (NPSH) limitations, and the SG will not be able to provide steam for the turbine driven feed eri.al pumps. However, to ensure that the SGs can be used as a heat sink, a----E

- driven feed*[ pump is needed, because it is independent of steam. Condey sate puml s, startu pumps, or the iotor 4riven au-'-

The Startup Feed Pump eedwater pump can be used. If RCPs are available, the steam generator level need not be adjusted. If RCPs are not available, the water level I-Motor Driven "-

is must be adjusted for natural circulation. The high entry point in the J Feedwater Pump generatorlsh ldb accessible from the Ieed er pumps so that natural further circulation can be'stimulated The SGs are primarily a backup to the S,if needed. - DHR pumps, which are used for forced flow. By requiring the SGs to be a backup heat removal path, the option to increase RCS pressure and temperature for heat removal in MODE 4 is provided.

BVVOG STS B 3.4.7-1 Rev. 3.1, 12/01/05 Attachment 1, Volume 9, Rev. 0, Page 147 of 415

Attachment 1, Volume 9, Rev. 0, Page 148 of 415 B 3.4.7 (D INSERT I Reactor coolant natural circulation is not normally used; however, the natural circulation flow rate is sufficient for core cooling and boron mixing. If entry into natural circulation is required, the reactor coolant at the highest elevation of the hot leg must be maintained subcooled for single phase circulation. When in natural circulation, it is preferable to remove heat using both SGs to avoid idle loop stagnation that might occur if only one SG were in service. One generator will provide adequate heat removal.

Insert Page B 3.4.7-1 Attachment 1, Volume 9, Rev. 0, Page 148 of 415

Attachment 1, Volume 9, Rev. 0, Page 149 of 415 RCS Loops - MODE:5, Loops Filled B 3.4.7 BASES APPLICABLE No safety analysesare performed with initial conditions in MODE 5.

SAFETY ANALYSES Srelated to loss of RCS loops RCS Loops - MODE 5 (Loops Filled) satisfies Criterion 4 of 0

10 CFR 50.36(c)(2)(ii). ED The purpose of this LCO is to require thatlat least

_ _ rre of the DHR loops *.RCroDDR, LCO one of these loops be,"- be OPERABLE

[orboth SGs with/secondary side wateranI~vel e_

and'in operationwIt ditional DHR loop OPF.RABLI

[50,. lOne DHR Ioo1 provides su cie t orce circu ation to prorm e safety functions f the reactor coolant under these conditions. he second DHR loop is n rmally 5 maintained as backup to the operatin DHR loop to provide red, dancy for decay he removal. However, if t e standby DHR loop is not IE2 OPERABLE a sufficient alternate me hod of providing redundany heat removal pa s is to provide both SG with their secondary side Oter levels ? [5 %. Should the operatin DHR loop fail, the SGs co/Id be used to re ove the decay heat.

Note 1 permits the D R pumps to be removed from opertion 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. The circumstances for stoppino both DHR trains are to be limite to situations where: (a) Pressure and temperature increases can be ma ntained well within the allowable p essure (P/T and low temperature ove pressure protection) and 10°F sub ooling limits or (b) Alternate heat p ths through the SGs are in operati n.

The Note prohibits oron dilution with coolant at boron concentrations less than required assure the SDM of LCO 3.1.1 is intained when DHR forced flow is stopped because an even concent ation distribution cannot be ensure Core outlet temperature is to be aintained at least IOF below satura ion temperature so that no vapor bble would form and possibly caus a natural circulation flow obstruct on. In this MODE, the generators ar used as a backup for decay heat emoval and, to 0 ensure their avail bility, the RCS loop flow path is to be maintained with subcooled liquid.

In MODE 5, it is ometimes necessary to stop all R P or DHR pump forced circulatio This is permitted to change ope tion from one DHR train to the othe perform surveillance or startup te ting, perform the transition to an from the DH R System, or to avoid operation below the RCP minimum PSH limit. The time period is a ptable because natural circula on is acceptable for heat removal, he reactor coolant temperature n be maintained subcooled, and b ron stratification affecting rea vity control is not expected.

BWOG STS B 3.4.7-2 Rev. 3.1, 12/01/05 Attachment 1, Volume 9, Rev. 0, Page 149 of 415

Attachment 1, Volume 9, Rev. 0, Page 150 of 415 B 3.4.7 0- INSERT 2 The LCO allows the two loops that are required to be OPERABLE to consist of any combination of RCS or DHR System loops. Any one loop in operation provides enough flow to remove the decay heat from the core with forced or natural circulation. The second loop that is required to be OPERABLE provides a redundant path for heat removal.

Insert Page B 3.4.7-2 Attachment 1, Volume 9, Rev. 0, Page 150 of 415

Attachment 1, Volume 9, Rev. 0, Page 151 of 415 RCS Loops - MODE 5, Loops Filled B 3.4.7 BASES LCO (continued)

Note 2 allows one DIP loop to be inoperable for a peno of up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months provided that the oth r loop is OPERABLE and in oper *on. This permits periodic surveillan c/tests to be performed on the inop rable loop during the only time when uch testing is safe and possible.

Note 3 provides f r an orderly transition from MODE to MODE 4 during a planned heatu by permitting DHR loops to not be in operation when at 0

least one RCP i in operation. This Note provides f r the transition to MODE 4 where an RCP is permitted to be in oper ion and replaces the

,RCS circulatio function provided by the DHR oo;s.

I NET 3 -- ---

L An OPERABLE DHR loop is composed of an OPERABLE DHR pump 0

and an OPERABLE DHR heat e "iana-R. e INSERT 47 0 DHR pumps are OPERABLE if they are capable of being powered and INSERT 5 are able to provide flow if required. A SG can fform as a heat sink-en it as an aduate water evel and is OP RABLE. 00 APPLICABILITY In MODE 5 with loops filled, forced circulation is provided by this LCO to remove decay heat from the core and to provide proper boron mixing.

One loop of DHR provides sufficient circulation for these purposes.

Operation in other MODES is covered by:

LCO 3.4.4, "RCS Loops - MODES 1 and 2ý'

LCO LCO 3.4.5, 3.4.6, "RCS Loops - MODE 3e" "RCS Loops - MODE 0

LCO 3.4.8, "RCS Loops - MODE 5, Loops Not Fille LCO 3.9.4, "Decay Heat Removal (DHR) and Coolant Circulation - High Water Level" 1MOQ.E 6 and 0

LCO 3.9.5, "Decay Heat Removal (DHR) and Coolant Circulation - Low Water Level j 0

ACTIONS A.1. A.2. B-tand B.21 required RCS loopor I'fione DHR loop is OPERABLEand an yrequired SG has se dary side/

ýwater level < [50/] or one required PFIR loop inoperable /redundancy for heat removal is lost. Action must be initiated to restore a secondFD RI loop to OPERABLE statuslor initiate/laction to restore the s/*condary si e&

fwat-er lvel in tyte bus, and action/hiust be taken immedi~ely. Either I Required Actj~n will restore redodant decay heat removal pathsi The immediate Completion Time reflects the importance of maintaining the availability of two paths for heat removal.

BWOG STS B 3.4.7-3 Rev. 3.1, 12101/05 Attachment 1, Volume 9, Rev. 0, Page 151 of 415

Attachment 1, Volume 9, Rev. 0, Page 152 of 415 B 3.4.7 0 INSERT 3 An OPERABLE RCS loop consists of an SG that is OPERABLE. An OPERABLE SG requires > 35 inches of secondary side water level above the lower tube sheet. In addition, the steam generator maximum level must be maintained low enough such that the steam generator remains capable of heat removal by maintaining a steam flow path (i.e., < 625 inches full range level). Furthermore, the SG must be capable of transferring heat from the reactor coolant at a controlled rate.

O INSERT 4 cooler. Furthermore, the two DHR loops share the same suction path through DH-1 1 and DH-12. Therefore, when both DHR loops are being used to meet the LCO requirements, control power is required to be removed from DH-11 and DH-12 valve operators, or manual valves DH-21 and DH-23 are required to be open.

O INSERT 5 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 3.4.7-3 Attachment 1, Volume 9, Rev. 0, Page 152 of 415

Attachment 1, Volume 9, Rev. 0, Page 153 of 415 RCS Loops - MODE 5, Loops Filled B 3.4.7 BASES ACTIONS (continued)

If no required R] loop is in operation, [excep( as proyided inNote 1,1 or no required DRl loop is OPERABLE, all operations involving introduction of coolant into the RCS with boron concentration less than required to meet the minimum SDM of LCO 31 -1 must be suspended and action to restore a D-R loop to OPERABLE status and operation must be initiated.

0 The required margin to criticality must not be reduced in this type of operation. Suspending the introduction of coolant into the RCS of coolant with boron concentration less than required to meet the minimum SDM of LCO 3.1.1 is required to assure continued safe operation. With coolant added without circulation, unmixed coolant could be introduced to the core, however coolant added with boron concentration meeting the minimum SDM maintains acceptable margin to subcritical operations.

The immediate Completion Time reflects the importance of maintaining operation for decay heat removal.

SURVEILLANCE SR 3.4.7.1 RS2 REQUIREMENTS This SR requires in operation. verification Verification every flow includes 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months that the require rate, temperature, DHR statuLs or pump loop is [o aua circulation monitoring, which help ensure that forcedfow is providing heat removal, The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Frequency has been shown by operating practice to be sufficient to regularly assess degradation. In addition, control room indication and alarms will normally indicate loop status.

SR 3.4.7.2 required Verifying the SGs are OPERABLE by ensuring their secondary side water 35 inches above the lower tube sheet levels are >1[5N %ensures that redundant heat removal paths are availablelif the se nd DHR loop is rt OPERABLE. If both DHR loops are OPERABLE, this Surveillance is not needed. The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Frequency 0 has been shown by operating practice to be sufficient to regularly assess degradation and verify operation Within §dfety analyses gssumptions 0

BWOG STS B 3.4.7-4 Rev. 3.1, 12/01/05 Attachment 1, Volume 9, Rev. 0, Page 153 of 415

Attachment 1, Volume 9, Rev. 0, Page 154 of 415 RCS Loops - MODE 5, Loops Filled B 3.4.7 BASES SURVEILLANCE REQUIREMENTS (continued)

SR 3.4.7.3 Verification that each required DHR pump :is OPERABLE ensures that redundant paths for heat removal are available. The requirement also ensures that the additional loop can be placed in operation if needed to maintain decay heat removal and reactor coolant circulation. If the 35 inches above the lower tube sheet condary side water level is 1-5/ oin both SGs, this Surveillance is not needed. Verification is performed by verifying proper breaker alignment 0 and power available to each required pump. Alternatively, verification that a pump is in operation also verifies proper breaker alignment and power availability. The Frequency of 7 days is considered reasonable in view of other administrative controls available and has been shown to be acceptable by operating experience.

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.

REFERENCES None.

BWOG STS B 3.4.7-5 Rev. 3.1, 12/01/05 Attachment 1, Volume 9, Rev. 0, Page 154 of 415

Attachment 1, Volume 9, Rev. 0, Page 155 of 415 JUSTIFICATION FOR DEVIATIONS ITS 3.4.7 BASES, RCS LOOPS - MODE 5, LOOPS FILLED

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

2. Changes are made (additions, deletions, and/or changes) to the ISTS Bases which reflect the plant specific nomenclature, number, reference, system description, analysis, or licensing basis description.

3. 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 made to be consistent with the Specification.

5. Changes have been made to allow natural circulation flow to meet the LCO requirements. In addition, due to these changes, other associated changes to the NOTES, ACTIONS, and Surveillances have been made to be consistent with changes made to the Specification.

Davis-Besse Page 1 of 1 Attachment 1, Volume 9, Rev. 0, Page 155 of 415

Attachment 1, Volume 9, Rev.

0, Page 156 of 415 Specific No Significant Hazards Considerations (NSHCs)

Attachment 1, Volume 9, Rev. 0, Page 156 of 415

Attachment 1, Volume 9, Rev. 0, Page 157 of 415 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.4.7, RCS LOOPS - MODE 5, LOOPS FILLED There are no specific NSHC discussions for this Specification.

Davis-Besse Page 1 of 1 Attachment 1, Volume 9, Rev. 0, Page 157 of 415

Attachment 1, Volume 9, Rev. 0, Page 158 of 415 ATTACHMENT 8 ITS 3.4.8, RCS LOOPS - MODE 5, LOOPS NOT FILLED Attachment 1, Volume 9, Rev. 0, Page 158 of 415

, Volume 9, Rev. 0, Page 159 of 415 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) , Volume 9, Rev. 0, Page 159 of 415

Attachment 1, Volume 9, Rev. 0, Page 160 of 415 ITS 3.4.8 IT.S 3/4.- FL4

-OR COOLTST STYS,1rM'

a. At least tw cocloamc loops [hall sha1 be LCO 3.4.8 OPERABLE:

1.* Reactor Coolant Loop 1 &Md Its associated steam See ITS genertor,3.4.5, ITS 3.4.6.

and

2. Reactor Coolant Loop 2 and *ts assOciated stemr - ITS 3.4.7

-* g.e.iaraoT'.

ACTION A ACTION B LCO 3.4.8 NOTE 1 provided (1) no operations are pen the reactor coolant system boron ci tPmperature is maintained at least Page 1 of 2 Attachment 1, Volume 9, Rev. 0, Page 160 of 415

Attachment 1, Volume 9, Rev. 0, Page 161 of 415 ITS 3.4.8 ITS 3/4.4 REACTOR COOLANT SYSTDI A06 SURVEILLANCE REOUTREMENTS 1The reuired decay heat renol1 loop(s) shall be dereined OPERABLE FseeTFi l3.4.5, J above 4 the lover rover

.1.213 tube sheet onceinhes per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months vete if an associated tut s reaLctor per tonce cooleant\

2L /E. 3.4.5 andl IS3.4.7 ours if no reactor coolant pumps are operati See3.ITS 4.4.1.2.3 At least one coolant loop shall be verified to be in operation and -.

SR 3.4.8.1 circub a re actor ouent at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

Add proosed S33.48.

4.4..2. At eas oopshal onecooantbeveriiedto b inopertioMan DAVIS-BESSE, UNIT 1 3/4 4-2a Aaendment No. Ap, 13 5 Page 2 of 2 Attachment 1, Volume 9, Rev. 0, Page 161 of 415

Attachment 1, Volume 9, Rev. 0, Page 162 of 415 DISCUSSION OF CHANGES ITS 3.4.8, RCS LOOPS - MODE 5, LOOPS NOT FILLED 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 3.4.1.2.d states that the provisions of Specifications 3.0.3 and 3.0.4 are not applicable. ITS 3.4.8 does not include this exception. This changes the CTS by deleting the specific exception to Specifications 3.0.3 and 3.0.4.

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 5, which is the Applicability of ITS 3.4.8. Therefore, this exception to CTS 3.0.3 is redundant and unnecessary.

CTS 3.0.4 provides requirements to preclude changing MODES with inoperable equipment. However, ITS LCO 3.0.4 has been modified to allow MODE changes under certain circumstances. This is justified in the Discussion of Changes for ITS Section 3.0. Therefore, this specific exception to CTS 3.0.4 is not needed in the ITS. This change is designated as administrative because it does not result in a technical change to the CTS.

A03 CTS 3.4.1.2 includes all MODE 5 coolant loop requirements in one Specification.

ITS 3.4.8 includes only the MODE 5, Loops Not Filled requirements. The MODE 5, Loops Filled requirements are included in ITS 3.4.7. This changes the CTS by splitting the MODE 5 requirements into two Specifications.

This change is acceptable since all facets of MODE 5 operation are covered in the two ITS Specifications. This change is designated as administrative because it does not result in any technical changes.

A04 CTS 3.4.1.2 Action a states that when less than the required reactor coolant loops are OPERABLE, action must be immediately initiated to restore the required loops. CTS 3.4.1.2 Action b states that when no coolant loops are in operation, all operations involving a reduction in boron concentration of the RCS must be suspended and action must be immediately initiated to return the required loop to operation. ITS 3.4.8 ACTION A specifies the Required Actions when one of the two required DHR loops is inoperable. Required Action A.1 is to immediately initiate action to restore the DHR loop to OPERABLE status.

ITS 3.4.8 ACTION B specifies the Required Actions when two required DHR loops are inoperable and when no required DHR loop is in operation. The Required Actions are to immediately suspend operations that would cause introduction of coolant into the RCS with boron concentration less than required to meet the requirements of LCO 3.1.1, and to immediately initiate action to restore one DHR loop to OPERABLE status and operation. This changes the CTS by revising the Actions to immediately suspend operations that would cause introduction of coolant into the RCS with boron concentration less than required Davis-Besse Page 1 of 6 Attachment 1, Volume 9, Rev. 0, Page 162 of 415

Attachment 1, Volume 9, Rev. 0, Page 163 of 415 DISCUSSION OF CHANGES ITS 3.4.8, RCS LOOPS - MODE 5, LOOPS NOT FILLED to meet the requirements of LCO 3.1.1 when two required DHR loops are inoperable and to break up the Actions for one and two inoperable required DHR loops into two separate Actions.

This change is acceptable because it results in no technical changes to the CTS.

When both required DHR loops are inoperable, in all likelihood no DHR loops will be in operation. With no DHR loops in operation at the same time as both required DHR loops are inoperable, the same ITS ACTION (ACTION B) would be required. Therefore, since ITS 3.4.8 ACTION B would also require entry when no DHR loops are in operation, the identical actions would be required (i.e.,

immediately suspend operations that would cause introduction of coolant into the RCS with boron concentration less than required to meet the requirements of LCO 3.1.1). This change is designated as administrative because it does not result in any technical changes to the CTS.

A05 CTS 3.4.1.2 footnote

states the decay heat removal (DHR) loops normal or emergency power may be inoperable in MODE 5. ITS 3.4.8 has not retained this specific footnote allowance. This changes the CTS by deleting a specific footnote allowance concerning power supplies.

This change is acceptable because the ITS definition of OPERABLE -

OPERABILITY requires an OPERABLE component to have only a normal or an emergency power source. This change to the CTS definition of OPERABLE -

OPERABILITY is discussed in the ITS Section 1.0 Discussion of Changes.

Given this change to the definition of OPERABLE - OPERABILITY, a specific allowance for the DHR loops is not required. This change is designated as an administrative change since it does not result in a technical change to the CTS.

A06 CTS 4.4.1.2.1 states that the required decay heat removal loop(s) shall be determined OPERABLE per Specification 4.0.5, the inservice testing Surveillance Requirements for ASME Code Class 1, 2, and 3 components. ITS 3.4.8 does not contain this explicit Surveillance Requirement. This changes the CTS by deleting the explicit requirement to perform the inservice testing Surveillance Requirements for ASME Code Class 1, 2, and 3 component.

The purpose of CTS 4.4.1.2.1 is to ensure the appropriate inservice testing Surveillance Requirements for ASME Code Class 1, 2, and 3 components are performed for the required decay heat removal loops. The inservice testing requirements of CTS 4.0.5 are retained in ITS 5.5.7, "Inservice Testing Program."

See the Discussion of Changes for ITS 5.5 for any changes to the requirements of CTS 4.0.5. The explicit cross reference is not necessary because when the system is determined to be inoperable when tested in accordance with the inservice testing program, the plant procedures will require the Decay Heat Removal System to be declared inoperable and the appropriate ITS 3.4.8 ACTIONS will be entered when applicable. This change is designated as administrative because it does not result in technical changes to the CTS.

Davis-Besse Page 2 of 6 Attachment 1, Volume 9, Rev. 0, Page 163 of 415

Attachment 1, Volume 9, Rev. 0, Page 164 of 415 DISCUSSION OF CHANGES ITS 3.4.8, RCS LOOPS - MODE 5, LOOPS NOT FILLED MORE RESTRICTIVE CHANGES M01 CTS 3.4.1.2 footnote **contains an allowance for the decay heat removal pumps to be de-energized for up to one hour. ITS LCO 3.4.8 Note 1 allows all DHR pumps to be removed from operation for _ 15 minutes only when switching from one loop to the other, and also requires that no draining operations to further reduce the RCS water volume are permitted (part c). This changes the CTS by reducing the time allowed for the DHR pump to be de-energized from 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months to 15 minutes, restricts the allowance to only pump switching operations, and adds a restriction that no draining operations are permitted to further reduce the RCS water volume.

The purpose of the CTS 3.4.1.2 footnote ** in MODE 5 with loops not filled is to allow the DHR loops to be switched from one to the other. This change is acceptable because ITS LCO 3.4.8 Note 1 provides sufficient time to perform loop switching operations and provides adequate controls. Stopping all operating DHR loops when the RCS is not filled should be limited to short periods of time because of the reduced inventory of water available to absorb decay heat.

Stopping all DHR pumps during loop swapping operations may be necessary.

Fifteen minutes is sufficient time to perform the loop swapping operation without excessive increases in RCS average temperature due to lack of decay heat removal. Adding the additional condition that no draining operations be performed when the pumps are stopped is reasonable given the low RCS water level and the unavailability of the DHR pumps to add inventory to the RCS, if needed. This change is more restrictive because it reduces the time a DHR loop may be out of service and adds an additional restriction.

M02 CTS 3.4.1.2 footnote ** part (2) allows the DHR pumps to be de-energized provided the core outlet temperature is maintained at least 10OF below saturation temperature. ITS LCO 3.4.8 Note 1 provides a similar allowance, but requires the maximum RCS temperature to be < 190 0 F. This changes the CTS by requiring the RCS temperature to be < 190OF instead of 10°F below saturation temperature.

The purpose of CTS 3.4.1.2 footnote ** part 2 is to help ensure the RCS temperature does not reach the boiling point. With the RCS loops not filled, the RCS pressure would be at atmospheric pressure. Thus 10°F below saturation temperature is 2020 F. This change is acceptable because the proposed change increases the margin to the boiling point since it requires the maximum RCS temperature be < 1901F. Furthermore, the 190OF limit is 10OF below the MODE 5 to MODE 4 transition temperature of 200 0 F. This change is more restrictive because it requires the unit to be maintained at a lower RCS temperature when the required DHR pump is not in operation.

M03 ITS SR 3.4.8.2 requires verification that correct breaker alignment and indicated power are available to each required pump. A Note further explains that the Surveillance 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. This Surveillance is not required by the CTS. This changes the CTS by requiring verification of correct breaker alignment and indicated power availability on required DHR pumps that are not in operation.

Davis-Besse Page 3 of 6 Attachment 1, Volume 9, Rev. 0, Page 164 of 415

Attachment 1, Volume 9, Rev. 0, Page 165 of 415 DISCUSSION OF CHANGES ITS 3.4.8, RCS LOOPS - MODE 5, LOOPS NOT FILLED The purpose of ITS SR 3.4.8.2 is to ensure a standby pump is available to provide RCS cooling should the operating pump fail. This change is acceptable because the verification of proper breaker alignment and power availability ensures that an additional DHR pump can be placed in operation, if needed, to maintain decay heat removal and reactor coolant circulation. This change is designated as more restrictive because it requires performance of the Surveillance on the non-operating pump.

RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES LA01 (Type 1 - Removing Details of System Design and System Description, Including Design Limits) CTS 3.4.1.2.a and 3.4.1.2.c contain a description of what constitutes an OPERABLE coolant loop. ITS 3.4.8 does not include this description of what constitutes an OPERABLE coolant loop. This changes the CTS by moving the details of what constitutes an OPERABLE coolant loop 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 a requirement for the RCS 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.

LA02 (Type 3 - Removing ProceduralDetails for Meeting TS Requirements or Reporting Requirements) CTS 4.4.1.2.3 states that the required coolant loops shall be verified to be in operation and circulating reactor coolant at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . ITS SR 3.4.8.1 states that the required DHR loop shall be verified to be in operation every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . This changes the CTS by moving the Surveillance Requirement to verify that the coolant loops are circulating reactor coolant to the Bases.

The removal of this detail for performing Surveillance Requirements from the Technical Specifications is acceptable because this type of information is not necessary to be in the Technical Specifications in order to provide adequate protection of the public health and safety. The ITS retains the requirement that a DHR loop be in operation. This will require recirculation of reactor coolant since the ITS Bases specify that verification that a DHR loop is in operation includes flow rate, temperature, or pump status monitoring, which helps ensure that forced flow is providing heat removal. Also, this change is acceptable because these Davis-Besse Page 4 of 6 Attachment 1, Volume 9, Rev. 0, Page 165 of 415

Attachment 1, Volume 9, Rev. 0, Page 166 of 415 DISCUSSION OF CHANGES ITS 3.4.8, RCS LOOPS - MODE 5, LOOPS NOT FILLED types of procedural details will be adequately controlled in the ITS Bases.

Changes to the Bases are controlled by the Technical Specification Bases Control Program in Chapter 5. This program provides for the evaluation of changes to ensure the Bases are properly controlled. This change is designated as a less restrictive removal of detail change because procedural details for meeting Technical Specification requirements are being removed from the Technical Specifications.

LESS RESTRICTIVE CHANGES L01 (Category I - Relaxation of LCO Requirements) CTS 3.4.1.2 places OPERABILITY requirements for the DHR loops to be OPERABLE and operating.

ITS 3.4.8 specifies the same requirements; however, a new allowance is provided. ITS LCO 3.4.8 Note 2 allows one of the required DHR loops 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 the other DHR loop is OPERABLE and in operation. This changes the CTS by adding this new allowance.

The purpose of CTS LCO 3.4.1.2 is to ensure there is sufficient forced circulation to provide forced flow for decay heat removal and transport. This change is acceptable because the LCO requirements continue to ensure that the structures, systems, and components are maintained consistent with the UFSAR analyses and licensing basis. This allowance provided by ITS 3.4.8 Note 2 still ensures a DHR loop is OPERABLE and in operation. Thus, decay heat removal and transport is still provided during this 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months time period. 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 1 - Relaxation of RequiredAction) CTS LCO 3.4.1.2 footnote **,in part, states that all decay heat removal (DHR) pumps may be de-energized for up to 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months provided no operations are permitted that would cause dilution of the Reactor Coolant System boron concentration. CTS 3.4.1.2 Action b states that when no coolant loops are in operation, all operations involving a reduction in boron concentration of the RCS must be suspended. The ITS LCO 3.4.8 Note 1 allows all DHR pumps to be removed from operation for a certain period of time provided no operations are permitted that would cause introduction of coolant into the RCS with boron concentration less than required to meet the requirements of LCO 3.1.1, "SHUTDOWN MARGIN (SDM)." ITS 3.4.8 Required Action B.1 states that operations that would cause introduction of coolant into the RCS with boron concentration less than required to meet the requirements of LCO 3.1.1 must be suspended. This relaxes the CTS Action and LCO footnote by revising the action and footnote from suspending reductions in boron concentration to suspending introduction of coolant into the RCS with a boron concentration less than required to meet LCO 3.1.1.

The purpose of the CTS LCO 3.4.1.2 footnote ** and CTS 3.4.1.2 Action b is to ensure that "pockets" of coolant with boron concentration less than that required to maintain the SDM are not created when there is no forced flow through the reactor. This change is acceptable because the Required Actions are used to establish remedial measures that must be taken in response to the degraded Davis-Besse Page 5 of 6 Attachment 1, Volume 9, Rev. 0, Page 166 of 415

Attachment 1, Volume 9, Rev. 0, Page 167 of 415 DISCUSSION OF CHANGES ITS 3.4.8, RCS LOOPS - MODE 5, LOOPS NOT FILLED 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 and the low probability of a DBA occurring during the repair period. As long as coolant with boron concentration less than that required to meet the SDM requirement in LCO 3.1.1 is not introduced into the RCS, there is no possibility of creating "pockets" of coolant with less than the required boron concentration. 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 6 of 6 Attachment 1, Volume 9, Rev. 0, Page 167 of 415

Attachment 1, Volume 9, Rev. 0, Page 168 of 415 Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs)

Attachment 1, Volume 9, Rev. 0, Page 168 of 415

Attachment 1, Volume 9, Rev. 0, Page 169 of 415 CTS RCS Loops - MODE 5, Loops Not Filled 3.4.8 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.8 RCS Loops - MODE 5, Loops Not Filled 3-4.1.2 LCO 3.4.8 Two decay heat removal (DHR) loops shall be OPERABLE and one DHR loop shall be in operation.

- - - - ------ j 3.4.1.2 I All DHR pumps may be removed from operation for _<15 minutes footnote -. when switching from one loop to another provided:

E]a. The maximum RCS temperatui "Fj ( ()

b. No operations are permitted that would cause introduction of coolant into the RCS with boron concentration less than required to meet the SDM of LCO 3.1rA and

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

DOC L01 2. One DHR loop may be inoperable for _2 hours for urveillance testing provided that the other DHR loop is OPERABLE and in 0

operation.

APPLICABILITY: MODE 5 with RCS loops not filled.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME Action a A. One required DHR loop A.1 Initiate action to restore Immediately inoperable. DH R loop to OPERABLE status.

BVWOG STS 3.4.8-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 0, Page 169 of 415

Attachment 1, Volume 9, Rev. 0, Page 170 of 415 CTS RCS Loops - MODE 5, Loops Not Filled 3.4.8 ACTIONS (continued)

CONDITION REQUIREDACTION COMPLETION TIME Action a, B. No required DHR loop B.1 Suspend operations that Immediately Action b OPERABLE. would cause introduction of coolant into the RCS with OR boron concentration less than required to meet SDM Required DHR loop not of LCO 3.1.1.

in operation.

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

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY 4.4.12.3 SR 3.4.8.1 Verify required DHR loop is in operation. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months DOC M03 SR 3.4.8.2 ----------- NOTE -----

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.

Verify correct breaker alignment and indicated 7 days power available to each required DHR pump.

BWOG STS 3.4.8-2 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 0, Page 170 of 415

Attachment 1, Volume 9, Rev. 0, Page 171 of 415 JUSTIFICATION FOR DEVIATIONS ITS 3.4.8, RCS LOOPS - MODE 5, LOOPS NOT FILLED

1. Removed brackets and provided plant specific limit.

2. 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.

3. Typographical error corrected.

4. The title of the LCO has been provided since this is the first reference to the LCO.

Davis-Besse Page 1 of 1 Attachment 1, Volume 9, Rev. 0, Page 171 of 415

Attachment 1, Volume 9, Rev. 0, Page 172 of 415 Improved Standard Technical Specifications (ISTS) Bases Markup and Justification for Deviations (JFDs)

Attachment 1, Volume 9, Rev. 0, Page 172 of 415

Attachment 1, Volume 9, Rev. 0, Page 173 of 415 RCS Loops -:MODE 5, Loops Not Filled B 3.4.8 B 3.4 REACTOR COOLANT SYSTEM (RCS)

B 3.4.8 RCS Loops - MODE 5, Loops Not Filled BASES BACKGROUND In MODE 5 with loops not filled, the primary function of the reactor coolant is the removal of decay heat and transfer of this heat to the decay heat I coles Iremoval (DHR) 1eat an er. The steam generators (SGs) are not available as a heat sink when the loops are not filled. The secondary 0 function of the reactor coolant is to act as a carrier for the soluble neutron poison, boric acid.

Loops are not filled when thelreacW FRCS draining is initiated (hot legs not completely filled).

Additionally, the RCS inventory is further reduced to a water levelgT~iwithin the 0

Lj horizontal portion of the hot leg'as might be the case for refueling or maintenance on the reactor coolant pumps or SGs. GL 88-17 (Ref. 1) expresses concerns for loss of decay heat removal for this operating condition. With water at this low level, the margin above the decay heat suction piping connection to the hot leg is small. The possibility of loss of level or inlet vortexing exists and if it were to occur, the operating DHR pump could become air bound and fail resulting in a loss of forced flow for heat removal. As a consequence the water in the core will heat up and could boil with the possibility of core uncovering due to boil off. Because the containment hatch may be open at this time, a pathway to the outside for fission product release exists if core damage were to occur.

In MODE 5 with loops not filled, only DHR pumps can be used for coolant circulation. The number of pumps in operation can vary to suit the operational needs- The intent of this LCO is to provide forced flow from at least one DHR pump for decay heat removal and transport, to require that two paths be available to provide redundancy for heat removal.

APPLICABLE No safety analyses are performed with initial conditions in MODE 5 with SAFETY loops not filled. The flow provided by one DHR pump is adequate for ANALYSES heat removal and for boron mixing.

RCS Loops - MODE 5 (Loops Not Filled) satisfies Criterion 4 of 10 CFR 50.36(c)(2)(ii).

LCO The purpose of this LCO is to require that a minimum of two DH R loops be OPERABLE and that one of these loops be in operation. An OPERABLE loop is one that has the capability of transferring heat from the reactor coolant at a controlled rate. Heat cannot be removed via the DHR ,4stem unless forced flow is used. A minimum of one running decay heat removal pump meets the LCO requirement for one loop in operation. An additional DHR loop is required to be OPERABLE to provide redundancy for heat removal.

BWOG STS B 3.4.8-1 Rev. 3.0, 03131/04 Attachment 1, Volume 9, Rev. 0, Page 173 of 415

Attachment 1, Volume 9, Rev. 0, Page 174 of 415 RCS Loops - MODE:5, Loops Not Filled B 3.4.8 BASES LCO (continued) 4 Note 1 permits the DHR .pumps to be removed from operation for

[ [-*-15 minutes when switching from one tr in o the other. The circumstances for stopping both DHR pumps are to be limited to situations where the outage time is shortland temperature is maintained ()

a 60] F0. The Note prohibits boron dilution with coolant at boron concentrations less than required to s draining operations%

the SDM of LCO 3.1.1 [

forced ow s o e

-. that could reduce the RCS water volume 0

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 that the other loop is OPERABLE and in operation. This permits periodic surveillance tests to be performed on the inoperable loop during the only time when these tests are safe and possible.

An OPERABLE DHR loop is composed of an OPERABLE DHR pump capable of providing forced flow to an OPERABLE DHR neat p)chane INSERT I

DHR pumps are OPERABLE if they are capable of being powered and are able to provide flow if required. INSE J APPLICABILITY In MODE 5 with loops not filled, this LCO requires core heat removal and coolant circulation by the DHR System.

Operation in other MODES is covered by:

LCO 3.4.4, "RCS Loops- MODES land0 LCO 3.4.5, "RCS Loops - MODE LCO 3.4.6, "RCS Loops - MODE LCO 3.4.7, "RCS Loops - MODE 5, Loops Fille LCO 3.9.4, "Decay Heat Removal (DHR) and Coolant Circulation High Water Level" M E6and0 LCO 3.9.5, "Decay Heat Removal (DHR) and Coolant Circulation - Low Water Lever KM ACTIONS A.1 Var i If one required DHR loop is inoperable, redundancy for heat removal is lost. Required Action A.1 is to immediately initiate activities to restore a second loop to OPERABLE status. The immediate Completion Time reflects the importance of maintaining the availability of two paths for heat removal.

BVVOG STS B 3.4.8-2 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 0, Page 174 of 415

Attachment 1, Volume 9, Rev. 0, Page 175 of 415 B 3.4.8 0 INSERT 1 Furthermore, the two DHR loops share the same suction path through DH-1 1 and DH-12. Therefore, when both DHR loops are being used to meet the LCO requirements, control power is required to be removed from DH-11 and DH-12 valve operators, or manual valves DH-21 and DH-23 are required to be open.

0 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 3.4.8-2 Attachment 1, Volume 9, Rev. 0, Page 175 of 415

Attachment 1, Volume 9, Rev. 0, Page 176 of 415 RCS Loops - MODE 5, Loops Not Filled B 3.4.8 BASES ACTIONS (continued)

B.1 and B.2 If no required loop is OPERABLE or the required loop is not in operation except as provided by Note 1 in the LCO, the Required Actionirequire0 immediate suspension of all operations involving introduction of coolant into the RCS with boron concentration less than required to meet the _,>

minimum SDM of LCO 3.1.1 and I initiation of action to limrosdia restore one DHR loop to OPERABLE status and operation.

The Req - ed Action for restoraion does not applyý6 the condition of both loops not in operation When the exception Nte in the LCO is in i forcLr. Suspending the introduction of coolant into the RCS of coolant with 0

boron concentration less than required to meet the minimum SDM of LCO 3.1.1 is required to assure continued safe operation. With coolant added without forced circulation, unmixed coolant could be introduced to the core, however coolant added with boron concentration meeting the minimum SDM maintains acceptable margin to subcritical operations.

The immediate Completion Time reflects the importance of maintaining operations for decay heat removal. The action to restore must continue until one loop is restored.

SURVEILLANCE SR 3.4.8.1 REQUIREMENTS This Surveillance requires verification every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months that the required loop is in operation. Verification includes flow rate, temperature, or pump status monitoring, which help ensure that forced flow is providing heat removal. The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months interval has been shown by operating practice to be sufficient to regularly assess degradation and verify operation within safety analyses assumptions.

SR 3.4.8.2 Verification that each required pump is OPERABLE ensures that redundancy for heat removal is provided. The requirement also ensures that an additional loop 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 each required pump. Alternatively, verification that a pump is in operation also verifies proper breaker alignment and power availability. The Frequency of 7 days is considered reasonable in view of other administrative controls available and has been shown to be acceptable by operating experience.

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.

BWOG STS B 3.4.8-3 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 0, Page 176 of 415

Attachment 1, Volume 9, Rev. 0, Page 177 of 415 RCS Loops - MODE 5, Loops Not Filled B 3.4.8 BASES REFERENCES 1. Generic Letter 88-17, October 17, 1988.

BWOG STS B 3.4.8-4 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 0, Page 177 of 415

Attachment 1, Volume 9, Rev. 0, Page 178 of 415 JUSTIFICATION FOR DEVIATIONS ITS 3.4.8 BASES, RCS LOOPS - MODE 5, LOOPS NOT FILLED

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. 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.

3. Typographical error corrected.

4. Changes made to be consistent with the Specification.

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

6. This description is not necessary. When using the Note allowance, ACTION B is not required to be entered (as described in the first sentence of ACTIONS B.1 and B.2 Bases). In addition, the deleted wording implies that only Required Action B.2 does not apply, when in actuality, neither of the Required Actions apply.

Davis-Besse Page 1 of 1 Attachment 1, Volume 9, Rev. 0, Page 178 of 415

Attachment 1, Volume 9, Rev. 0, Page 179 of 415 Specific No Significant Hazards Considerations (NSHCs)

Attachment 1, Volume 9, Rev. 0, Page 179 of 415

Attachment 1, Volume 9, Rev. 0, Page 180 of 415 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.4.8, RCS LOOPS - MODE 5, LOOPS NOT FILLED There are no specific NSHC discussions for this Specification.

Davis-Besse Page 1 of 1 Attachment 1, Volume 9, Rev. 0, Page 180 of 415

, Volume 9, Rev. 0, Page 181 of 415 ATTACHMENT 9 ITS 3.4.9, PRESSURIZER , Volume 9, Rev. 0, Page 181 of 415

, Volume 9, Rev. 0, Page 182 of 415 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) , Volume 9, Rev. 0, Page 182 of 415

Attachment 1, Volume 9, Rev. 0, Page 183 of 415 Ao ITS 3.4.9 ITS REACTOR COOLANT SYSTEM PRESSURIZER UMITING C0IM-ON1R OPRAB-ON LCO 3.4.9 "A4 Tim pressuzer shal be OPERABLE wilt LCO 3.4.9.a b. Awater levelI 45 1nd228 Inches.

IadAdd proposed LCO 3.4.9.b)

APMAB Y MOE APPUCABIUTY: MODES 1 and 3p I Add proposed MODE 3 Applicabi ACTION A-ACTION 1 -

n Add proposed ACTIONS C and Dt M(

SURVEIL.ANCE JIREMENTS., ,

SR 3.4.9.1 4AA The pressudar shal be demonstrated OPERABLE by vefyin pressurizer level to be witn limits at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

Add proposed SR 3.4.9.2 M*

Amendment No. 255 DAVIS-BESME UNIT 1 3/4 4-5 Page 1 of 1 Attachment 1, Volume 9, Rev. 0, Page 183 of 415

Attachment 1, Volume 9, Rev. 0, Page 184 of 415 DISCUSSION OF CHANGES ITS 3.4.9, PRESSURIZER 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 3.4.4.a states that the pressurizer shall be OPERABLE with a steam bubble.

ITS 3.4.9 does not retain this requirement. This changes the CTS by not specifically requiring the pressurizer to be OPERABLE with a steam bubble.

This change is acceptable because when the unit is in MODE 1, 2, or 3 and the pressurizer water level is maintained at less than 228 inches, a steam bubble will exist. Since the ITS still requires the pressurizer water level to be less than 228 inches, a steam bubble will be present and there is no need to specifically require the steam bubble. The change is designated as administrative because it does not result in a technical change to the CTS.

A03 CTS 3.4.4 Action states that if the inoperable pressurizer is not restored to OPERABLE status within the allowed time, to be in HOT STANDBY (MODE 3) with the control rod drive trip breakers open within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . Under similar conditions, ITS 3.4.9 ACTION B states to be in MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in MODE 4 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . This changes the CTS by eliminating the requirement to open the control rod drive trip breakers. The change associated with entering MODE 4 is discussed in DOC M02.

This change is acceptable because it results in no technical change to the Technical Specifications. Although CTS 3.4.4 Action appears to require the control rod drive trip breakers to be opened within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months (if the pressurizer is not restored to OPERABLE status within the allowed restoration time), they are not actually required to be opened. The Applicability of CTS 3.4.4 is MODES 1 and 2. CTS 3.0.1 states that "Limiting Conditions for Operation and ACTION requirements shall be applicable during the OPERATIONAL MODES or other conditions specified for each specification." Therefore, the CTS 3.4.4 Action to open the control rod drive trip breakers ceases to be applicable once the unit enters MODE 3, and the Action is exited. As a result, deleting this action results in no operational difference from the CTS Action. This change is designated as administrative because it does not result in a technical change to the CTS.

MORE RESTRICTIVE CHANGES M01 CTS 3.4.4 does not contain requirements for the pressurizer heaters. ITS LCO 3.4.9.b has been added requiring the pressurizer to be OPERABLE with a minimum of 112 kW of essential pressurizer heaters OPERABLE. ITS 3.4.9 ACTIONS C and D have been added to provide compensatory measures when the new requirement is not met. ITS 3.4.9 ACTION C, which applies when the Davis-Besse Page 1 of 3 Attachment 1, Volume 9, Rev. 0, Page 184 of 415

Attachment 1, Volume 9, Rev. 0, Page 185 of 415 DISCUSSION OF CHANGES ITS 3.4.9, PRESSURIZER capacity of pressurizer heaters is less than 150 kW, requires restoration of the essential pressurizer heater capability within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . If the heater capability is not restored within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , ITS 3.4.9 ACTION D requires the unit to be in MODE 4 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . In addition, SR 3.4.9.2 has been added, and requires verification that the essential pressurizer heater capacity is greater than or equal to 150 kW every 24 months.

This change is acceptable because the pressurizer heaters are used to maintain the steam and water at the saturation temperature corresponding to the desired RCS pressure. The addition of the LCO, ACTIONS and Surveillance Requirement will assure that this capability is available. This change is designated as more restrictive because additional LCO requirement and associated ACTIONS and a Surveillance Requirement have been added.

M02 CTS 3.4.4 only requires the pressurizer to be OPERABLE in MODES 1 and 2. If the pressurizer is inoperable, the CTS Actions allows 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months to restore the pressurizer to OPERABLE status or the unit must be in HOT STANDBY (MODE 3) with the control rod drive trip breakers open within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

ITS 3.4.9 requires the pressurizer to be OPERABLE in MODES 1, 2, and 3. If the pressurizer is not restored to OPERABLE status under the same conditions as the CTS (water level not within limit) within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , the unit must be in MODE 3 in 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in MODE 4 in 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . This changes the CTS by expanding the Applicability of the Pressurizer to include MODE 3 and requiring the unit to exit this new Applicability within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . The deletion of the Action to open the control rod drive trip breakers is discussed in DOC A03.

The purpose of the ITS MODE 3 Applicability is to prevent solid water RCS operation during heatup and cooldown to avoid rapid pressure rises caused by normal operational perturbation. This change is acceptable because it provides appropriate requirements in MODE 3 to achieve this purpose. This change is designated as more restrictive because it requires the pressurizer to be OPERABLE under more conditions (MODE 3) than is currently required.

RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES None LESS RESTRICTIVE CHANGES L01 (Category I - Relaxation of LCO Requirements) CTS 3.4.4.b states that the pressurizer shall be OPERABLE with a water level between 45 and 228 inches.

ITS LCO 3.4.9.a states that the pressurizer shall be OPERABLE with a pressurizer water level < 228 inches. This changes the CTS by eliminating the lower water level limit of 45 inches.

Davis-Besse Page 2 of 3 Attachment 1, Volume 9, Rev. 0, Page 185 of 415

Attachment 1, Volume 9, Rev. 0, Page 186 of 415 DISCUSSION OF CHANGES ITS 3.4.9, PRESSURIZER The purpose of the CTS 3.4.4.b lower limit is to preserve the steam space during normal operation, allowing both sprays and heaters to maintain the design operating pressure. The lower level limit prevents the low level interlock from de-energizing the pressurizer heaters during steady state operations. This change is acceptable because the low water level limit is not necessary for accident mitigation. The pressurizer water level is routinely monitored by operations personnel to ensure a low level in the pressurizer does not occur, similar to other plant parameters not specified in the Technical Specifications.

Therefore, the low level limit is not necessary to be included in the Technical Specifications. This change is designated as less restrictive because less stringent LCO requirements are being applied in the ITS than are being applied in the CTS.

Davis-Besse Page 3 of 3 Attachment 1, Volume 9, Rev. 0, Page 186 of 415

Attachment 1, Volume 9, Rev. 0, Page 187 of 415 Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs)

Attachment 1, Volume 9, Rev. 0, Page 187 of 415

Attachment 1, Volume 9, Rev. 0, Page 188 of 415 CTS Pressurizer 3.4.9 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4-9 Pressurizer LCO 3.4.4 LCO 3.4.9 The pressurizer shall be OPERABLE with:

LCO 3.4.4.b a. Pressurizer water level < 0--ince nd---- 00 DOC M01 b. A minimum of [

150 esna l of ressurizer ea ers OPERABLE [and 0

Icapable ot being powered from an emergen y power supplyý.

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

OPERABILITY req irements on pressurizer heaters *io not apply in MODE 4. 0 APPLICABILITY: MODES 1, 2, and 3M IMODE_4Awth RCS tempepatife _ [275'1-,

0 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME Action A. Pressurizer water level A.1 Restore level to within limit. 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months not within limit.

Action B. Required Action and B.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time of Condition A not AND met.

B.2 Be in MODE 4 wittCS jte rn),rature ýf-[*05] =F]

[2-1 hours 0 DOC M01 C. Capacity o pressurizer 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months heaters[capa ple of bein powerep by I eme gen pwer I C. rizer heater 0 sup ly] less than I.imit.

BWOG STS 3.4.9-1 Rev. 3.0, 03/31104 Attachment 1, Volume 9, Rev. 0, Page 188 of 415

Attachment 1, Volume 9, Rev. 0, Page 189 of 415 CTS

'Pressurizer 3.4.9 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME DOC M01 D. Required Action and D.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time of Condition C not AND met.

D.2 Be in MODE 4. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY 4.4.4 SR 3.4.9.1 Verify pressurizer water level !9F inches 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months 0 DOC M01 SR 3.4.9.2 ]Verify >_[126] kWof pre= urizer heaters are capa e of being powe 1rd from an emergency power supply.

months24 00 SR 3.4.9.3 [VVfy emergency power supply for pressurize heaters is OPERABLE.

[18] months]

0D

- capacity of essential pressurizer heaters is > 150 kW. I BWNOG STS 3.4.9-2 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 0, Page 189 of 415

Attachment 1, Volume 9, Rev. 0, Page 190 of 415 JUSTIFICATION FOR DEVIATIONS ITS 3.4.9, PRESSURIZER

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

2. The bracketed requirement that the pressurizer heaters be capable of being powered from an emergency power supply has been deleted. The essential heaters, which are the heaters used to meet the LCO requirement, are always powered from the emergency power supply (i.e., they are powered from the essential buses). This is consistent with the ISTS SR 3.4.9.3 Bases, which states that the SR is not applicable ifthe heaters are permanently powered by 1 E power supplies.

3. ISTS 3.4.9 includes the Applicability of MODE 4 with RCS temperature > 275 0 F. The ISTS Bases states that the reason for the MODES 3 and 4 Applicability is to prevent solid water RCS operation during heatup and cooldown to avoid rapid pressure rises caused by normal operational perturbations. However, the temperature cross-over point between MODES 3 and 4 for Davis-Besse is 2801F. In the ISTS, the temperature cross-point is 3300 F. Thus, the Davis-Besse MODE 3 Applicability requirement is essentially equivalent to the ISTS 3.4.9 Applicability of MODE 4 with RCS temperature > 275 0 F (only a 50 F difference exists). Therefore, ITS 3.4.9 does not include the MODE 4 Applicability; only the MODES 1, 2, and 3 Applicability is maintained. Due to this change, the NOTE to the LCO has been deleted and the associated Required Action (Required Action B.2) and Completion Time has been modified to be consistent with the normal time provided in the ISTS to be in MODE 4.

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

Davis-Besse Page 1 of 1 Attachment 1, Volume 9, Rev. 0, Page 190 of 415

Attachment 1, Volume 9, Rev. 0, Page 191 of 415 Improved Standard Technical Specifications (ISTS) Bases Markup and Justification for Deviations (JFDs)

Attachment 1, Volume 9, Rev. 0, Page 191 of 415

Attachment 1, Volume 9, Rev. 0, Page 192 of 415 Pressurizer B 3.4:9 B 3.4 REACTOR COOLANT SYSTEM (RCS)

B 3.4.9 Pressurizer BASES BACKGROUND The pressurizer provides a point in the RCS where liquid and vapor are maintained in equilibrium under saturated conditions for pressure control purposes to prevent bulk boiling in the remainder of the RCS. Key functions include maintaining required primary system pressure during steady state operation and limiting the pressure changes caused by reactor coolant thermal expansion and contraction during normal load transients.

The pressure control components addressed by this LCO include the pressurizer water level he required heatersý and their trols and e mergency per suppliesl Pressurizer safety valves andApressurizer 2he

ý operated relief valveM (PORV\/ are addressed by LCO 3.4.10, "Pressurizer Safety Valves," and LCO 3.4.11, "Pressurizer * (o Operated Relief Valve (PORV)," respectively.

The maximum water level limit has been established to ensure that a liquid to vapor interface exists to permit RCS pressure control during normal operation and proper pressure response for anticipated design basis transients. The water level limit thus serves two purposes:

a. Pressure control during normal operation maintains subcooled reactor coolant in the loops and thus is in the preferred state for heat transport nd

b. By restricting the level to a maximum, expected transient reactor coolant volume increases (pressurizer insurge) will not cause excessive level changes that could result in degraded ability for pressure control.

The maximum water level limit permits pressure control equipment to function as designed. The limit preserves the steam space during normal operation, thus both sprays and heaters can operate to maintain the design operating pressure. The level limit also prevents filling the pressurizer (water solid) for anticipated design basis transients, thus ensuring that pressure relief devices (PORVs or code safety valves) can control pressure by steam relief rather than water relief. Ifthe level limits were exceeded prior to a transient that creates a large pressurizer insurge volume leading to water relief, the maximum RCS pressure might exceed the design Safety Limit (SL) of 2750 psig or damage may occur to the PORVs or pressurizer code safety valves.

B'AOGSTS B 3.4.9-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 0, Page 192 of 415

Attachment 1, Volume 9, Rev. 0, Page 193 of 415 Pressurizer

"'There are two essentialhetr bjanks.* B 3.4.9 Iwith each bank powered from a Sseparate essential bus and each BASES bank having a capacity of 126 kW.

BACKGROUND (continued)

The pressurizer heaters are .used to maintain a pressure in the RCS so reactor coolant in the loops is subcooled and thus in the preferred state for .heat transport to the steam generators (SGs). This function must be maintained with a loss of offsite power. Consequently, the emphasis of this LCO is to ensure that the essential power supplies and the associated heaters are adequate to maintain pressure for RCS loop subcooling with an extended loss of offsite power.

A minimum required available capacity of IT kw W ensures that the RCS pressure can be maintained. Unless adequate heater capacity is available, reactor coolant subcooling cannot be maintained indefinitely.

Inability to control the system pressure and maintain subcooling under conditions of natural circulation flow in the primary system could lead to loss of single phase natural circulation and decreased capability to remove core decay heat.

APPLICABLE In MODES 1 and 2, the LCO requirement for a steam bubble is reflected SAFETY implicitly in the accident analyses. INo safety analys re performedi ANALYSES Ilower IAODES. All analyses performed from a critical reactor condition assume the existence of a steam bubble and saturated conditions in the 0D pressurizer. In making this assumption, the analyses neglect the small fraction of noncondensible gases normally present.

Safety analyses presented in thL-FSAR do not take credit for pressurizer heater operation; however, an implicit initial condition assumption of the 0

safety analyses is that the RCS is operating at normal pressure.

The maximum level limit is of prime interest for the loss of main feedwater (LOMFW) event. Conservative safety analyses assumptions for this event indicate that it produces the largest increase of pressurizer level caused by a moderate frequency event. Thus this event has been selected to establish the pressurizer water level limit. Assuming proper response action by emergency systems, the level limit prevents water relief through the pressurizer safety valves. Since prevention of water relief is a goal for abnormal transient operation, rather than an SL, the value for pressurizer level is nominal and is not adjusted for instrument error.

BWOG STS B 3.4.9-2 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 0, Page 193 of 415

Attachment 1, Volume 9, Rev. 0, Page 194 of 415 Pressurizer B 3.4.9 BASES APPLICABLE SAFETY ANALYSES (continued)

Evaluations performed for the design basis large break loss of coolant lower accident (LOCA), which assumed air ~aximum level than assumed for the LOMFW event have been made. The higher pressurizer level LOWLassumed for thL is the basis for the volume of reactor coolant 0 released to e containment. The containment analysis performed using the mass and energy release demonstrated that the maximum resulting containment pressure was within design limits.

The requirement for emergency power supplies is based on NUREG-0737 (Ref. 1). The intent is to allow maintaining the reactor coolant in a subcooled condition with natural circulation at hot, high pressure conditions for an undefined, but extended, time period after a loss of offsite power. While loss of offsite power is an initial condition or coincident event assumed in many accident analyses, maintaining hot, high pressure conditions over an extended time period is not evaluated as part _ý of SAR accident analyses.

U The maximum pressurizer water level limit satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii). Although the heaters are not specifically used in accident analysis, the need to maintain subcooling in the long term during loss of offsite power, as indicated in NUREG-0737 (Ref. 1), is the reason for providing an LCO.

LCO The LCO requirement for the pressurizer to be OPERABLE with a water level [eve 0] inches ensures that a steam bubble exists. Limiting the maximum operating water level preserves the steam space for pressure 0

control. The LCO has been established to ensure the capability to establish and maintain pressure control for steady state operation and to minimize the consequences of potential overpressure transients.

Requiring the presence of a steam bubble is also consistent with analytical assumptions.

".Sinceeach essential* The LCO requires OPERABLE a minimum Iland:/caPable of powerecl of being pressurizer kof trom heaters powerI(

an/emergency 126 k both essential supp. suc, the LCO addresses both the heatprs and the power banks are used to meet Isupplies. The minimum heater capacity required is sufficient to maintain the LCO requirement. ýthesystem near normal operating pressure when accounting for heat losses through the pressurizer insulation. By maintaining the pressure near the operating conditions, a wide margin to subcooling can be per bank obtained in the loops. The exact design value of~l12f]k is derived from 00 the use of nine heaters rated at 14 kW each. The amount needed to maintain pressure is dependent on the insulation losses, which can vary due to tightness of fit and condition.

BWOG STS B 3.4.9-3 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 0, Page 194 of 415

Attachment 1, Volume 9, Rev. 0, Page 195 of 415 Pressurizer B 34.9 BASES APPLICABILITY The need for pressure control is most pertinent when core heat can cause the greatest effect on RCS temperature, resulting in the greatest effect on pressurizer level and RCS pressure control. Thus Applicability has been designated for MODES 1 and 2. The Applicability is also provided for MOEn[i, ýfor glessurizer water level, for MOQE 4 with RCS ternerature ->[27] 0F. The purpose is to prevent solid water RCS 0 operation during heatup and cooldown to avoid rapid pressure rises caused by normal operational perturbations, such as reactor coolant a

pump startup. IThe12 eentantesignated as the ature of [275]t F hasn btm cutoff for applicabilit bcuse LCO 3.41.112, 'Lo Teprature Overpressure Protekstion (LTOP) System," provides /arequirement for I 0 pressurizer level be/ow [275]*F. The LCO does not apply to MODE 5 withI loops filled becaus* LCO 3.4.12 applies. The LCO does not apply to MODES 5 and 6 vyith partial loop operation.j In MODES 1, 2, and 3, there is the need to maintain the availability of pressurizer heaters capable of being powered from an emergency power supply. In the event of a loss of offsite power, the initial conditions of these MODES give the greatest demand for maintaining the RCS in a hot pressurized condition with loop subcooling for an extended period. R Applicability is modified by a Note stating that the OPERABILITY I requirements on pressurizer heaters do not apply in MODE 4.FFor 0

MODE 4, 5, or 6, it is not necessary to control pressure (by heaters) to ensure loop subcooling for heat transfer when the Decay Heat Removal System is in service, and therefore the LCO is not applicable.

ACTIONS A.1 With pressurizer water level in excess of the maximum limit, action must be taken to restore pressurizer operation to within the bounds assumed in the analysis. This is done by restoring the pressurizer water level to within the limit. The 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months Completion Time is considered to be a reasonable time for draining excess liquid-B.1 and B.2 If the water level cannot be restored, reducing core power constrains heat input effects that drive pressurizer insurge that could result from an anticipated transient.I By shutting don the reactor and r ucing reactor coolant temp ture to at least M/E 3, the potential ermal energy of the reactor coolant mass for LOOA mass and energ eleases is reduced.

INSERT 1 0

BWVOG STS B 3.4.9-4 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 0, Page 195 of 415

Attachment 1, Volume 9, Rev. 0, Page 196 of 415 B 3.4.9 yj) INSERT 1 Therefore, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and to MODE 4 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . The Completion Time of 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months is reasonable, based on operating experience, to reach MODE 3 from full power conditions in an orderly manner and without challenging plant systems. Similarly, the Completion Time of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months to reach MODE 4 is reasonable based on operating experience to achieve power reduction from full power conditions in an orderly manner and without challenging plant systems.

Insert Page B 3.4.9-4 Attachment 1, Volume 9, Rev. 0, Page 196 of 415

Attachment 1, Volume 9, Rev. 0, Page 197 of 415 Pressurizer B 3.4.9 BASES ACTIONS (continued)

Six hours is a reasonab e time based upon operating exerience to reach MODE 3 from full pow r without challenging plant syst91ms and operators.

Further pressure and emperature reduction to MODE4 with RCS temperature S [275]0 places the plant into a MODE yhere the LCO is not applicable. The [2 hour Completion Time to reach he nonapplicable 0 MODE is reasona le based upon operating experie ce.

C.1 f the [emergen ]ipower supplrs to the heaters ar not capable of providing [16JkW, or thefpressurizer heaters are inoperable, restoration j is required in 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . The Completion Time of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months is reasonable considering the anticipation that a demand caused by loss of offsite power will not occur in this period. Pressure control may be maintained during this time using normal station powered heaters.

D.1 and D.2

_ -essentialI If ressurizer heater capability cannot be restored within the allowed Completion Time of Required Action C.1, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and to MODE 4 within tL.(

j- 6flling 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . The Completion Time of 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months is reasonable, based on operating experience, to reach MODE 3 from full power conditions in an orderly manner and without challenging plant systems. Similarly, the Completion Time of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months to reach MODE 4 is reasonable based on operating experience to achieve power reduction from full power conditions in an orderly manner and without challenging plant systems.

SURVEILLANCE SR 3.4.9.1 REQUIREMENTS This SR requires that during steady state operation, pressurizer water level is maintained below the nominal upper limit to provide a minimum space for a steam bubble. The Surveillance is performed by observing the indicated level. The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months interval has been shown by operating practice to be sufficient to regularly assess the level for any deviation and verify that operation is within safety analyses assumptions. Alarms are also available for early detection of abnormal level indications.

BWOG STS B 3.4.9-5 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 0, Page 197 of 415

Attachment 1, Volume 9, Rev. 0, Page 198 of 415

.Pressurizer

B 3.4.9 BASES SURVEILLANCE REQUIREMENTS (continued)

MSR 3.4.9.2 0 esetial The SR requires the power supplies are capable of producing the minimum power and thelasso iatedl pressurizer heaters are verified -to be 0 at their design rating. (This may be done by testing the power supply output and by performing an electrical check on heater element continuity and resistance.) The Frequency oft[ 8 months is considered adequate 0

to detect heater degradation and has been shown by operating experience to be acceptable. R]

0

[SR 3.4.9.3 This SR is not a plicable if the heaters are perm nently powered by 1E power suppli s.

This Surveillanc demonstrates that the heater can be manually 0 transferred to, a d energized by, emergency p oer supplies. The Frequency of [1 ] months is based on a typical fuel cycle and is consistent with imilar verifications of emergen y power. I REFERENCES 1. NUREG-0737, November 1980.

BVVOG STS B 3.4.9-6 Rev. 3.0, 03/31104 Attachment 1, Volume 9, Rev. 0, Page 198 of 415

Attachment 1, Volume 9, Rev. 0, Page 199 of 415 JUSTIFICATION FOR DEVIATIONS ITS 3.4.9 BASES, PRESSURIZER

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

2. Changes are made (additions, deletions, and/or changes) to the ISTS Bases which reflect the plant specific nomenclature, number, reference, system description, analysis, or licensing basis description.

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

4. Changes made to be consistent with the Specification.

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

Davis-Besse Page 1 of 1 Attachment 1, Volume 9, Rev. 0, Page 199 of 415

Attachment 1, Volume 9, Rev. 0, Page 200 of 415 Specific No Significant Hazards Considerations (NSHCs)

Attachment 1, Volume 9, Rev. 0, Page 200 of 415

Attachment 1, Volume 9, Rev. 0, Page 201 of 415 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.4.9, PRESSURIZER There are no specific NSHC discussions for this Specification.

Davis-Besse Page 1 of 1 Attachment 1, Volume 9, Rev. 0, Page 201 of 415

, Volume 9, Rev. 0, Page 202 of 415 ATTACHMENT 10 ITS 3.4.10, PRESSURIZER SAFETY VALVES , Volume 9, Rev. 0, Page 202 of 415

, Volume 9, Rev. 0, Page 203 of 415 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) , Volume 9, Rev. 0, Page 203 of 415

Attachment 1, Volume 9, Rev. 0, Page 204 of 415 ITS 3.4.10 ITS REACTOR COOLANT SYSTEM SAFETY VALVES AND PILOT OPERATED RELIEF VALVE - OPERATING LIMITING CONDITION FOR OPERATION LCO 3.4.10 3.4.3 All pressurizer code safet valves shall be OPERABLE with a lift setting of < 2525 psig.* When not isolated, the pressurizer pilot operated See ITS relief valve shall have a trip setpoint of 1 2435 psig and an allowable value F 3.411 lof >2435 psig.**-F APPLICABILITY: MODES 1, 2 and 3.

ACTION:

ACTION A With one pressurizer code safety valve inoperable, either restore the iinoperable valve to OPERABLE status within 15 minutes obe in HOT SHUTDOWN ACTION B --- _tin ours.

[Add proposed Required Ation B.1 _ ME Add proposed ACTION B for two pressurizer safety valves inoperable. M02 SURVEILLANCE REQUIREMENTS I A dd proposed r es tl mi SR 3.4.10.1 4.4.3 For the pressurizer code safety valves, there are no additional Surveillance Requirements other than those required by Specification 4.0.5. See For the pressurizer pilot operated relief valve a CHANNEL CALIBRATION check 3. ITS]

shall be performed each REFUELING INTERVAL. I The lift/setting pressure shall frrespond to ambient copditions of theL valve zj nominal operatinq temperature--and pressure.

- o ,

I* Allowable value for CHANNEL CALIBRATION check.j I ITS 1 3.4 11 See DAVIS-BESSE, UNIT I 3/4 4-4 Amendment No. 33,60,128,135, 218 Page 1 of 1 Attachment 1, Volume 9, Rev. 0, Page 204 of 415

Attachment 1, Volume 9, Rev. 0, Page 205 of 415 DISCUSSION OF CHANGES ITS 3.4.10, PRESSURIZER SAFETY VALVES 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.

MORE RESTRICTIVE CHANGES M01 CTS 3.4.3 Action requires, in part, that with one pressurizer code safety valve inoperable, to either restore it within 15 minutes or be in HOT SHUTDOWN (MODE 4) within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . ITS 3.4.10 ACTION A requires that with one pressurizer safety valve inoperable, to restore the valve to OPERABLE status within 15 minutes. If not restored, ITS 3.4.10 ACTION B requires the unit to be in MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and MODE 4 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . This changes the CTS by requiring entry into MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months when a shutdown is required.

This change is acceptable because the requirement to place the unit in MODE 3 ensures an intermediate shutdown condition is reached in a shorter period of time. The 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months Completion Time is based on operating experience and the need to reach the required condition from full power in an orderly manner and without challenging unit systems. This change is designated as more restrictive because it imposes a time requirement on when the unit must be in MODE 3.

M02 CTS 3.4.3 Action does not provide any actions for when two pressurizer safety valves are inoperable. Therefore, CTS 3.0.3 would be entered requiring entry into HOT STANDBY (MODE 3) within 7 hours8.101852e-5 days 0.00194 hours 1.157407e-5 weeks 2.6635e-6 months and HOT SHUTDOWN (MODE 4) within 13 hours1.50463e-4 days 0.00361 hours 2.149471e-5 weeks 4.9465e-6 months . ITS 3.4.10 ACTION B, which applies when two pressurizer safety valves are inoperable, requires a shutdown to MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and to MODE 4 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . This changes the CTS by providing one less hour to shut down the unit to both MODE 3 and MODE 4 following discovery of two inoperable pressurizer safety valves.

The purpose of requiring a shutdown when both pressurizer safety valves are inoperable is due to the plant is not meeting the overpressure protection analysis assumptions. This change is acceptable because it provides an adequate period of time to be in a MODE in which the requirement does not apply, commensurate with the severity of the inoperability. The Completion Times of 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months are reasonable, based on operating experience, for reaching MODES 3 and 4, respectively, from full power in an orderly manner and without challenging unit systems. This change has been designated as more restrictive because it reduces the Completion Times to be in MODES 3 and 4.

M03 CTS 4.4.3 requires a verification that the pressurizer safety valve lift setting is within the limit of CTS 3.4.3 (i.e., < 2525 psig). ITS SR 3.4.10.1 includes a similar requirement, but also requires that following testing, the lift setting must Davis-Besse Page 1 of 2 Attachment 1, Volume 9, Rev. 0, Page 205 of 415

Attachment 1, Volume 9, Rev. 0, Page 206 of 415 DISCUSSION OF CHANGES ITS 3.4.10, PRESSURIZER SAFETY VALVES be within + 1% of the nominal setting (2500 psig). This changes the CTS by requiring a minimum pressurizer safety valve setpoint after testing of > 2475 psig.

The purpose of CTS 4.4.3 is to ensure the pressurizer safety valves are set within the accident analysis setpoint. This change is acceptable because the valves must be set in accordance with the Inservice Test Program requirements.

The pressurizer safety valves are ASME Code Section III relief valves, thus they must be set to + 1% of the nominal setpoint following testing. This change is designated as more restrictive since a new requirement is specified in the ITS that is not included in the CTS.

RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES LA01 (Type 3 - Removing ProceduralDetails for Meeting TS Requirements or Reporting Requirements) CTS LCO 3.4.3 is modified by a note (footnote *) that states that the pressurizer safety valves lift setting pressure shall correspond to ambient conditions of the valve at nominal operating temperature and pressure.

This information is not provided in ITS 3.4.10. This changes the CTS by moving this information 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. ITS 3.4.10 still retains a requirement for the valves to be OPERABLE. Under the definition of OPERABILITY, the pressurizer safety valves must be capable of lifting at the assumed conditions, which includes the ambient operating conditions of the pressurizer safety valves themselves. Also, this change is acceptable because these types of procedural details will be adequately controlled in the ITS Bases. Changes to the Bases are controlled by the Technical Specification Bases Control Program in Chapter 5.

This program provides for the evaluation of changes to ensure the Bases are properly controlled. This change is designated as a less restrictive removal of detail change because procedural details for meeting Technical Specification requirements are being moved from the Technical Specifications to the ITS Bases.

LESS RESTRICTIVE CHANGES None Davis-Besse Page 2 of 2 Attachment 1, Volume 9, Rev. 0, Page 206 of 415

Attachment 1, Volume 9, Rev. 0, Page 207 of 415 Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs)

Attachment 1, Volume 9, Rev. 0, Page 207 of 415

Attachment 1, Volume 9, Rev. 0, Page 208 of 415 Pressurizer Safety Valves CTS 3.4.10 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.10 Pressurizer Safety Valves 3.4.3 LCO 3.4.10 Two pressurizer safety valves shall be OPERABLE with lift settings

[24755 sig andj_< 1252EV psig.

0 APPLICABILITY: MODES 1, 2, and 3r FMODE 4 with 011 RCS cold leg temperatures > [283]°q 0

.... ...- 1--. . . . ..... NO TE-The lift settings are rot required to be within the LCO Ii nits for entry into MODES 3 and 4 for the purpose of setting the pressuri.,er safety valves 0 under ambient (hot) conditions. This exception is allov ed for [36] hours following entry into I ODE 3 provided a preliminary col setting was made prior to heatup..

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME Action A. One pressurizer safety A.1 Restore valve to 15 minutes valve inoperable. OPERABLE status.

Action B. Required Action and B.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time not met. AND 0 OR ° B.2 Be in MODE 4 wi ny [ 0 RCS old leg temp rature I DOC M02 Two pressurizer safety !5 [28 ]-F1 valves inoperable.

BWOG STS 3.4.10-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 0, Page 208 of 415

Attachment 1, Volume 9, Rev. 0, Page 209 of 415 CTS Pressurizer Safety Valves 3.4.10 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY 4.4.3 SR 3.4.10.1 Verify each pressurizer safety valve is OPERABLE In accordance in accordance with the Inservice Testing Program. with the Inservice Following testing, lift settings shall be within +/- 1%. Testing Program BWOG STS 3.4.10-2 Rev. 3.0, 03131/04 Attachment 1, Volume 9, Rev. 0, Page 209 of 415

Attachment 1, Volume 9, Rev. 0, Page 210 of 415 JUSTIFICATION FOR DEVIATIONS ITS 3.4.10, PRESSURIZER SAFETY VALVES

1. ISTS LCO 3.4.10 requires both a minimum and maximum lift setting value for the pressurizer safety valves. Davis-Besse is only including the maximum lift setting in ITS LCO 3.4.10, consistent with current licensing basis. The overpressure protection analysis assumes a maximum lift setting for the pressurizer safety valves; a minimum lift setting is not assumed. However, the minimum lift setting is being included in the Davis-Besse ITS as part of ITS SR 3.4.10.1, the pressurizer safety valve lift setting Surveillance. ITS SR 3.4.10.1 requires the as-left lift setting to be

+ 1%, which is consistent with the ASME Code requirements. Thus, the pressurizer safety valves will be considered OPERABLE provided their lift settings are

< 2525 psig, but when tested the as-left lift settings will be > 2475 psig and

_<2525 psig.

2. ISTS 3.4.10 Applicability of MODE 4 with all RCS cold leg temperatures > 283 0 F is not included in the Davis-Besse ITS. This is consistent with the current licensing basis. The temperature cross-over point between MODES 3 and 4 for Davis-Besse is 2800 F. In the ISTS, the temperature cross-point is 3300 F. Thus, the Davis-Besse MODE 3 Applicability requirement is actually more restrictive than the ISTS 3.4.10 Applicability of MODE 4 with RCS temperature > 2830 F. Therefore, ITS 3.4.10 does not include the MODE 4 Applicability; only the MODES 1, 2, and 3 Applicability is maintained. In addition, due to this change, ISTS 3.4.10 Required Action and associated Completion Time have been changed to only require being in MODE 4 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Completion Time is consistent with the time to be in MODE 4 in other actions (e.g., ITS LCO 3.0.3).

3. As described in the Applicability Section of the ISTS Bases, this Note is included to allow testing of the pressurizer safety valves at high pressure and temperature near their normal operating range. The Davis-Besse pressurizer safety valves discharge directly to the containment atmosphere. In-situ testing is not performed at Davis-Besse; the pressurizer safety valves lift settings are verified at a vendor test facility. Thus, the Note allowance is not needed and has been deleted.

4. This change has been made consistent with the Writer's Guide for the Improved Technical Specifications TSTF-GG-05-01, Section 4.1.6.i.5.ii.

Davis-Besse Page 1 of 1 Attachment 1, Volume 9, Rev. 0, Page 210 of 415

Attachment 1, Volume 9, Rev. 0, Page 211 of 415 Improved Standard Technical Specifications (ISTS) Bases Markup and Justification for Deviations (JFDs)

Attachment 1, Volume 9, Rev. 0, Page 211 of 415

Attachment 1, Volume 9, Rev. 0, Page 212 of 415 Pressurizer Safety Valves B 3.4.10 B 3.4 REACTOR COOLANT SYSTEM (RCS)

B 3.4.10 Pressurizer Safety Valves BASES BACKGROUND The purpose of the two spring loaded pressurizer safety valves is to provide RCS overpressure protection. Operating in conjunction with the Reactor Protection System (RPS), two valves are used to ensure that the pSafety Limit (SL) of 2750 psig is not exceeded for analyzed transients O during operation in MODES 1 and 2. Two safety valves are used for SI 4 and MODE 3an ortost MoL.4o For the remainder of MODE ,A----EJ IMIODE-5 , and MODE 6 with the reactor head on, overpressure protection is provided by operating procedures and LCO 3.4.12, "Low Temperature 0

0 Overpressure Protection (LTOP)ye The self actuated pressurizer safety valves are designed in accordance with the requirements set forth in the ASME Boiler and Pressure Vessel Code,Section III (Ref. 1). The required lift pressure ift2500 sietg is%.

amet condis stedmwithmMODe pressurizer to Tpressurizerasaeetyivalvesegeneratesohe rquenchires 0 located ininthe increase cvlstainment.

tepmperature Theorcharge downstream of the flow is valve,'and safety indicateen bybyanan dincrease etabishedn.h tank teoinperature and level. a vito a separate tee opening directly into 1000 The psig.1, 2, setting is lift 3.'This for (D

0 safetyFlow containment.

pressurizer through valves the l generates requirement the ambient conditions associatedabove for lifting pressures with MODES and requires dichrg Thespipe.r sensorsrprovideig~'ebae n h[1%tleac acoustic levels or vibration that is either that the valves be set hot or that a correlation between hot and cold detected by piezoelectric sensors on theI discharge pipe. These sensors provideI settings be established. [(nominal operating temperature and pressure)}

vlve position indication (open/closed) inJ the control room. *'The pressurizer safety valves are part of the primary success path and mitigate the effects of postulated accidents. OPERABILITY of the safety valves ensures that the RCS pressure will be limited to 110% of design 10 (D

pressure. The consequences of exceeding the ASME pressure limit could include damage to RCS components, increased leakage, or a requirement to perform additional stress analyses prior to resumption of reactor operation. U APPLICABLE All accident analyses in the FSAR that require safety valve actuation SAFETY assume operation of both pressurizer safety valves to limit increasing ANALYSES reactor coolant pressure. The overpressure protection analysis (Ref. 1) is also based on operation of both safety valves and assumes that the valves open at the high range of the setting (2500 psig system design pressure plus 1%). These valves must accommodate pressurizer BWOG STS B 3.4.10-1 Rev. 3.1, 12/01/05 Attachment 1, Volume 9, Rev. 0, Page 212 of 415

Attachment 1, Volume 9, Rev. 0, Page 213 of 415 Pressurizer Safety Valves B 3.4.10 BASES APPLICABLE SAFETY ANALYSES (continued) insurges that could occur during a startup, rod withdrawal, ejected rod, loss of main feedwater, or main feedwater line break accident. The f startup accident establishes the minimum safety valve capacity. The startup accident is assumed to occurlat < 150 owe Single failure of a 1E i safety valve is neither assumed in the accident analysis nor required to be addressed by the ASME Code. Compliance with this Specification is required to ensure that the accident analysis and design basis calculations remain valid.

Pressurizer safety valves satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii).

LCO The two pressurizer safety valves are set to open at the RCS design pressure (2500 psig) and within the ASME specified tolerance to avoid exceeding the maximum RCS design pressure SL, to maintain accident 3 analysis assumptions and to comply with 6SMSECode requirementsý, The Jupper ower pressureI to ene limitaeased on the %D1 (D

tolerance requirements for lifting pressures above 1000 psig.

The limit protected by this Specification is the reactor coolant pressure 0

boundary (RCPB) SL of 110% of design pressure. Inoperability of one or both valves could result in exceeding the SL if a transient were to occur.

The consequences of exceeding the ASME pressure limit could include damage to one or more RCS components, increased leakage, or additional stress analysis being required prior to resumption of reactor operation.

APPLICABILITY In MODES 1, 2, and 3, Wand portio sof MODE 4 abov e LTOP cut in ltempature, OPERABILITY of two valves is required because the combined capacity is required to keep reactor coolant pressure below 110% of its design value during certain accidents. MODE 3 nd oconservatively rtion included, although the listed accidents may not require both safety valves for protection.

0 The LCO is not applicable in MODE 4/when any/RCSolle (/

ýtemperatur* is s [283]°F/and MODE 5 because LTOP protection is provided. Overpressure protection is not required in MODE 6 with the reactor vessel head detensioned.

The Note allows e ry into MODES 3 and 4"with the lift seftt gs outside the LCO limits. is permits testing an /examination oft/ safety valves 0

at high pressur and temperature ne/a their normal op ating range, but only after th alves have had a pr iminary cold sett' g. The cold setting gives ass nce that the valves re OPERABLE n r their design BWOG STS B 3.4.10-2 Rev. 3.1, 12/01/05 Attachment 1, Volume 9, Rev. 0, Page 213 of 415

Attachment 1, Volume 9, Rev. 0, Page 214 of 415 Pressurizer Safety Valves B 3.4.10 BASES LCO (continued) condition. Only o e valve at a time will removed from /ervice for testing. The [3 hour exception is b-ed on an 18 hou outage time for each of the o valves. The 18 .ho period is derived/from operating 0

experien hat hot testing can performed in thi. timeframe.

ACTIONS A.1 With one pressurizer safety valve inoperable, restoration must take place within 15 minutes. The Completion Time of 15 minutes reflects the importance of maintaining the RCS overpressure protection system. An inoperable safety valve coincident with an RCS overpressure event could challenge the integrity of the RCPB.

B.1 and 8.2 If the Required Action cannot be met within the required Completion Time or if both pressurizer safety valves are inoperable, the plant must be brought to a MODE in which the requirement does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and to MODE 4 with any RCS cold le erature _ [283]°F within 0

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . The 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months allowed is reasonable, based on operating experience, to reach MODE 3 from full power conditions in an orderjlL--i manner and without challenging plant systems. Similarly, the [ ours 0

allowed is reasonable, based on operating experience, to reach MODE 4 without challenging plant systems. With any RCS col e temperature at or below [2% , ov rpressure protection is provided by TOP. The change from MODE 1, 2, or 3 to MODE 4 reduces the R S energy (core power and pressur ), lowers the potential for large pres urizer insurges, 0

and thereby removps the need for overpressure protection by two pressurizer safety Palves.

SURVEILLANCE SR 3.4.10.1 REQUIREMENTS SRs are specified in the Inservice Testing Program. Pressurizer safety Fo valves are to be tested in accordance with the requirements of the ASME Code (Ref.2), which provides the activities and the Frequency necessary 0 F2 to satisfy the SRs. No additional requirements are specified.

023 inaccordance with Reference 1/

I-The pressurizer safety valve setpoint is M+/-3.% for OPERABILITY;en however, the valves are reset to +/- 1% during the Surveillance to all for REFERENCES , ASME Code for Operation and Maintenance of Nuclear Power 0 Plantj.

0

1. ASME Boiler and Pressure Vessel Code,Section III. I 1 Et wit1 6 d a BWOGSTS B 3.4.10-3 Rev. 3.1, 12/01/05 Attachment 1, Volume 9, Rev. 0, Page 214 of 415

Attachment 1, Volume 9, Rev. 0, Page 215 of 415 JUSTIFICATION FOR DEVIATIONS ITS 3.4.10 BASES, PRESSURIZER SAFETY VALVES

1. Changes are made to be consistent with changes to the Specification.

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

3. Changes are made (additions, deletions, and/or changes) to the ISTS Bases which reflect the plant specific nomenclature, number, reference, system description, analysis, or licensing basis description.

Davis-Besse Page 1 of 1 Attachment 1, Volume 9, Rev. 0, Page 215 of 415

Attachment 1, Volume 9, Rev. 0, Page 216 of 415 Specific No Significant Hazards Considerations (NSHCs)

Attachment 1, Volume 9, Rev. 0, Page 216 of 415

Attachment 1, Volume 9, Rev. 0, Page 217 of 415 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.4.10, PRESSURIZER SAFETY VALVES There are no specific NSHC discussions for this Specification.

Davis-Besse Page 1 of 1 Attachment 1, Volume 9, Rev. 0, Page 217 of 415

Attachment 1, Volume 9, Rev. 0, Page 218 of 415 ATTACHMENT 11 ITS 3.4.11, PRESSURIZER PILOT OPERATED RELIEF VALVE (PORV)

Attachment 1, Volume 9, Rev. 0, Page 218 of 415

415 1, Volume 9, Rev. 0, Page 219 of Attachment (CTS) Markup Current Technical Specification (DOCs) and Discussion of Changes 415 1, Volume 9, Rev. 0, Page 219 of Attachment

Attachment 1, Volume 9, Rev. 0, Page 220 of 415 ITS 3.4.11 ITS REACTOR COOLANT SYSTEM SAFETY VA[VES AND PILOT OPERATED RELIEF VALVE - OPERATING LIMITING CONDITION FOR OPERATION See ITS LCO3.4.11 3.4.3 All pressurizer code safety valves shall be OPERABLE with a lift 3.4.10 Fse-tting of <' 2525 psig.*l When not isolated, the pressurizer pilot operated

.rel ief valve Ishal I 49ve a *trip setpqi'nt of k 2Z5Z psigý/nd/an alTTowaW auI--*(A1 APPLICABILITY: MODES 1, 2 and 3. LA02 ACTION:

With one pressurizer code safety valve inoperable, either restore the See ITS inoperable valve to OPERABLE status within 15 minutes or be in HOT SHUTDOWN- - 3.4.10 J within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

Add proposed ACTIONS A. B. and C M0l SURVEILLANCE RE*UIREMENTS 4.4 F or the pressurizer code safety valves, there are no additional SeeTS Surveillance Requirements other than those required by Specification 4.0.5. 3.4.109 tor the pressurizer pilot operated relief valve a LHANNEL ALIMA110N checkLA0 shall be performed each/REFUELING INTERVAL, /

- - -LAdd1 proposed S R 3.4.11.1 and SIR 3.4.11.2J M02

The lift setting pressure shall correspond to ambient conditions of the valve at nominal operating temperature'and pressure.

F See ITS 3.4.10 Alowble/ value for CHANNEL CALIýRATION check.I LA02 DAVIS-BESSE, UNIT 1 3/4 4-4 Am~endment No. 336,2,3,218 Page 1 of 1 Attachment 1, Volume 9, Rev. 0, Page 220 of 415

Attachment 1, Volume 9, Rev. 0, Page 221 of 415 DISCUSSION OF CHANGES ITS 3.4.11, PRESSURIZER PILOT OPERATED RELIEF VALVE (PORV)

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.

MORE RESTRICTIVE CHANGES M01 CTS 3.4.3 does not provide any actions for when the pressurizer pilot operated relief valve (PORV) or block valve are inoperable and not isolated. Therefore, CTS 3.0.3 would be entered, requiring entry into HOT STANDBY (MODE 3) within 7 hours8.101852e-5 days 0.00194 hours 1.157407e-5 weeks 2.6635e-6 months and HOT SHUTDOWN (MODE 4) within 13 hours1.50463e-4 days 0.00361 hours 2.149471e-5 weeks 4.9465e-6 months . With the PORV inoperable, ITS 3.4.11 ACTION A requires the block valve to be closed within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and power removed from the block valve within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . With the block valve inoperable, ITS 3.4.11 ACTION B requires the block valve to be closed within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and power removed from the block valve within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . If either of these actions are not met, ITS 3.4.11 ACTION C requires a shutdown to MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and to MODE 4 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . This changes the CTS by stating the ACTIONS rather than deferring to CTS 3.0.3 and by adding the requirement to remove power from the block valve.

The purpose of CTS 3.0.3 is to place the unit outside the MODE of Applicability within a reasonable amount of time in a controlled manner. CTS 3.4.3 is silent on these actions, deferring to CTS 3.0.3 for the actions to accomplish this. This portion of the change is acceptable because the ACTIONS specified in ITS 3.4.11 adopt ISTS structure for placing the unit outside the MODE of Applicability without changing the time specified to enter MODE 3 and MODE 4.

Furthermore, power must be removed from the block valve to reduce the potential of inadvertent depressurization that would occur if the PORV failed open. This is acceptable because it ensures an inadvertent depressurization cannot occur due to a failed open PORV. This change is designated as more restrictive because an additional requirement is included in the ITS that is not in the CTS.

M02 CTS 4.4.3 does not specify Surveillance Requirements to cycle the pressurizer pilot operated relief valve (PORV) and the block valve. ITS SR 3.4.11.1 requires performance of one complete cycle of the block valve every 92 days. This Surveillance Requirement is modified by a Note stating that the Surveillance is not required to be performed with the block valve closed in accordance with the Required Action of the LCO. ITS SR 3.4.11.2 requires cycling of the PORV every 24 months. This changes the CTS by adding specific requirements to cycle the block valve and the PORV.

The purpose of ITS SR 3.4.11.1 and SR 3.4.11.2 is to ensure the PORV and associated block valve are operating correctly so the potential for a small break Davis-Besse Page 1 of 3 Attachment 1, Volume 9, Rev. 0, Page 221 of 415

Attachment 1, Volume 9, Rev. 0, Page 222 of 415 DISCUSSION OF CHANGES ITS 3.4.11, PRESSURIZER PILOT OPERATED RELIEF VALVE (PORV)

LOCA through the PORV pathway is minimized, or if a small break LOCA were to occur through a failed open PORV, the block valve could be manually operated to isolate the path. In addition, ITS SR 3.4.11.2 ensures the PORV can be opened as necessary if needed during a steam generator tube rupture (SGTR) event. This change is acceptable because it provides specific requirements for testing of the block valve and the PORV. This change is designated as more restrictive because it adds Surveillance Requirements for the block valve and the PORV to the ITS that are not in the CTS.

RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES LA01 (Type 1 - Removing Details of System Design and System Description, Including Design Limits) CTS 3.4.3 provides the trip setpoint for pilot operated relief valve (PORV). ITS 3.4.11 does not retain this detail. This changes the CTS by moving the details of the trip setpoint to the Bases.

The removal of this detail, which is related to system design, from the Technical Specification is acceptable because this type of information is not necessary to be in the Technical Specifications to provide adequate protection of public health and safety. The PORV is not assumed to open automatically in any safety analysis. It is utilized to depressurize the RCS for mitigation of a SGTR event when offsite power is unavailable. However, UFSAR analysis for the SGTR assumes that offsite power is available. The ITS still retains a requirement for the PORV to be OPERABLE. Also, this change is acceptable because these types of procedural details will be adequately controlled in the ITS Bases.

Changes to the Bases are controlled by the Technical Specification Bases Control Program in Chapter 5. This program provides for the evaluation of changes to ensure the Bases are properly controlled. This change is designated as a less restrictive removal of detail change because procedural details for meeting Technical Specification requirements are being moved from the Technical Specifications to the ITS Bases.

LA02 (Type 4 - Removal of LCO, SR, or other TS requirementto the TRM, UFSAR, ODCM, QAPM, IST Program, or liP) CTS 3.4.3 provides the Allowable Value for PORV opening and footnote **states that this Allowable Value is for the CHANNEL CALIBRATION. CTS 4.4.3 requires a CHANNEL CALIBRATION of the pressurizer pilot operated relief valve (PORV) each REFUELING INTERVAL.

ITS 3.4.11 does not retain these requirements. This changes the CTS by moving the CHANNEL CALIBRATION and associated Allowable Value to the Technical Requirements Manual (TRM).

The removal of these details from the Technical Specifications is acceptable because this type of information is not necessary to be included in the Technical Specifications to provide adequate protection of public health and safety. The PORV is not assumed to open automatically in any safety analysis. It is utilized Davis-Besse Page 2 of 3 Attachment 1, Volume 9, Rev. 0, Page 222 of 415

Attachment 1, Volume 9, Rev. 0, Page 223 of 415 DISCUSSION OF CHANGES ITS 3.4.11, PRESSURIZER PILOT OPERATED RELIEF VALVE (PORV) to depressurize the RCS for mitigation of a SGTR event when offsite power is unavailable. However, UFSAR analysis for the SGTR assumes that offsite power is available. ITS 3.4.11 now requires that the PORV and the block valve be cycled through at least one complete cycle instead of the CHANNEL CALIBRATION. (See DOC M02 for the addition of the ITS SR 3.4.11.1 and ITS SR 3.4.11.2). 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 Surveillance Requirement, including its acceptance criteria, is being removed from the Technical Specifications.

LESS RESTRICTIVE CHANGES None Davis-Besse Page 3 of 3 Attachment 1, Volume 9, Rev. 0, Page 223 of 415

Attachment 1, Volume 9, Rev. 0, Page 224 of 415 Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs)

Attachment 1, Volume 9, Rev. 0, Page 224 of 415

Attachment 1, Volume 9, Rev. 0, Page 225 of 415 CTS Pressurizer PORV 3.4.11 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.11 ,Pressurizer Po-er Operated Relief Valve (PORV) 0 3.4.3 LCO 3.4.11 The PORV and associated block valve shall be OPERABLE.

APPLICABILITY: MODES 1, 2, and 3.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME DOC M01 A. PORV inoperable. A.1 Close block valve. 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months AND A.2 Remove power from block 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months valve.

DOC M01 B. Block valve inoperable. B.1 Close block valve. 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months AND B.2 Remove power from block 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months valve.

DOC Mni C. Required Action and C.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time not met. AND C.2 Be in MODE 4. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months BWOG STS 3.4.11-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 0, Page 225 of 415

Attachment 1, Volume 9, Rev. 0, Page 226 of 415 CTS Pressurizer PORV 3.4.11 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY DOC M02 SR 3.4.11.1 Not required to be performed with block valve closed in accordance with the Required Actions of this LCO.

Perform one complete cycle of the block valve. 92 days DOC M02 SR 3.4.11.2 Perform one complete cycle of the PORV. Imonih~ Q SR 3.4.11.3 [Veri PORV and block valve are capable of bei g 18 months]

0 powe ed from an emergency power source.

BWOG STS 3.4.11-2 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 0, Page 226 of 415

Attachment 1, Volume 9, Rev. 0, Page 227 of 415 JUSTIFICATION FOR DEVIATIONS ITS 3.4.11, PRESSURIZER PILOT OPERATED RELIEF VALVE (PORV)

1. Changes are made (additions, deletions, and/or changes) to the ISTS which reflects the plant specific nomenclature, number, reference, system description, analysis, or licensing basis description.

2. ISTS SR 3.4.11.2 requires performance of one complete cycle of the PORV every 18 months. The ISTS SR 3.4.11.2 Bases states the 18 month frequency is based on a typical refueling cycle. Therefore, the Frequency has been changed to align it with the Davis-Besse refueling cycle, which is 24 months.

3. The bracketed Surveillance Requirement that the PORV and associated block valve are verified to be capable of being powered from an emergency power source has been deleted. The PORV and associated block valve are always powered from the emergency power supply (i.e., they are powered from the essential buses). This is consistent with the ISTS SR 3.4.11.3 Bases, which states that the SR is not applicable ifthe valves are permanently powered by 1 E power supplies.

Davis-Besse Page 1 of 1 Attachment 1, Volume 9, Rev. 0, Page 227 of 415

Attachment 1, Volume 9, Rev. 0, Page 228 of 415 Improved Standard Technical Specifications (ISTS) Bases Markup and Justification for Deviations (JFDs)

Attachment 1, Volume 9, Rev. 0, Page 228 of 415

Attachment 1, Volume 9, Rev. 0, Page 229 of 415 Pressurizer PORV B,3A.11 B 3.4 REACTOR COOLANT SYSTEM (RCS)

B 3:4.11 Pressurizer P7a' rOperated Relief Valve (PORV)

BASES BACKGROUND The pressurizer is equipped with three devices for pressure relief functions: two American Society of Mechanical Engineers (ASME) pressurizer safety valves that are safety grade components and one PORV that is not a safety grade device. The PORV is an electromatic pilot operated valve that is automatically opened at a specific set pressure when the pressurizer pressure increases and is automatically closed on decreasing pressure. The PORV may also be manually operated using controls installed in the control room.

An electric motor operated, normally open, block valve is installed between the pressurizer and the PORV. The function of the block valve is to isolate the PORV. Block valve closure is accomplished manually using controls in the control room and may be used to isolate a leaking PORV to permit continued power operation. Most importantly, the block valve is to be used to isolate a stuck open PORV to isolate the resulting small break loss of coolant accident (LOCA). Closure terminates the RCS depressurization and coolant inventory loss.,the essentialwbuses, thich are pwrdfrom either the ofite The PORV, its block valve, and their controls are powered from no pe) epowue ieý ýut are also caj fig powered fron]Yemergency*

p . Power supplies for the PORV are separate from those for the block valve. Power supply requirements are defined in NUREG-0737, ParagraphX, 1 (Ref. 1). L 2435 psig) (D The PORV setpoin is above the high pressure reactor trip setpoint and below the opening setpoint for the pressurizer safety valve as required by IE Bulletin 79-05B (Ref. 2). The purpose of the relationship of these setpoints is to limit the number of transient pressure increase challenges that might open the PORV, which, if opened, could fail in the open position. A pressure increase transient would cause a reactor trip, reducing core energy, and for many expected transients, prevent the pressure increase from reaching the PORV setpoint. The PORV setpoint thus limits the frequency of challenges from transients and limits the possibility of a small break LOCA from a failed open PORV.

ISRT 1 ý - 0 Placing the setpoint below the pressurizer safety valve opening setpoint reduces the frequency of challenges to the safety valves, which, unlike the PORV, cannot be isolated if they were to fail open. The PORV setpoint is therefore important for limiting the possibility of a small break LOCA.

BWOG STS B 3.4.11-1 Rev. 3.1, 12/01/05 Attachment 1, Volume 9, Rev. 0, Page 229 of 415

Attachment 1, Volume 9, Rev. 0, Page 230 of 415 B 3.4.11 (O INSERT 1 The PORV is also set such that it will open before the pressurizer safety valves are opened. However, it should not open on any anticipated transients. Reference 3 identified that the turbine trip from full power would cause the largest overpressure transient. The Reference 3 analysis demonstrated that with an RPS RC High Pressure trip setpoint of 2355 psig, the resulting overshoot in RCS pressure would be limited to 50 psi. Consequently, the minimum PORV setpoint needs to accommodate both the RCS pressure overshoot and the RPS instrument string error of 30 psi.

Insert Page B 3.4.11-1 Attachment 1, Volume 9, Rev. 0, Page 230 of 415

Attachment 1, Volume 9, Rev. 0, Page 231 of 415 Pressurizer PORV B 3.4.11 BASES BACKGROUND (continued)

The primary purpose of this LCO is to ensure that the PORV and the block valve are operating correctly so the potential for a small break LOCA through the .PORV pathway is minimized, or if a small break LOCA were to occur through a failed open PORV, the block valve could be manually operated to isolate the path.

The PORV may be manually operated to depressurize the RCS as deemed necessary by the operator in response to normal or abnormal transients. The PORV may be used for depressurization when the pressurizer spray is not available; a condition that would be encountered during loss of offsite power. Steam generator tube rupture (SGTR) is one event that may require use of the PORV if the sprays are unavailable.

The PORV may also be used for feed and bleed core cooling in the case of multiple equipment failure events that are not within the design basis, such as a total loss of feedwater.

The PORV functions as an automatic overpressure device and limits challenges to the safety valves. Although the PORV acts as an overpressure device for operational purposes, safety analyses Mdo not take credit for PORV actuation, but]do take credit for the safety valves.

The PORV also pr vides low temperature overpre sure protection (LTOP) during heatup and cooldown. LCO 3.4.12,1"Low Temperature Overpressure Prot ction (LTOP) System," addres es this function.

APPLICABLE The PORV small break LOCA break size is bounded by the spectrum SAFETY of piping breaks analyzed for plant licensing. Because the PORV small ANALYSES break LOCA is located at the top of the pressurizer, the RCS response characteristics are different from RCS loop piping breaks; analyses have been performed to investigate these characteristics.

The possibility of a small break LOCA through the PORV is reduced when the PORV flow path is OPERABLE and the PORV opening setpoint is established to be reasonably remote from expected transient challenges.

The possibility is minimized if the flow path is isolated.

The PORV opening setpoint has been established in accordance with Reference 2. It has been set so expected RCS pressure increases from anticipated transients will not challenge the PORV, minimizing the possibility of a small break LOCA through the PORV.

BWOG STS B 3.4.11-2 Rev. 3.1, 12/01/05 Attachment 1, Volume 9, Rev. 0, Page 231 of 415

Attachment 1, Volume 9, Rev. 0, Page 232 of 415 Pressurizer PORV B 3.4.11 BASES APPLICABLE SAFETY ANALYSES (continued)

Overpressure protection is provided by safety valves, and analyses do not take credit for the PORV opening for accident mitigation.

Operational analyses that support the emergency operating procedures utilize the PORV to depressurize the RCS for mitigation of SGTR when the pressurizer spray system is unavailable (loss of offsite power). RFSAR Q safety analyses for SGTR have been performed assuming that offsite power is available and thus pressurizer sprays (or the PORV) are available.

The PORV and its block valve satisfies Criterion 4 of 10 CFR 50.36(c)(2)(ii).

LCO The LCO requires the PORV and its associated block valve to be OPERABLE. The block valve is required to be OPERABLE so it may be used to isolate the flow path if the PORV is not OPERABLE. If the block valve is not OPERABLE, the PORV may be used for temporary isolation.

APPLICABILITY In MODES 1, 2, and 3, the PORV and its block valve are required to be OPERABLE to limit the potential for a small break LOCA through the flow path. A likely cause for PORV LOCA is a result of pressure increase transients that cause the PORV to open. Imbalances in the energy output of the core and heat removal by the secondary system can cause the RCS pressure to increase to the PORV opening setpoint. Pressure increase transients can occur any time the steam generators are used for heat removal. The most rapid increases will occur at higher operating power and pressure conditions of MODES 1 and 2.

Pressure increases are less prominent in MODE 3 because the core input energy is reduced, but the RCS pressure is high. Therefore, the applicability is pertinent to MODES 1, 2, and 3. The LCO is not applicable in MODE 4 when both pressure and core energy are decreased and the pressure surges become much less sig1nificant.ITh PORV setpoint is reduced for LTOP in MODES 4, 5, arid 6 with the reactor vessel he~bd in place. LCO 3.4.12 addresses tl~e PORV 0 requirements in Vnese MODES.7 ACTIONS A-1 and A.2 With the PORV inoperable, the PORV must be restored or the flow path mutiolated 7; The block

ý valv-sh"ý__tiv" -auld 0

EH)isolatwithin 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . be closed and power must be removed from the block valve to reduce the potential W inadvertent op ning an eprelsurizatior.

0 depressurization that would occur if the PORV failed open BWOG STS B 3.4.11-3 Rev. 3.1, 12/01/05 Attachment 1, Volume 9, Rev. 0, Page 232 of 415

Attachment 1, Volume 9, Rev. 0, Page 233 of 415 Pressurizer PORV B 3.4.11 BASES ACTIONS (continued)

B.1 and B.2 If the block valve is inoperable, it must be restored to OPERABLE status within 1 hour- The prime importance for the capability to close the block valve is to isolate a stuck open PORV. Therefore, if the block valve cannot be restored to OPERABLE status within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , the Required Action is to close the block valve and remove power within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months rendering the PORV isolated. The 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months Completion Times are consistent with an allowance of some time for correcting minor problems, restoring the valve to operation, and establishing correct valve positions and restricting the time without adequate protection against RCS depressurization.

C.1 and C.2 Ifthe Required Action and associated Completion Time cannot be met, the plant must be brought to a MODE in which the requirement does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and to MODE 4 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . The 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months allowed is reasonable, based on operating experience, to reach MODE 3 from full power conditions in an orderly manner and without challenging plant systems. Similarly, the 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months allowed is reasonable, based on operating experience, to reach MODE 4 from full power conditions in an orderly manner and without challenging plant systems.

SURVEILLANCE SR 3.4.11.1 REQUIREMENTS Block valve cycling verifies that it can be closed if needed. The basis for the Frequency of 92 days is the ASME Code (Ref. *. Block valve 4 (D cycling, as stated in the Note, is not required to be performed when it is closed for isolation; cycling could increase the hazard of an existing degraded flow path.

Any combination of indications (e.g., acoustic, system response)

SR 3.4.11.2 may be used to confirm a complete cycle of the PORV.

PORV cycling demonstrates its function.The Frequency of Mmontnths is 0 0 based on a typical refueling cycle and industry accepted practice.

BWOG STS B 3.4.11-4 Rev. 3.1, 12101/05 Attachment 1, Volume 9, Rev. 0, Page 233 of 415

Attachment 1, Volume 9, Rev. 0, Page 234 of 415 Pressurizer PORV B 3.4.11 BASES SURVEILLANCE REQUIREMENTS (continued)

SR 3.4.11.3 This Surveillance is ot required for plants with perma ent 1 E power supplies to the valv s.

This SR demonstrats that emergency power can be rovided and is 0

performed by transf rring power from the normal supl ly to the emergency supply and cycling e valves. The Frequency of 18 r onths is based on a typical refueling cle and industry accepted practic .

REFERENCES 1. NUREG-0737, Paragraph _,.[1, November 1980.

0

2. NRC IE Bulletin 79-05B, April 21, 1979.

EJ-4*. ASME Code for Operation and Maintenance of Nuclear Power 0 Plants.

3. BAW-1890, September 1985. 0 BWVOG STS B 3.4.11-5 Rev. 3.1, 12/01/05 Attachment 1, Volume 9, Rev. 0, Page 234 of 415

Attachment 1, Volume 9, Rev. 0, Page 235 of 415 JUSTIFICATION FOR DEVIATIONS ITS 3.4.11 BASES, PRESSURIZER PILOT OPERATED RELIEF VALVE (PORV)

1. Changes are made to be consistent with changes to the Specification.

2. Changes are made (additions, deletions, and/or changes) to the ISTS Bases which reflect the plant specific nomenclature, number, reference, system description, analysis, or licensing basis'description.

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

4. Changes made to be consistent with the Specification.

5. Editorial change for clarity.

6. Typographical error corrected.

Davis-Besse Page 1 of 1 Attachment 1, Volume 9, Rev. 0, Page 235 of 415

Attachment 1, Volume 9, Rev. 0, Page 236 of 415 Specific No Significant Hazards Considerations (NSHCs)

Attachment 1, Volume 9, Rev. 0, Page 236 of 415

Attachment 1, Volume 9, Rev. 0, Page 237 of 415 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.4.11, PRESSURIZER PILOT OPERATED RELIEF VALVE (PORV)

There are no specific NSHC discussions for this Specification.

Davis-Besse Page 1 of 1 Attachment 1, Volume 9, Rev. 0, Page 237 of 415

Attachment 1, Volume 9, Rev. 0, Page 238 of 415 ATTACHMENT 12 ITS 3.4.12, LOW TEMPERATURE OVERPRESSURE PROTECTION (LTOP)

Attachment 1, Volume 9, Rev. 0, Page 238 of 415

, Volume 9, Rev. 0, Page 239 of 415 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) , Volume 9, Rev. 0, Page 239 of 415

Attachment 1, Volume 9, Rev. 0, Page 240 of 415 ITS 3.4.12 ITS REACTOR COOLANT SYST EM SAFETY VALVES - SHUTDOWN LIMITING CONDITION FOR LCO 3.4.12 3.4.2 Decay Beat Removal System relief valve DH-lift setting of < 330 psig and isolation valves rntn nnior ln .th.4r vnlvw nl~rntnrs removed.

APPLICABILITY: MODES 4 and 5.

ACTION: Add proposed MODE 6 Applicability A. WithFDH g not OPERABLE:

ACTION C 1. Make the valve OPERABLE vithin eight hours; or. LAU1

2. a. Within next one hour, disable the capability of both high pressure injection (HPI) pumps to inject vater into the reactor coolant system; and

b. Within next eight hours:

ACTION D- 4i 1. Disable the automatic transfer of makeup pump suction to the borated vater storage tank on low makeup tank level; and

2. Reduce makeup tank level to < 73 inches and reduce reactor coolant system pressure and pressurizer level vithin the acceptable region on Figures 3.4-2a (in MODE 4) and 3.4-2b (in MODE 5).

ACTION A B. WithJOH-li DS-121closed, openIH-21 ad DH-23 vithin one hour.

I - LA01 ACTION B C. With the control pover not removed from ODH-li myd DH-12Tremove the pover to the valve operators lat the Motor Controý Cters f ithin one hour-SURVEILLANCE REQUIREMENTS 4.4.2 Decay Beat Removal System relief valve shall be determined OPERABLE: LA01 SR 3.4.12.2 a. per the surveillance requirements of Specification 4.0.5.

b. at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by verifying either:

SR 3.4.12.1 1. isolation valvesIOH-li d DH-12 open vith control pover removed from their valve operators; or LA01 Required Action B.2 2. valvesIDH-21 d OH open.

The lift setting pessure shall correspond to blent conditions of the LA0 valve at nominal erating te mrature and prefsure.

DAVIS-BESSE, UNIT I 314 4-3 Amendment No. 57,170,135 Page 1 of 3 Attachment 1, Volume 9, Rev. 0, Page 240 of 415

Attachment 1, Volume 9, Rev. 0, Page 241 of 415 ITS 3.4.12 ITS Figure 3.4.12-1 Fieure 3.4-2a I Reactor Coolant System Pressure - Pressurizer Level Limits for inoperable Decay Heat Removal System Relief Valve in MODE 4 400 350

~UlIACCZPTABLE* REGn[OF 300 MODE 4 ac 250 0.

200 0.

U 150 ACCEPTABLE REGION h~zz 100

~ 4 \

50 I I I I , m

\~V

.NOTE: NOT OORRECTED FOR INOTR1ThENT ERROR -_

r-LGi

!I I I I I I I I 0 40 80 120 160 200 240 Initial Pressurizer Level (Inches) 4 DAVIS-BESSE, UNIT I 3/4 4- a Amendment No. A' 116 Page 2 of 3 Attachment 1, Volume 9, Rev. 0, Page 241 of 415

Attachment 1, Volume 9, Rev. 0, Page 242 of 415 ITS 3.4.12 ITS Figure 3.4.12-2 Fieure 3.4-2b I Reactor Coolant System Pressure - Pressurizer Level Limits for inoperable Decay Heat Removal System Relief Value in MODE 5 Add proposed MODE 6 Me 400 LAO1

NOTE. --NOT C-RRCTED FOR INSM, ERROR 350 w

300 C"

Ce*

250 I,,

0 UNACCEP!?TABLE RE*CIO -

200 I-e I 0 150 L

100 MODE 5 50 ACCEPTABLE REGTON 0 40ouI lo0% ZVU Z-VU Initial Pressurizer Level (Inches) 3/4 4-4b Amendment ?No. A7, 116 DAVIS-BESSE, UNIT I Page 3 of 3 Attachment 1, Volume 9, Rev. 0, Page 242 of 415

Attachment 1, Volume 9, Rev. 0, Page 243 of 415 DISCUSSION OF CHANGES ITS 3.4.12, LOW TEMPERATURE OVERPRESSURE PROTECTION (LTOP)

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.

MORE RESTRICTIVE CHANGES M01 CTS 3.4.2 is applicable in MODES 4 and 5. CTS Figure 3.4-2b is applicable in MODE 5. ITS LCO 3.4.12 is applicable in MODES 4 and 5, and MODE 6 when the reactor vessel head is on. In addition, Figure 3.4.12-2 is applicable in MODE 5 and MODE 6 when the reactor vessel head is on. This change expands the Applicability of the low temperature overpressure protection components to be OPERABLE in MODE 6 when the reactor vessel head is on.

The purpose of CTS 3.4.2 is to ensure that there is a sufficient low temperature protection during shutdown conditions. The definition of MODE 6 in ITS Table 1.1-1 clearly states that MODE 6 is when one or more reactor vessel head closure bolts are less than fully tensioned. Therefore, this change will require the MODE 6 Applicability when one or more reactor vessel head closure bolts are less than fully tensioned, until the vessel head is removed. This change is necessary because an overpressure event could occur in this situation and a relief path is still necessary until the head is physically removed. This change is designated as more restrictive because it adds additional requirements to the CTS.

RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES LA01 (Type I - Removing Details of System Design and System Description, Including Design Limits) CTS 3.4.2 is modified by a note (footnote *) that states that the decay heat removal relief valve lift setting pressure shall correspond to normal operating temperature and pressure. CTS LCO 3.4.2, Actions A, B, and C, and Surveillance Requirement 4.4.2 provides specific valve numbers for certain Decay Heat Removal System valves. CTS 3.4.2 Action c requires power to the valve operators be removed at the motor control centers. CTS Figures 3.4-2a and 3.4-2b (used when a Decay Heat Removal System relief valve is inoperable) include a Note that states the Figures are not corrected for instrument error. ITS 3.4.12 does not include these details. Furthermore, ITS 3.4.12 uses the plant specific names for the associated valves, and requires control power to be Davis-Besse Page 1 of 2 Attachment 1, Volume 9, Rev. 0, Page 243 of 415

Attachment 1, Volume 9, Rev. 0, Page 244 of 415 DISCUSSION OF CHANGES ITS 3.4.12, LOW TEMPERATURE OVERPRESSURE PROTECTION (LTOP) removed from the RCS to DHR system isolation valves. This changes the CTS by moving the valve numbers, the information concerning the lift settings, the details concerning how to remove power from the valves, and that the Figures are not corrected for instrument error 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 Specification to provide adequate protection of public health and safety. ITS 3.4.12 still retains a requirement for the valves to be OPERABLE, uses the plant specific names for the valves, requires control power to be removed from the valves, and the Figures to be used when a Decay Heat Removal System relief valve is inoperable. Under the definition of OPERABILITY, the Decay Heat Removal System relief valve must be capable of lifting at the assumed conditions, which includes ambient operating conditions of the Decay Heat Removal System relief valve itself. Also this change is acceptable because these types of procedural details will be adequately controlled in the ITS Bases. Changes to the Bases are controlled by the Technical Specification Bases Control Program in Chapter 5. This program provides for 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 moved from the .Technical Specifications to the ITS Bases.

LESS RESTRICTIVE CHANGES None Davis-Besse Page 2 of 2 Attachment 1, Volume 9, Rev. 0, Page 244 of 415

Attachment 1, Volume 9, Rev. 0, Page 245 of 415 Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs)

Attachment 1, Volume 9, Rev. 0, Page 245 of 415

Attachment 1, Volume 9, Rev. 0, Page 246 of 415 0

CTS LTOP tern 3.4.12 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.12 Low Temperature Overpressure Protection (LTOP) 3.4.2 LCO 3.4.12 An LTOP Syste shall be OPERABL with a maximum of [one] makeup pump capable f injecting into the RC high pressure inje ion (HPI) deactivated, an the core flood tanks CFTs) isolated and RT 1

- - - -- - --. NO ES - - -. .. .

1. [Two malp eup pumps] may be c pable of injecting f r _ 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months for pump s p operations.

2. CFT ma be unisolated when FT pressure is less han the maximur RCS pressure for th existing RCS temp rature allowed by the p essure and temperat re limit curves provi ed in the PTLR.

a. Pressu izer level - [220] inch s and an OPERABL power operated relief v lve (PORV) With a lift etpoint of _ [555] p ig or

b. The R S depressurized and n RCS vent of Ž [0. 5] square inch.

APPLICABILITY: MoD4 kvhen any/iACS cold legýmperature id:_ [283]'F, I MOQDE 5,1 MODE 6 when the reactor vessel head is on.

ACTIONS CONDITION REQUIRED ACTION [ COMPLETION TIME A. More than [onq] makeup A.1 Initiote action to verify only Immediately pump capable/of [on :] makeup pump is EETE13 injecting into tie RCS. a asble of injecting into th RS.

B. HPI activate 1d. B.1 itiate action to verify H I Immediately Jeactivated./

BWOG STS 3.4.12-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 0, Page 246 of 415

Attachment 1, Volume 9, Rev. 0, Page 247 of 415 3.4.12 CTS O* INSERT 1 3.4.2 The Decay Heat Removal (DHR) System relief valve shall be OPERABLE with:

a. A lift setting of < 330 psig; and

b. The Reactor Coolant System (RCS) to DHR System isolation valves open with control power removed.

O INSERT 2 Action B A. DHR System relief A.1 Open RCS to DHR System 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months valve inoperable due to isolation bypass valves.

one or more RCS to DHR System isolation AND valves closed.

A.2 Verify RCS to DHR System Once per isolation bypass valves open. 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months

. Action C B. DHR System relief B.1 Remove control power from 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months valve inoperable due to RCS to DHR System isolation one or more RCS to valves.

DHR System isolation valves with control power not removed.

Action A. 1 C. DHR System relief C.1 Restore DHR System relief 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months valve inoperable for valve to OPERABLE status.

reasons other than Condition A or B.

Insert Page 3.4.12-1a Attachment 1, Volume 9, Rev. 0, Page 247 of 415

Attachment 1, Volume 9, Rev. 0, Page 248 of 415 3.4.12 CTS 0 INSERT 2 (continued)

Action A.2 D. Required Action and D.1 Disable capability of both high 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months Associated Completion pressure injection pumps to Time not met. inject water into the RCS.

AND D.2 Disable makeup pump suction 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months automatic transfer to the borated water storage tank on low makeup tank level.

AND D.3 Verify makeup tank level 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months

< 73 inches.

AND D.4 Verify RCS pressure and 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months pressurizer level in Acceptable Region of Figure 3.4.12-1 or 3.4.12-2, as applicable.

Insert Page 3.4.12-1b Attachment 1, Volume 9, Rev. 0, Page 248 of 415

Attachment 1, Volume 9, Rev. 0, Page 249 of 415 CTS 0

LTOP 3.4.12 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME C. A CFT not is lated when C.1 I.late affected CFT. 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months CFT pressur is greater than or equa to the maximum R S pressure for existing tt mperature allowed in th PTLR.

D. Required A ion C.1 not D.1 Ii crease RCS tempera ure 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months met within t e required t > 1751F.

Completion ime.

OR D.2 epressurize affected FT 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months

< [555] psig.

E. Pressurizer level E.1 estore pressurizer le el to 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months

> [220] inch es. [220] inches.

F. Required ion E.1 not F.1 lose and maintain cl sed 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months met within he required he makeup control va ye Completior Time. aInd its associated isol tion valve.

AND F.2 Stop RCS heatup. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months G. PORV ino erable. G.1 Restore PORV to 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months OPERABLE status.

H. Required ction G.1 not H.1 Reduce makeup tan level 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months met withi the required to * [70] inches.

Completi n Time.

AND H.2 Deactivate low low keup 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months tank level interlock t the borated water storag tank suction valves.

BVV.OG STS 3.4.12-2 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 0, Page 249 of 415

Attachment 1, Volume 9, Rev. 0, Page 250 of 415 CTS 0

'LTOP 3.4.12 ACTIONS (continued)

CONDITION REQUIRED ACTION ] COMPLETION TIME

1. Pressurizer leve 1.1 Depr ssurize RCS and 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months

> [220] inches. esta lish RCS vent of

_[0. 5] square inch.

AND PORV inopera le.

OR LTOP Syste inoperable fo any reason other than Condition A rough Condition H 4.4.2.b BWOKG STS 3.4.12-3 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 0, Page 250 of 415

Attachment 1, Volume 9, Rev. 0, Page 251 of 415 CTS 0

LTOP 3.4.12 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.4.12.6 erify required RCS vent .75] square inch is 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months for pen. unlocked open vent valve(s)

AND 31 days for other vent path(s)

SR 3.4.1 Peo 9ANNEL FUN,'T'ONAL TE5f for PORVM' Within [12] ho rsn 4.4.2.a Verify DHR System relief valve lift setpoint 5 330 psig in] RCS temr rature accordance with the Inservice Testing (IST) Program. to AN

Dr[283 D F "N

IS T P:ro~g~r~a~mj with the

[In accordance 31 aysthereafte SR 3.4.12.8 /Perform CHANNEL CABRATION for PORV. [18] months 4-ý-RT 3 BWsJOG STS 3.4.12-4 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 0, Page 251 of 415

Attachment 1, Volume 9, Rev. 0, Page 252 of 415 3.4.12 CTS 0 INSERT 3 400 Figure 3.4-2a 350 - - - -UNACCEPTABLE REGION

- - - .hil a - ~ ia -T -

N (J2 300 S

a - - - -

- - a a -

a, S

C-I- 250 a,

I-S 200 C

=

'a a

0

,150 C

0

'a a

S

.ACCEP-TA1LE REGION 100 50

- -- - a - - -

0 40 80 120 160 200 240 Initial Pressurizer Level (Inches)

Figure 3.4.12-1 RCS Pressure Versus Pressurizer Level Limit for Inoperable DHR System Relief Valve in MODE 4 Insert Page 3.4.12-4a Attachment 1, Volume 9, Rev. 0, Page 252 of 415

Attachment 1, Volume 9, Rev. 0, Page 253 of 415 3.4.12 CTS 0 INSERT 4 400 Figure 3.4-2b 350 300 I-O

0 Oa 200 Cr 150 100 50 0

Inicial Pressurizer level (Inches)

Figure 3.4.12-2 RCS Pressure Versus Pressurizer Level Limit for Inoperable DHR System Relief Valve in MODE 5 and MODE 6 when the reactor vessel head is on Insert Page 3.4.12-4b Attachment 1, Volume 9, Rev. 0, Page 253 of 415

Attachment 1, Volume 9, Rev. 0, Page 254 of 415 JUSTIFICATION FOR DEVIATIONS ITS 3.4.12, LOW TEMPERATURE OVERPRESSURE PROTECTION (LTOP)

1. ISTS 3.4.12 has been changed to be consistent with the Davis-Besse current licensing basis and analysis basis. The Davis-Besse low temperature overpressure protection analysis between 280°F (MODE 4 entry temperature) and 140OF only requires the Decay Heat Removal (DHR) System relief valve to be OPERABLE with a setpoint of < 330 psig to protect the RCS from an overpressure condition. This relief valve performs the same function as the PORV in the ISTS. Between 2801F and 140 0 F, the analysis does not require the high pressure injection (HPI) pumps to be incapable of injecting, the core flooding tanks to be isolated, or the pressurizer level to be within a certain limit. The CTS Actions only require the HPI pumps to be disabled and the pressurizer level to be within a certain limit if the DHR System relief valve is inoperable. To ensure the relief valve remains connected to the RCS, the CTS requires the Reactor Coolant System (RCS) to DHR isolation valves to be open with control power removed. If these requirements are not met, the CTS provides specific Actions to take. These requirements have been maintained in the ITS. In addition, the current Surveillances to ensure the LCO is met have also been provided. Finally, Davis-Besse has added the ISTS MODE 6 Applicability to be as consistent with the ISTS as possible, while still maintaining the specific analysis assumption requirements.

Davis-Besse Page 1 of 1 Attachment 1, Volume 9, Rev. 0, Page 254 of 415

Attachment 1, Volume 9, Rev. 0, Page 255 of 415 Improved Standard Technical Specifications (ISTS) Bases Markup and Justification for Deviations (JFDs)

Attachment 1, Volume 9, Rev. 0, Page 255 of 415

Attachment 1, Volume 9, Rev. 0, Page 256 of 415 I All changes are unless otherwise noted P

Low Temperat1re Overpressure P ection LTOIj:

B 3.4.12 B 3.4 REACTOR COOLANT SYSTEM (RCS)

B 3;4.12 Low Temperature Overpressure Protection (LTOP) 0 BASES BACKGROUND --------------- --- REVIE - VE 'S NOTE --------

For plants for ich the NRC has aplroved LTOP setpoi rts based on non-10 CFR 5P, Appendix G, methooology, as allowed ir* NRC Generic 0 Letter 88-11, he following Bases mrst be revised accor ingly.

LTOP tcontrols RCS pressure at low temperatures so the integrity of the reactor coolant pressure boundary (RCPB) is not compromised by violating the pressure and temperature (P/T) requirements of 10 CFR 50, Appendix G (Ref. 1). The reactor vessel is the limiting RCPB component for providing such protection. LCO 3.4.3, "RCS Pressure and Temperature (P/T) Limits," provides the allowable combinations for operational pressure and temperature during cooldown, shutdown, and heatup to keep from violating the Reference 1 limits.

The reactor vessel material is less tough at reduced temperatures than at normal operating temperature. Also, as vessel neutron irradiation accumulates, the material becomes less resistant to pressure stress at low temperatures (Ref. 2). RCS pressure must be maintained low when temperature is low and must be increased only as temperature is increased.

Operational maneuvering during cooldown, heatup, or any anticipated operational occurrence must be controlled to not violate LCO 3.4.3.

Exceeding these limits could lead to brittle fracture of the reactor vessel.

LCO 3.4.3 presents requirements for administrative control of RCS pressure and temperature to prevent exceeding the P/T limits.

This LCO provides RCS overpressure protection in the applicable MODES by ensuring an adequate pressure relief capacity n- a minimum/

theDecy Hat emoalcoolant additi n capability. Te pr ssure rel e capacity equires eith~er tDHR) System relieog valve. the power op rated relief valve ( RV) lift setpoint to b reduced and/

pressurizer oolant level at or bel w a maximum limit r the RCS depressuri d and with an RCS ent of sufficient size to handle the limiting tra sient during LTOP.

The LTOP appr ach to protecting thý vessel by limiting /oolant addition capability allovys a maximum of [on] makeup pump, ar/d requires deactivating ýPI, and isolating the/lore flood tanks (CFs).

BWOG STS B 3.4.12-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 0, Page 256 of 415

Attachment 1, Volume 9, Rev. 0, Page 257 of 415 I

All changes are unless otherwise noted 9 ILow Termper re Overpressure P ection LTOPI e B 3.4.12 0

BASES BACKGROUND (conntinued)

Should more th n [one] HPI pump injeo on an HPI actuati n, the pressurizer leve and PORV or anothe* RCS vent cannot revent overpressurizinO the RCS. Even with nly one HPI pump PERABLE, the vent cannot prevent RCS overpre surization.

The pressurize level limit provides a ompressible vapor pace or cushion (either steam or nitrogen) th t can accommodate a coolant insurge and pr vent a rapid pressure increase, allowing t e operator time to stop the inc ase. The PORV, wit reduced lift setting or the RCS vent is the ov pressure protectionn .d that acts as b ckup to the vice operator in ter inating an increasind pressure event.

With HPI dea ivated, the ability to rovide RCS coolant addition is restricted. To balance the possible eed for coolant addition, the LCO does not req ire the Makeup Syste to be deactivated. Due to the lower pressures as ociated with the LTO MODES and the e pected decay heat levels, t e Makeup System ca provide flow with t e OPERABLE makeup pu through the makeup control valve.

PORV Require nts As designed fo the LTOP System, e ch PORV is signale to open if the RCS pressure pproaches a limit set in the LTOP actuati n circuit. The LTOP actuatio circuit monitors RCS pressure and deter ines when an overpressure ondition is approache . VAen the monito ed pressure meets or exce ds the setting, the P RV is signaled to o en. Maintaining the setpoint hin the limits of the L 0 ensures the Ref rence 1 limits will be met in ny event analyzed fo LTOP.

When a POR is opened in an incr asing pressure tran ient, the release of coolant ca ses the pressure incr ase to slow and re rse. As the PORV relea s coolant, the RCS p essure decreases ntil a reset pressure is r ached and the valve i signaled to close. The pressure continues to ecrease below the r et pressure as the alve closes.

RCS Vent Req irements Once the RCS is depressurized, a v nt exposed to the c ntainment atmosphere 11maintain the RCS a ambient containme t pressure in an RCS overpre sure transient, if the r lieving requiremen of the maximum credible LTO transient do not ex ed the capabilities the vent. Thus, the vent path must be capable of reyving the flow of th limiting LTOP transient an maintaining pressure below P/T limits- T e required vent capacity ma be provided by one r more vent paths.

BWOG STS B 3.4.12-2 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 0, Page 257 of 415

Attachment 1, Volume 9, Rev. 0, Page 258 of 415 B 3.4.12 (O INSERT 1 The DHR System relief valve provides overpressure protection for the RCS during low temperature operations. RCS and DHR Systems are monitored for temperature and pressure. Maintaining the relief setpoint within the limits of the LCO ensures the Reference 1 limits will be met in any event in the LTOP analysis.

If system pressure exceeds the lift setpoint of the DHR System relief valve, it will open. As the relief valve opens, coolant is released and pressure decreases. When the relief valve reset is reached, below the LTOP pressure limit, the relief valve closes.

Insert Page B 3.4.12-2 Attachment 1, Volume 9, Rev. 0, Page 258 of 415

Attachment 1, Volume 9, Rev. 0, Page 259 of 415 I All changes are unless otherwise noted 9 jLow Temper re Overpressure P ection TOPf B 3.4.12 0

BASES BACKGROUND (continued)

For the remaining portions of MODE 3, overpressure protection is provided by operating procedures. 12 For an RCS ver to meet the flow ca acity, it requires rerving a pressurizer safety valve, locking the RV in the open Ibsition and disabling its .blck valve in the open position, or similarly establishing a vent by openipg an RCS vent valve! The vent path(s) riust be above the leve of reactpSr coolant, so as not t* drain the RCS wh n open.

APPLICABLE Safety analyses (Ref. 3) demonstrate that the reactor vessel can be portions of SAFETY adequately protected against overpressurization transients durngn MODE ANALYSES shutdown. In MODES I'M2, andn/and i M 4 w-with K tempera ure iexcec*'ding [283]°F1 the pressurizer safety valves will prevent RCS 4

-pressure from exceeding the Referencse 1 limits. At nominally 2[3 F *.-

[ , overpressure prevention falls to [ OPERABLEPV and a 280 DHR System relief valve. **a rest rted coolant suffcient size RCS levelvent.

in theEach pressuriz of tls r or toen a depressur haaltied, ed RCS and Below 140°F7, credible 1 o~pesr reie capability. I s en asaI ie overp re ssurization sources J ivpesr eircp oy are secured.

The actual temperature at which the pressure in the P/T limit curve falls below the pressurizer safety valve setpoint increases as vessel material toughness decreases due to neutron embrittlement. Each time the P/T limit curves are revised, the LTOP t will be re-evaluated to ensure that its functional requirements can still be met with the PR and \

DHR System relief valve and pressurizer lev metod or the nepresstize and vented RCS co dition.

operating procedures.

nransients that are capable or overpressurizing the IKLCS nave been Pi) identified and evaluated. These transients relate to either mass input or FCore Flooding" High Pressure Injection (HI heat input: actuating th-[g-ffij-System, discharging the PT , ýenergizing Tanks (CFTs) a the pressurizer heaters, failing the makeup control valve open, losing decay heat removal,*starting a reactor coolant pump (RCP) a arge temperatur mismatch between the primary and seconany a olant Isystems, ?hd adding nitrogen to thg pressurizel.

HPI actuation ard CFT discharge are he transients that r ult in exceeding P/T lnhtits within < 10 minut s, in which time no operator action is assumed to t ke place. In the rest, operator action afte that time precludes over ressurization. The ar lyses demonstrate that the time allowed for op rator action is adequa e, or the events are self limiting and do not exceed IT limits.

The following re required during th LTOP MODES to nsure that transients do ot occur, which eithe of the LTOP overpr ssure protection means cann handle:

INSERT 2 BWOG STS B 3.4.12-3 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 0, Page 259 of 415

Attachment 1, Volume 9, Rev. 0, Page 260 of 415 B 3.4.12 O INSERT 2 The DHR System relief valve (DH-4849), which is in the suction line to the decay heat pumps, has been sized to pass 1800 gpm at the nominal set pressure of 320 psig. The flow rate is based on the maximum developed runout flow (900 gpm per pump) with both HPI pumps running simultaneously. This flow rate is considered to cause the worst credible pressure transient. The opening of a CFT isolation valve was not considered because power is removed from the valve once it is closed upon plant cooldown and depressurization. Other postulated occurrences, makeup control valve failing to open, loss of DHR System cooling, all pressurizer heaters energizing, do not produce a pressure excursion as severe as that produced by the two HPI pumps.

Although the pressurizer, by procedure, cannot be solid, for the purpose of analysis it was considered to go solid during the transient. The DHR System relief valve is a Seismic Class I Nuclear Class 2 bellows type of safety-relief valve. It should be noted that the postulation of both HPI pumps starting during DHR System operation is made only for the purpose of sizing the DHR System relief valve. The possibility of this event occurring due to either a single operator error or a single spurious signal is precluded by the design of the Safety Features Actuation System.

Insert Page B 3.4.12-3 Attachment 1, Volume 9, Rev. 0, Page 260 of 415

Attachment 1, Volume 9, Rev. 0, Page 261 of 415 I

All changes are unless otherwise noted 9 ILow Temper re Overpressure P R ction MTOP: tern B 3.4.12 0

BASES APPLICABLE SAFETY ANALYSES (continued)

a. Deactivatin all but [one] makeu pump,

b. Deactivati g HPI, and

c. Immobili ing CFT discharge is71ation valves in the Irclosed positions.

DHR System relief valve The Reference 3 analyses demonstrate the[ can maintain RCS pressure below limits 1when only one makeup ump is actuated.1 Consequent y, t LCO allows only [one] m eup pump to be OPERABLE in re LTOP MODE .

Since the PORV' cannot do this for on HPI pump and th* RCS vent cannot do this fbr even one pump, th# LCO also requires/the HPI actuation circu s deactivated and th7 CFTs isolated./

The isolated 9FTs must have their iischarge valves clo ed and the valve power breake)'s fixed in their open Iositions. The analy es show the effect of CFTIdischa roe i' over a nrrower RCS temoe 'ature ranae The DHR System relief 1

(175-F and lIow) than that of the ;CO([283]-F-and b low).

valve is placed in e' echp'iiics analyses established th4temperature of LTOP service before RCS -------------

temperature is reduced FF racur below 280 ___ IApplicability. d [283]fF. Above this temperature, the pressurizer safety and operating procedures valvesprovide the reactor vessel pressure protection. The vessel pmaterials were assumed to have a neutron irradiation accumulation equal to 21 effective full power years (EFPYs) of operation.

This LCO will dýactivate the HPI act ation when the RC temperature is

_ [283]'F. Th consequences of a Shall break LOCA in/LTOP MODE 4 conform to 10 CFR 50.46 and 10 CrR 50, Appendix K Refs. 4 and 5),

requirements by having a maximur of [one] makeup p mp OPERABLE.

Reference 3 contains the acceptance limits that satisfy the LTOP requirements. Any change to the RCS must be evaluated against these analyses to determine the impact of the change on the LTOP acceptance limits.

PORV Perfor nce The fracture chanics analyses st~w that the vessel i protected when the PORV is t to open at s [555] Isig. The setpoint i derived by modeling the erformance of the L'OP System, assu ing the limiting allowed LTO transient of uncontr lied HPI actuation f one pump.

BVWOG STS B 3.4.12-4 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 0, Page 261 of 415

Attachment 1, Volume 9, Rev. 0, Page 262 of 415 I

All changes are a unless otherwise noted Low Temperre Overpressure ection TOPI: tr B 3.4.12 Q

BASES APPLICABLE SAFETY ANALYSES (continued)

These analysed consider pressure oy'ershoot and under oot beyond te PORV opening andPORV stroke times./The closing, resulting setpoint from signal proces ing and valv a or below the deriv*' limit ensure*

the Referen 1 limits will be met.

PFIT The Iimits PORVconflict set oint withwillthe beLTOP re-evaluat ana 6d yssifor compliance it .The P/TwheI4 limitsthe arerevisedJ periodically modified as the reactor vessel material toughness decreases As required by License due to embrittlement induced by neutron irradiation. Revised P/T limits Condition 2.C(3)(d), prior to operation beyond 21 Effective are determined using neutron fluence projections and the results of Full Power Years, a examinations of the reactor vessel material irradiation surveillance reanalysis and proposed specimens. IThe Base,'or LCO 3.4.3 Oscuss these eaminations.I modifications, as necessary, to ensure continued means of protection for LTOP events will be provided to the NRC. The PORV is co sideredrst case an activesingi corfiponent. Therefore, its failure I represents the LTOP active failure.

Pressurizer Lev I Performance Analyses of op rator response time s ow that the pressu zer level must be maintained [220] inches to provi e the 10 minute acti n time for correcting trans ents.

The pressurize level limit will also be re-evaluated for or pliance each time P/T surveillance.

material limit c rves are revised bas d on the results oft e vessel RCS Ven Permance With the RCS epressurized, analys s show a vent of [0. 5] square inches is capa le of mitigating the tr nsient resulting fro full opening of the makeup ntrol valve while the keup pump is pro ding RCS makeup. The pacity of a vent this size is greater than he flow resulting from this credi le transient at 100 p g back pressure, ich is less than the maximum RCS pressure on the /T limit curve in LC 3.4.3.

The RCS ven size will also be re-e aluated for complia ce each time P/T limit curves a revised based on th results of the vess I material surveillance.

The vent is p ssive and is not subj to active failure.

The LTOP S stem satisfies Criterio 2 of 10 CFR 50.36 c)(2)(ii).

BWOG STS B 3.4.12-5 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 0, Page 262 of 415

Attachment 1, Volume 9, Rev. 0, Page 263 of 415 I All changes are unless otherwise noted 1Low Temper re Overpressure P ection TOP] t B 3.4.12 BASES LCO The LCO requires an LTOP System O ERABLE with a li ted coolant input capability nd a pressure relief pability. To limit co lant input, the LCO requires a aximum of [one] ma eup pump OPERAOLE, the HPI deactivated, an the CFT discharge is lation alves close and immobilized. Fqr pressure relief, it re dires either the pre surizer coolant at or below a ximum level and the RV OPERABLEO th a lift setting INSERT 3 at the LTOP Ii t or the RCS depress rized and a vent es ablished.

The LCO is mo ified by two Notes. 1 allows (two m keup pumps] to be made capa le of injecting for < 1 our during pump sw p operations.

One hour provi es sufficient time to s fely complete the a ual transfer and to complet the administrative ontrols and surveillan requirements at..

associated wit the swap. The intent is to minimize the a ual time that more than [on ] makeup pump is ph sically capable of in ection. Note 2 states that CF isolation is only requ ed when the CFT p essure is more than or equal t the maximum RCS ressure for the exis ng RCS temperature, a allowed in LCO 3.4. . This Note permits the CFT discharge valv surveillance perfor d only under these pressure and temperature nditions.

The pressuriz r is OPERABLE with coolant level _ [22 ] inches.

The PORV is PERABLE when its lock valve is open, s lift setpoint is set at _ [555] sig and testing has p oven its ability to o n at that setpoint, and tive power is availa le to the two valve and their control circuits.

For the depr ssurized RCS, an RC vent is OPERABL when open with an area of at east [0.75] square in es.

APPLICABILITY S This LCO is applicable in MODE4 hen ay RCS cold leg te,,perature is

[* *~ ~ ] F, in M9DtE 51 and in MODE 6 when the reactor vessel head is

[2.5 ZLo-n-'. Th~e Applicability Itempera.lee of [283]7F is established by fracture a mechanics analyses. The pressurizer safety valves provide overpressure protection to meet LCO 3.4.3 P/T limitslabov e283]°F. With the vessel head off, overpressurization is not possible. I in MODES 1, 2,and 3 LCO 3.4.3 provides the operational P/T limits for all MODES.

LCO 3.4.10, "Pressurizer Safety Valves," requires the pressurizer safety valves OPERABLE to provide overpressure protection during MODES 1, 2, and I and MODE-4 above [2831.

BWOG STS B 3.4.12-6 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 0, Page 263 of 415

Attachment 1, Volume 9, Rev. 0, Page 264 of 415 B 3.4.12 (O INSERT 3 For low temperature overpressure protection, Davis-Besse relies on the four-inch DHR System relief valve (DH-4849) with a lift setpoint - 330 psig. This relief valve is located on the DHR System suction line from the RCS. The RCS to DHR System isolation valves (DH-1 1 and DH-1 2) must be open and control power removed from the valve operators for the DHR System relief valve to be OPERABLE. Control power can be removed either in the control room or at the motor control center (by removing fuses, opening breakers, or racking breakers out).

Insert Page B 3.4.12-6 Attachment 1, Volume 9, Rev. 0, Page 264 of 415

Attachment 1, Volume 9, Rev. 0, Page 265 of 415 All changes are unless otherwise noted 1Low Temper.re Overpressure Prection TOPI er B 3.4.12 BASES ACTIONS A.1 and B.1 With two or mor makeup pumps cap ble of injecting into he RCS or if the HPI is activ ted, immediate action are required to re der the other pump(s) inoper, ble or to deactivate HPI. Emphasis is on mmediate deactivation be use inadvertent inje tion with [one] or m re HPI pump OPERABLE is he event of greatest s gnificance, since it uses the INsERT 4 greatest press re increase in the sho est time. Also, the vent cannot mitigate overpr ssurization from the i jection of even one HPI pump.

The immediate Completion Times ref ect the urgency of q ickly proceeding wit the Required Action C.1 0.1 and .2 An unisolated FT requires isolation within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months only wen the CFT pressure is at r more than the maxi uum RCS pressure 'or the existing temperature al owed in LCO 3.4.3.

If isolation is n eded and cannot be ccomplished in 1 h ur, Required Action D.1 an Required Action D.2 rovide two options, either of which must be perfo med in 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . By ncreasing the RCS emperature to

> 175°F, the FT pressure of 600 p ig cannot exceed th LTOP limits if both tanks ar fully injected. Depre surizing the CFTs b low the LTOP limit of [555] ig also prevents ex eding the LTOP lim s in the same event.

The Completi n Times are based o operating experien that these activities can e accomplished in th se time periods an on engineering evaluations i dicating that a limiting LTOP event is not I ely in the allowed time.

E.1 F.1 and F.2 With the pre surizer level more thai [220] inches, the ti e for operator acton in a pr ssure increasing eve t is reduced. The stulated event most affecte in the LTOP MODE1 is failure of the ma eup control valve, which fills th pressurizer relatively rapidly. Restoratio is required within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

BWOG STS B 3.4.12-7 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 0, Page 265 of 415

Attachment 1, Volume 9, Rev. 0, Page 266 of 415 B 3.4.12 O* INSERT 4 A.1 and A.2 With the DHR System relief valve inoperable due to one or both RCS to DHR System isolation valves closed, the overpressure protection flow path is isolated. The flow path must be restored by opening the RCS to DHR System isolation bypass valves (DH-21 and DH-23),

within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . After opening, the RCS to DHR System isolation bypass valves must be verified open every 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

The 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months Completion Time reflects the importance of the action and provides time for a timely opening of the RCS to DHR System isolation bypass valves. To ensure they remain in the open position, the positions of the RCS to DHR System isolation bypass valves are required to be verified every 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . RCS to DHR System isolation bypass valves are manual valves and do not have remote position indication.

B.1 With control power available to one or both of the RCS to DHR System isolation valves, the overpressure protection flow path could be inadvertently isolated. The control power must be removed from the valves within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months to ensure the valves will remain open during system operation.

The 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months Completion Time reflects the importance of the action and provides time for a timely removal of control power.

C.1 If the DHR System relief valve is inoperable for reasons other than the relief flow path (Condition A or B), the DHR System relief valve must be restored to OPERABLE status within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . The 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months Completion Time is acceptable due to the low probability of an overpressure event.

D.1, D.2, D.3, and D.4 If any Required Action and Completion Time of Condition A, B, or C is not met, other compensatory actions must be taken to minimize the probability and consequences of an LTOP event. Without an OPERABLE relief path for overpressure protection, the RCS water addition capabilities must be limited. Within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months both HPI pumps must be disabled (e.g., by opening motor supply breakers), and within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months the makeup pump suction automatic transfer to the borated water storage tank on low makeup tank level must be disabled.

Makeup tank level must be verified to be < 73 inches within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months to minimize volume.

Furthermore, without an overpressure relief path, RCS pressure and pressurizer level must be verified to be in the Acceptable Region of Figure 3.4.12-1 or 3.4.12-2 (depending on the MODE) within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months to ensure an overpressure condition cannot occur. These Figures do not include instrument error uncertainties.

Insert Page B 3.4.12-7 Attachment 1, Volume 9, Rev. 0, Page 266 of 415

Attachment 1, Volume 9, Rev. 0, Page 267 of 415 I All changes are a unless otherwise noted 9 1Low Terner.tre Overpressure P er ction TOPlM em B 3.4.12 BASES ACTIONS (continued)

If restoration Required wit inF.11 hour1.273148e-4 days 0.00306 hours 1.818783e-5 weeks 4.1855e-6 months Actions in either and F.2 must ca b eperformed acco 1;pished, cannot bewithin hours to close the makeup control valve and its isola :ion valve. These Required Actions limit the makeu. capability, which is nt required with a hi ;h pressurizer level, and permi cooldown and depre surization to continue. Heatup must be stoppe Ibecause heat additi, n decreases the re *ctor coolant The Completior Times again are bas .d on operating exp rience that these activities *n be accomplished;ure in these time period and on engineering ev; luations indicating th ,t a limiting LTOP tr nsient is not likely in the allo Wed times G.1.H.1.and F.2 With the PORK restoration of t! einoperable, PORV within overpre ur required.

1 ho r"is If that ility relieving capa cannotis lost, be and accomplished, the ability of the Mak* up System to add

ter must be limited within ti e next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

if restoration cnnot be completed v thin 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , Requir d Action H.1 and Required Acti n H.2 must be pefre olimit RCS wa :er addition capability. M* keuPllis not deactivat d1to maintain the R, S coolant level.

Required Acti n H.1 and Required Jction H.2 require r* ucing the makeup tank I vel to 70 inches and deactivating the Io* low makeup tank level interlock to the borated water *.torage tank. T'his nakes the available mak .up water volume ins ifficient to exceed t* e LTOP limit by a makeup contr I>valve full opening.

These Compl ztion Times also cons der these activities *n be accomplishe* in these time periods. A limiting LTOP e *ent is not likely in those times.

Some PORV testing or maintenanc e can only be perfo meed at plant shutdown. S ch activity is permitt d if Required Actior H.1 and Required Action H.2 artknt compensat for PORV unavail* bility.

BWOG STS B 3.4.12-8 Rev. 3.0, 03131/04 Attachment 1, Volume 9, Rev. 0, Page 267 of 415

Attachment 1, Volume 9, Rev. 0, Page 268 of 415 All changes are ( y unless otherwise noted 9 JLow Ternperre Overpressure P ection TOPIM B 3.4.12 BASES ACTIONS (continued) 1.1 With the pressu izer level above [220 inches and the PO V inoperable or the LTOP S tem inoperable for a y reason other than cited in Condition A thr ugh H, Required Acti n 1.1 requires the R S depressurized nd vented within 12 h urs from the time either Condition started.

One or more v nts may be used. A ý nt size of a [0.75] quare inches is specified. Thi vent size assumes 1 0 psig backpressur . Because makeup may required, the vent si e accommodates i advertent full makeup syste operation. Such a v nt keeps the press re from full flow of [one] make p pump with a wide o en makeup control alve within the LCO limit.

The PORV ha a larger area and y be used for venti by opening and locking it oper This size RC vent or the PORVs a vent cannot maintai RCS pressure below LTOP I mits if the HPI and C T systems are nad ertently actuated.

Therefore, ve' ification of t cti ation of two HPI pu ps, HPI injection, and the CFT must accompany the epressurizing and enting. Since these ,ster are required deactiv ted by the LCO, S 3.4.12.1, SR 3.4.12.2, nd SR 3.4.12.3 requi e verification of thei deactivated status every 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

The Comple on Time is based on perating experien that this activity can be acco plished in this time p riod and on engine ring evaluations indicating th t a limiting LTOP tran ient is not likely in t is time.

SURVEILLANCE SR 3.4.12.1 S 3.4.12.2 andSR 34.12.3 REQUIREMENTS Verifications m st be performed that nly [one] makeup p mp is capable of injecting into the RCS, the HPI is activated, and the FT discharge isolation valve are closed and immo ilized. These Surv illances ensure the minimum olant input capability will not create an R S overpressure condition to allenge the LTOP Sy em. The Surveilla ces are required at 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months int rvals.

The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months i tervals are shown by operating practice t be sufficient to regularly ass ss conditions for pot tial degradation an verify operation within the sa ety analysis.

BVVOG STS B 3.4.12-9 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 0, Page 268 of 415

Attachment 1, Volume 9, Rev. 0, Page 269 of 415 I All changes are unless otherwise noted 9 ILow Temper re Overpressure P ection TOPIe B 3.4.12 BASES SURVEILLANCE REQUIREMENTS (continued)

SR 3.4.12.4 Verification of t pressurizer level at [220] inches by ob erving control room or other in ications ensures a c shion of sufficient s e is available to reduce the ra e of pressure increas from potential tran ients.

The 30 minute urveillance Frequen y during heatup an cooldown must be performed f r the LCO Applicabili period when temp rature changes can cause pre surizer level variation This Frequency ay be discontinued en the ends of these conditions are satis ied, as defined in plant proce ures. Thereafter, the urveillance is requ red at 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months intervals.

These Frequ ncies are shown by o erating practice su icient to regularly assess indica ions of potential degr dation and verify o eration within the safety analys s.

Eý' -*

SR 3.4.12.5 Verification thatihe PORV block valv is open ensures a f ow path to the PORV. This isquired at 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months in rvals.

T5 The interval h s been shown by ope/ating practice suffic ent to regularly assess condit for onspotential degr~ dation and verify o eration is within the safety an lysis. /

SR 3.42.

of ThethisROS LCO.ventF ~r f aatvent leastvalve

[0.75] squar inches must be ve ified open for 12 hours.

relief protection not is beiIoc!ed open,tothe Valv only s thatif the are vent g used sealed or *ecured satisfy Frequth ncy requirements is every considered "lo :ked" in the open position valve that is Io in this context. are Iorother vent path(*

safety valve, r bked, sealed, or secur in position, a rerr (e.g., a vent

d pen manway), the equired Frequency il ved pressurizer every 31 days.

Again, the Fr quency intervals consider adequacy to *egularly operating practie assess condi ions to determine verify operati for potential de n within the safety lradation and a nalysis.

BWOG STS B 3.4.12-10 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 0, Page 269 of 415

Attachment 1, Volume 9, Rev. 0, Page 270 of 415 B 3.4.12 O* INSERT 5 Verification of the flow path from the RCS to the DHR System relief valve is required every 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . This verification is performed by checking RCS to DHR System isolation valves in the open position with control power removed from the valve operator. This Surveillance ensures the overpressure relief flow path is aligned and remains aligned. Removal of control power ensures the flow path is not inadvertently closed.

The Frequency is adequate based on operating experience. Manual operation is required to close the isolation valves or energize control power. Valve operations are administratively controlled by procedure. In this configuration the isolation valves will not inadvertently close.

Insert Page B 3.4.12-10 Attachment 1, Volume 9, Rev. 0, Page 270 of 415

Attachment 1, Volume 9, Rev. 0, Page 271 of 415 I

All changes are unless otherwise noted 9 1Low Temper;re Overpressure Pection TOP tEr B 3.4.12 BASES SURVEILLANCE REQUIREMENTS (continued)

The passive v,0*t path arrangement"%ust only be open/to'be OPERABLE.

SR 3.4.12 CHANNEL F NCTIONAL TEST is r quired within [12] h urs after decreasing RC temperature to - [28 ]°F and every 31 da s thereafter to changes not shown ensure the setp int is proper for using the PORV for LTO . A successful test of the requi ed contact(s) of a ch nnel relay may be p rformed by the verification of t e change of state of a single contact of th relay. This clarifies what is an acceptable CHAN EL FUNCTIONAL EST of a relay.

This is accepta le because all of the ther required conta ts of the relay are verified by ther Technical Speci ications and non-Te hnical Specifications ests at least once per refueling interval wi applicable extensions. RV actuation is not eeded, as it could d pressurize the RCS.

The [12] hour Frequency considers he unlikelihood of a ow temperature overpressure vent during the time. The 31 day Freque cy is based on industry acce ted practice and is a ptable by experie ce with equipment re iability. I SR 3.4.12.8 The performan e of a CHANNEL CA IBRATION is requir d every

[18] months. T e CHANNEL CALIB ATION for the LTO setpoint ensures that th PORV will be actuat d at the appropriat RCS pressure by verifying th accuracy of the instr ment string. The c libration can only be perfor ed in shutdown.

The Frequen y considers a typical r fueling cycle and i dustry accepted practice.

REFERENCES 1. 10 CFR 50, Appendix G.

2. Generic Letter 88-11.

Z ý 3. FSAR, Section M - r gi 5 0

4. 10CFR 0.46. 0

5. '10 C7~0, Appendix K BWOG STS B 3.4.12-11 Rev. 3.0, 03131/04 Attachment 1, Volume 9, Rev. 0, Page 271 of 415

Attachment 1, Volume 9, Rev. 0, Page 272 of 415 B 3.4.12 0 INSERT 6 Verification of the DHR System relief valve lift setpoint must be performed to ensure LTOP requirements can be met. Overpressure protection of the RCS is ensured by the DHR System relief valve, which relieves pressure and prevents the RCS from exceeding the Pressure/Temperature Limits.

The DHR System relief valve setpoint is verified in accordance with the Inservice Testing (IST) Program forproper operation and correct lift setting of < 330 psig.

This lift setting pressure shall correspond to ambient conditions of the valve at nominal operating temperature and pressure. The IST Program specifies the testing and frequency, as directed by ASME Code.

Insert Page 3.4.12-11 Attachment 1, Volume 9, Rev. 0, Page 272 of 415

Attachment 1, Volume 9, Rev. 0, Page 273 of 415 JUSTIFICATION FOR DEVIATIONS ITS 3.4.12 BASES, LOW TEMPERATURE OVERPRESSURE PROTECTION (LTOP)

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 changes made to the Specification.

3. 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.

4. The Davis-Besse methods for LTOP have been included. With RCS temperature between approximately 500'F and 280'F, pressurizer safety valves cannot provide overpressure protection; LTOP is provided by operating procedures. Below 140'F, credible overpressurization sources are secured. These methods for LTOP have been previously reviewed and approved by the NRC, as documented in the NRC Safety Evaluation for Amendment 199, dated July 20, 1995.

Davis-Besse Page 1 of 1 Attachment 1, Volume 9, Rev. 0, Page 273 of 415

Attachment 1, Volume 9, Rev. 0, Page 274 of 415 Specific No Significant Hazards Considerations (NSHCs)

Attachment 1, Volume 9, Rev. 0, Page 274 of 415

Attachment 1, Volume 9, Rev. 0, Page 275 of 415 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.4.12, LOW TEMPERATURE OVERPRESSURE PROTECTION (LTOP)

There are no specific NSHC discussions for this Specification.

Davis-Besse Page 1 of 1 Attachment 1, Volume 9, Rev. 0, Page 275 of 415

, Volume 9, Rev. 0, Page 276 of 415 ATTACHMENT 13 ITS 3.4.13, RCS OPERATIONAL LEAKAGE , Volume 9, Rev. 0, Page 276 of 415

, Volume 9, Rev. 0, Page 277 of 415 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) , Volume 9, Rev. 0, Page 277 of 415

Attachment 1, Volume 9, Rev. 0, Page 278 of 415 IITS 3.4.13 IT._GS REACTOR COOLANT SYSTEM OPERATIONAL LEAKAGE LIMITING CONDITION FOR OPERATION 3.4.13 3.4.6.2 Reactor Coolant System operational leakage shall be limited to:

a. No PRESSURE BOUNDARY LEAKAGE,

b. 1 GPM UNIDENTIFIED LEAKAGE,

c. 150 gallons per day primary to secondary leakage through any one steam generator (SG),

d. 10 GPM IDENTIFIED LEAKAGE from the Reactor Coolant System,

e. 10 GP/M CONTROLLED LE 'AGE, and - - )

f" 5 GPM leakage from any Reactor Coolant System Pressure Isolation Valve as specified in Table 3.4-2. H See ITS 3.4.14J APPLICABILITY: MODES 1, 2, 3 and 4

ACTION:

ACTION B a. With any PRESSURE BOUNDARY LEAKAGE, or with primary to secondary leakage not within limit, be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN Within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

ACTION A b. With any Reactor Coolant System operational leakage greater than any one of the above limits, excluding PRESSURE BOUNDARY LEAKAGE or primary to secondary leakage, reduce the leakage rate to within limits within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or-be in at ACTION B ~ileast HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within d.TION The following 30 hour3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months slexcept reas permitted ry paragrayphintO3.c 3.elow witin4 In the event that integrity of any pressure isolation valves in Table 3.4-2 cannot be demonstrated, POWER OPERATION may continue, provided that at least two valves in each high pressure line having a non-functional valve are in and remain in, tihe mode corresponding to the isolated condition.(a) See3.4I4TS

d. The provisions of Section 3.0.4 are not applicable for entry into MODES 3 and 4 for the purpose of testing the isolation valves in Table 3.4-2.

I'Motor operated valves shall be placed in the closed position and power supplies deenergized. L S-Iee ISl 3.4.14 DAVIS-BESSE, UNIT 1 3/4 4-15 Otde .4ý20*&l-Amendment No. -135,-F,220-,, 276 Page 1 of 2 Attachment 1, Volume 9, Rev. 0, Page 278 of 415

Attachment 1, Volume 9, Rev. 0, Page 279 of 415 ITS 3.4.13 ITS REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS 4.4.6.2.1 Reactor Coolant System operational leakages shall be demonstrated to be within each of the above limits by:

a. Monitoring the containment atmosphere gaseous or particulat* radioactivity at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . / S0

b. Monitoring the containment sump level and flow indicationat least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

c. Measurement of the NTROLLED LEAKAGE from the eactor coolant pump seals L01 to the makeup syst when the Reactor Coolant System p essure is 2185 +/- 20 psig at least once 31 da .

SR 3.4.13.1 d. Performance of a Reactor Coolant System water inventory balance at least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months during steady state operation. (1)(2)

SR 3.4.13.2 e. Verifying that primary to secondary leakage is < 150 gallons per day through any one steam generator, at least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . (2) 4.4.6.2.2 Each Reactor Coolant System Pressure Isolation Valve specified in Table 3.4-2 shall be individually demonstrated OPERABLE by verifying leakage testing (or the equivalent) to be within its limit prior to entering MODE 2:

a. After each refueling outage,

b. Whenever the plant has been in COLD SHUTDOWN for 7 days, or more, and if leakage testing has not been performed in the previous 9 months, and

c. Prior to returning the valve to service following maintenance, repair or replacement See ITS 1 work on the valve. 3.4.14 ]

d. The provisions of Specification 4.0.4 are not applicable for entry into MODES 3 or 4.

4.4.6.2.3 Whenever the integrity of a pressure isolation valve listed in Table 3.4-2 cannot be demonstrated, determine and record the integrity of the high pressure flowpath on a daily basis.

Integrity shall be determined by performing either a leakage test of the remaining pressure isolation valve, or a combined leakage test of the remaining pressure isolation valve in a series with the closed motor-operated containment isolation valve. In addition, record the position of the closed motor-operated containment isolation valve located in the high pressure piping on a daily basis.

SR 3.4.13.1 NOTE 2 () Not applicable to primary to secondary leakage.

SR 3.4.13.1 NOTE 1, (2) Not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after establishment of steady state operation.

SR 3.4.13.2 NOTE DAVIS-BESSE, UNIT I 3/4 4-16 -Order-deted-420/ Amendment No. $4,-I3rl-Or.4~gf22O, 276 0

Page 2 of 2 Attachment 1, Volume 9, Rev. 0, Page 279 of 415

Attachment 1, Volume 9, Rev. 0, Page 280 of 415 DISCUSSION OF CHANGES ITS 3.4.13, RCS OPERATIONAL LEAKAGE 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.

MORE RESTRICTIVE CHANGES None RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES None LESS RESTRICTIVE CHANGES L01 (Category 1 - Relaxation of LCO Requirements) CTS 3.4.6.2.e requires that Reactor Coolant System leakage shall be limited to 10 gpm of CONTROLLED LEAKAGE. CTS 4.4.6.2.1 .c requires a verification that the CONTROLLED LEAKAGE is within the limit every 31 days. ITS LCO 3.4.13 does not retain these requirements. This changes the CTS by deleting this LCO requirement.

The purpose of CTS 3.4.6.2.e and its associated Surveillance is to ensure the CONTROLLED LEAKAGE does not exceed a specified limit. CONTROLLED LEAKAGE is seal water flow from the reactor coolant pumps seals. The CTS 3.4.6.2.e limit of 10 gpm is the design leakage rate through the pump seals and back to the makeup tank via the seal return lines. Thus a higher flow rate would indicate that the pumps seals are deteriorated or failed. However, a maximum seal water leakage (i.e. flow) is not an assumption of any accident or transient analysis, which is the reason it has been maintained in another pressurized water reactor ISTS (NUREG-1431, "Standard Technical Specifications - Westinghouse Plants," ISTS 3.5.5). For Davis-Besse, there is no need to quantify the normal CONTROLLED LEAKAGE since it is normal system operation and there is no loss from the RCS inventory. Furthermore, if the seal water flow increases greater than the current 10 gpm limit due to an upper seal failure, the increased flow would be directed to the containment normal sump.

The containment normal sump is the collecting sump that identified LEAKAGE is quantified (and limited to 10 gpm). Thus the increased seal water flow resulting Davis-Besse Page 1 of 2 Attachment 1, Volume 9, Rev. 0, Page 280 of 415

Attachment 1, Volume 9, Rev. 0, Page 281 of 415 DISCUSSION OF CHANGES ITS 3.4.13, RCS OPERATIONAL LEAKAGE from a failed or leaking upper seal would be detected and proper actions taken as necessary. Therefore this change is acceptable and is designated as less restrictive because an LCO requirement required in the CTS will not be required in the ITS.

L02 (Categoty 5 - Deletion of Surveillance Requirement) CTS 4.4.6.2.1 .a requires monitoring of the containment atmosphere gaseous or particulate radioactivity at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . CTS 4.4.6.2.1.b requires monitoring the containment sump level and flow indication at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . The ITS does not contain these Surveillance Requirements. This changes the CTS by eliminating these Surveillance Requirements.

This change is acceptable because the deleted Surveillance Requirements are not necessary to verify that the LCO is being met. Thus, appropriate Surveillance Requirements continue to be performed in a manner and at a Frequency necessary to give confidence that the LCO is being met. The indications in the deleted Surveillance Requirements are not necessarily indications of failure to meet the LCO on RCS operational LEAKAGE. These items do provide useful information and the containment atmosphere particulate monitor and the containment sump monitors are required to be OPERABLE and tested by ITS 3.4.15, "RCS Leakage Detection Instrumentation." However, under ITS SR 3.0.1, failure to meet the Surveillance results in failure to meet the LCO. As these indications do not necessarily indicate a failure to meet the LCO, it is not appropriate to retain these indications in this Specification. This change is designated as less restrictive because Surveillances which are required in the CTS will not be required in the ITS.

Davis-Besse Page 2 of 2 Attachment 1, Volume 9, Rev. 0, Page 281 of 415

Attachment 1, Volume 9, Rev. 0, Page 282 of 415 Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs)

Attachment 1, Volume 9, Rev. 0, Page 282 of 415

Attachment 1, Volume 9, Rev. 0, Page 283 of 415 CTS RCS Operational LEAKAGE 3.4.13 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.13 RCS Operational LEAKAGE 3.4.6.2 LCO 3.4.13 RCS operational LEAKAGE shall be limited to:

a. No pressure boundary LEAKAG

b. 1 gpm unidentified LEAKAG 0

c. 10 gpm identified LEAKAG and

d. 150 gallons per day primary to secondary LEAKAGE through any one steam generator (SG).

APPLICABILITY: MODES 1, 2, 3, and 4.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME Action b A. RCS operational A.1 Reduce LEAKAGE to within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months LEAKAGE not within limits.

limits for reasons other than pressure boundary LEAKAGE or primary to secondary LEAKAGE.

Action a, B. Required Action and B.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months Action b associated Completion Time of Condition A not AND met.

B.2 Be in MODE 5. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months OR Pressure boundary LEAKAGE exists.

OR Primary to secondary LEAKAGE not within limit.

BVVG STS 3.4.13-1 Rev. 3.1,12/01/05 Attachment 1, Volume 9, Rev. 0, Page 283 of 415

Attachment 1, Volume 9, Rev. 0, Page 284 of 415 CTS RCS Operational LEAKAGE 3.4.13 SURVEILLANCE REQUIREMENTS

-SURVEILLANCE FREQUENCY 4.4.6.2.1 .d SR 3.4.13.1 -... -...------ -.-- NOTES----------- -.-------..

(including 1. Not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months footnotes after establishment of steady state operation.

(1) and (2))

2. Not applicable to primary to secondary LEAKAGE.

Verify RCS operational LEAKAGE is within limits by 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months performance of RCS water inventory balance.

4.4.6.2. 1.e R 3.4.13.2 -.-.- ----- -......---NOTE -............----- -----

(including Not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after footnote (2)) establishment of steady state operation.

Verify primary to secondary LEAKAGE is 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months

<_150 gallons per day through any one SG.

BWOG STS 3.4.13-2 Rev. 3.1,12/01/05 Attachment 1, Volume 9, Rev. 0, Page 284 of 415

Attachment 1, Volume 9, Rev. 0, Page 285 of 415 JUSTIFICATION FOR DEVIATIONS ITS 3.4.13, RCS OPERATIONAL LEAKAGE

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.

Davis-Besse Page 1 of 1 Attachment 1, Volume 9, Rev. 0, Page 285 of 415

of 415 1, Volume 9, Rev. 0, Page 286 Attachment Bases Technical Specifications (ISTS) 10 Improved Standard Markup Deviations (JFDs) and Justification for of 415 1, Volume 9, Rev. 0, Page 286 Attachment

Attachment 1, Volume 9, Rev. 0, Page 287 of 415 RCS Operational LEAKAGE B 34.13 B 3.4 REACTOR COOLANT SYSTEM (RCS)

B 3.4.13 RCS Operational LEAKAGE BASES BACKGROUND Components that contain or transport the coolant to or from the reactor core make up the RCS. Component joints are made by welding, bolting, rolling, or pressure loading, and valves isolate connecting systems from the RCS.

During plant life, the joint and valve interfaces can produce varying amounts of reactor coolant LEAKAGE, through either normal operational wear or mechanical deterioration. The purpose of the RCS Operational LEAKAGE LCO is to limit system operation in the presence of LEAKAGE from these sources to amounts that do not compromise safety. This LCO specifies the types and amounts of LEAKAGE.

10 CFR 50, Appendix A, GDC 30 (Ref. 1), requires means for detecting Although not and, to the extent practical, identifying the source of reactor coolant committed to LEAKAGE. Regulatory Guide 1.45 (Ref. 2) escribes acceptable methods for selecting Leakage Detection Sys em The safety significance of RCS LEAKAGE varies widely depending on its source, rate, and duration. Therefore, detecting and monitoring reactor coolant LEAKAGE into the containment area is necessary. Quickly separating the identified LEAKAGE from the unidentified LEAKAGE is necessary to provide quantitative information to the operators, allowing them to take corrective action should a leak occur detrimental to the safety of the facility and the public.

A limited amount of leakage inside containment is expected from auxiliary systems that cannot be made 100% leaktight. Leakage from these systems should be detected, located, and isolated from the containment atmosphere, if possible, to not interfere with RCS leakage detection.

This LCO deals with protection of the reactor coolant pressure boundary (RCPB) from degradation and the core from inadequate cooling, in addition to preventing the accident analysis radiation release assumptions from being exceeded. The consequences of violating this LCO include the possibility of a loss of coolant accident (LOCA). However, the ability to monitor leakage provides advance warning to permit plant shutdown before a LOCA occurs. This advantage has been shown by "leak before break" studies.

BWAOG STS B 3.4.13-1 Rev. 3.1, 12/01/05 Attachment 1, Volume 9, Rev. 0, Page 287 of 415

Attachment 1, Volume 9, Rev. 0, Page 288 of 415 RCS Operational LEAKAGE B 3.4.13 BASES APPLICABLE -Except for primary to secondary LEAKAGE, the safety analyses do not SAFETY address operational LEAKAGE. However, other operational LEAKAGE ANALYSES is related to the safety analyses for LOCA; the amount of leakage can affect the probability of such an event. The safety analysis for an event resulting in steam discharge to the atmosphere assumes that primary to secondary LEAKAGE from all steam generators (SGs) isljl gallon per minute]or increases topJ1 gallon per minutef~as a resultof accident induced conditions. The LCO requirement to limit primary to secondary LEAKAGE through any one SG to less than or equal to 150 gallons per day is significantly less than the conditions assumed in the safety analysis.

Primary to secondary LEAKAGE is a factor in the dose releases outside M containment resulting from asteam line break SLB) accident. To a lesser extent, other accidents or transients involve secondary steam 0

release to the atmosphere, such as a steam generator tube rupture (SGTR). The leakage contaminates the secondary fluid.

TFSAR (Ref. 3) analysis for SGTR assumes the contaminated secondary fluid is only briefly released via safety valves and the majority is steamed to the condenser. The[p gprn]primary to secondary E

LEAKAGE safety analysis assumption is relatively inconsequential.

B*more limitin or site radiatn releases.m The safety analysis 0

LB accident assumes the entire [DJ1 gprrl primary to secondary LEAKAGE is through the affected generator as an initial condition. The dose consequences resulting from the LB accident are well within the limits defined in 10 CFR 100.

RCS operational LEAKAGE satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii).

LCO RCS operational LEAKAGE shall be limited to:

a. Pressure Boundary LEAKAGE No pressure boundary LEAKAGE is allowed, being indicative of material deterioration. LEAKAGE of this type is unacceptable as the leak itself could cause further deterioration, resulting in higher LEAKAGE. Violation of this LCO could result in continued degradation of the RCPB. LEAKAGE past seals and gaskets is not pressure boundary LEAKAGE.

BWOG STS B 3.4.13-2 Rev. 3.1, 12/01105 Attachment 1, Volume 9, Rev. 0, Page 288 of 415

Attachment 1, Volume 9, Rev. 0, Page 289 of 415 RCS Operational LEAKAGE B 3.4.13 BASES LCO (continued)

b. Unidentified LEAKAGE One gallon per minute (gpm) of unidentified LEAKAGE is allowed as a reasonable minimum detectable amount that the containment air monitoring and containment sump level monitoring equipment can detect within a reasonable time period. Violation of this LCO could result in continued degradation of the RCPB, if the LEAKAGE is from the pressure boundary.

c. Identified LEAKAGE Up to 10 gpm of identified LEAKAGE is considered allowable because LEAKAGE is from known sources that do not interfere with detection of unidentified LEAKAGE and is well within the capability of the RCS makeup system. Identified LEAKAGE includes LEAKAGE to the containment from specifically known and located sources, but r does f not include pressure boundary LEAKAGE or controlled reactor coolant pump (RCP) sea le off (a normal function not considered LEAKAGE). Violation of this LCO could result in continued degradation of a component or system.

d. Primary to Secondary LEAKAGE Through Any One SG The limit of 150 gallons per day per SG is based on the operational LEAKAGE performance criterion in NEI 97-06, Steam Generator Program Guidelines (Ref. 4). The Steam Generator Program operational LEAKAGE performance criterion in NEI 97-06 states, "The RCS operational primary to secondary leakage through any one SG shall be limited to 150 gallons per day." The limit is based on operating experience with SG tube degradation mechanisms that result in tube leakage. The operational leakage rate criterion in conjunction with the implementation of the Steam Generator Program is an effective measure for minimizing the frequency of steam generator tube ruptures.

APPLICABILITY In MODES 1, 2, 3, and 4, the potential for RCPB LEAKAGE is greatest when the RCS is pressurized.

In MODES 5 and 6, LEAKAGE limits are not required because the reactor coolant pressure is far lower, resulting in lower stresses and reduced potentials for LEAKAGE.

BWOG STS B 3.4.13-3 Rev. 3.1, 12/01/05 Attachment 1, Volume 9, Rev. 0, Page 289 of 415

Attachment 1, Volume 9, Rev. 0, Page 290 of 415 RCS Operational LEAKAGE B 3.4.13 BASES APPLICABILITY (continued)

LCO 3.4.14, "RCS Pressure Isolation Valve (PIV) Leakage," measures leakage through each individual PIV and can impact this LCO. Of the two PIVs in .series in each isolated line, leakage measured through one PIV does not.result in RCS LEAKAGE when the other is leaktight. If both valves leak and result in a loss of mass from the RCS, the loss must be included in the allowable identified LEAKAGE.

ACTIONS A.1 If unidentified LEAKAGE or identified LEAKAGE are in excess of the LCO limits, the LEAKAGE must be reduced to within limits within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . This Completion Time allows time to verify leakage rates and either identify unidentified LEAKAGE or reduce LEAKAGE to within limits before the reactor must be shut down. This action is necessary to prevent further deterioration of the RCPB.

B.1 and B.2 If any pressure boundary LEAKAGE exists or primary to secondary LEAKAGE is not within limit, or if unidentified or identified LEAKAGE cannot be reduced to within limits within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , the reactor must be brought to lower pressure conditions to reduce the severity of the LEAKAGE and its potential consequences. The reactor must be brought to MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and MODE 5 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . This action reduces the LEAKAGE and also reduces the factors that tend to degrade the pressure boundary.

The Completion Times allowed are reasonable, based on operating experience, to reach the required conditions from full power conditions in an orderly manner and without challenging plant systems. In MODE 5, the pressure stresses acting on the RCPB are much lower and further deterioration is much less likely.

SURVEILLANCE SR 3.4.13.1 REQUIREMENTS Verifying RCS LEAKAGE within the LCO limits ensures that the integrity of the RCPB is maintained. Pressure boundary LEAKAGE would at first appear as unidentified LEAKAGE and can only be positively identified by inspection. Unidentified LEAKAGE and identified LEAKAGE are determined by performance of an RCS water inventory balance.

The RCS water inventory balance must be performed with the reactor at r-*--- steady state operating conditions (stable temperature, power level,

-- pressurizer and makeup tank levels, m and letdown, [and RCP seall 0

BWOG STS B 3.4.13-4 Rev. 3.1, 12/01/05 Attachment 1, Volume 9, Rev. 0, Page 290 of 415

Attachment 1, Volume 9, Rev. 0, Page 291 of 415 RCS Operational LEAKAGE B 3.4.13 BASES SURVEILLANCE REQUIREMENTS (continued) linjection andret m flows . The Surveillance is modified by two Notes. \ k, 2 Note 1 states that this SK is not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after establishing steady state operation. The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months allowance provide Isufficient time to collect and process all necessary -dataafter stable plant\

(

/The accuracy of the conditions are established.

results will be impacted if any measured parameter s used to calculate the RCS eady operation is require o perform a proper Pater inven ory LEAKAGE is not ina lbalance pfnce calculations durirA maneuvering are not useful. or steady state condition. paaan on ppera etermination y/er inven ory e, stead state is defin d as stable RCS pressur%, temperature, power evel, pressurizer nd makeup returntank levels, flows. r!keup and letdown, ad RCP

ýpump seal ýniection and An early warning of pressure boundary LEAKAGE or unidentified LEAKAGE is provided by the automatic systems that monitor the containment atmosphere radioactivity and the containment sump level.

These leakage detection systems are specified in LCO 3.4.15, "RCS Leakage Detection Instrumentation."

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

The 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Frequency is a reasonable interval to trend LEAKAGE and recognizes the importance of early leakage detection in the prevention of accidents.

SR 3.4.13.2 This SR verifies that primary to secondary LEAKAGE is less than or equal to 150 gallons per day through any one SG. Satisfying the primary to secondary LEAKAGE limit ensures that the operational LEAKAGE performance criterion in the Steam Generator Program is met. Ifthis SR is not met, compliance with LCO 3.4.17, "Steam Generator Tube Integrity," should be evaluated. The 150 gallons per day limit is measured at room temperature as described in Reference 5. The operational LEAKAGE rate limit applies to LEAKAGE through any one SG. If it is not practical to assign the LEAKAGE to an individual SG, all the primary to secondary LEAKAGE should be conservatively assumed to be from one SG.

BWOG STS B 3.4.13-5 Rev. 3.1, 12/01/05 Attachment 1, Volume 9, Rev. 0, Page 291 of 415

Attachment 1, Volume 9, Rev. 0, Page 292 of 415 RCS Operational LEAKAGE B 3.4.13 BASES SURVEILLANCE REQUIREMENTS (continued)

The Surveillance is modified by a Note which states that the Surveillance is not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after establishment of steady state operation. For RCS primary to secondary LEAKAGE 1% determination, steady state is defined as stable epower level[ pressurizer and makeug tank levels, fiakeupý land letc bwn, and RCP sealinjection and return Aows.

The Surveillance Frequency of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months is a reasonable interval to trend primary to secondary LEAKAGE and recognizes the importance of early leakage detection in the prevention of accidents. The primary to secondary LEAKAGE is determined using continuous process radiation monitors or radiochemical grab sampling in accordance with the EPRI guidelines (Ref. 5).

REFERENCES 1. 10 CFR 50, Appendix A, GDC 30.

2. Regulatory Guide 1.45, May 1973.

3. _ISA R, Pggiect'o 00©

4. NEI 97-06, "Steam Generator Program Guidelines."

5. EPRI, "Pressurized Water Reactor Primary-to-Secondary Leak Guidelines."

BWOG STS B 3.4.13-6 Rev. 3.1, 12/01/05 Attachment 1, Volume 9, Rev. 0, Page 292 of 415

Attachment 1, Volume 9, Rev. 0, Page 293 of 415 JUSTIFICATION FOR DEVIATIONS ITS 3.4.13 BASES, RCS OPERATIONAL LEAKAGE

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

2. Changes are made (additions, deletions, and/or changes) to the ISTS Bases which reflect the plant specific nomenclature, number, reference, system description, analysis, or licensing basis description.

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

4. Duplicate discussion deleted. Steady state operation is discussed in the previous paragraph.

5. Editorial change for clarity.

Davis-Besse Page 1 of 1 Attachment 1, Volume 9, Rev. 0, Page 293 of 415

Attachment 1, Volume 9, Rev. 0, Page 294 of 415 Specific No Significant Hazards Considerations (NSHCs)

Attachment 1, Volume 9, Rev. 0, Page 294 of 415

Attachment 1, Volume 9, Rev. 0, Page 295 of 415 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.4.13, RCS OPERATIONAL LEAKAGE There are no specific NSHC discussions for this Specification.

Davis-Besse Page 1 of 1 Attachment 1, Volume 9, Rev. 0, Page 295 of 415

Attachment 1, Volume 9, Rev. 0, Page 296 of 415 ATTACHMENT 14 ITS 3.4.14, RCS PRESSURE ISOLATION VALVE (PIV) LEAKAGE Attachment 1, Volume 9, Rev. 0, Page 296 of 415 .

, Volume 9, Rev. 0, Page 297 of 415 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) , Volume 9, Rev. 0, Page 297 of 415

Attachment 1, Volume 9, Rev. 0, Page 298 of 415 ITS 3.4.14 ITS REACTOR COOLANT SYSTEM OPERATIONAL LEAKAGE LIMITING CONDITION FOR OPERATION 3.4.6.2 Reactor oolant System operational leakage shall be limited to:

No PRESSURE BOUNDARY LEAKAGE,

b. I GPM UNIDENTIFIED LEAKAGE,

c. 150 gallons per day primary to secondary leakage through any one steam generator (SG),

See TS]

d. 10 GPM IDENTIFIED LEAKAGE from the Reactor Coolant System, C. 10 GPM CONTROLLED LEAKAGE. and LCO 3.4.14 f. ý eakage from any Reactor Coolant System Pressure Isolation Valve as par 1pecifle4 in Table .4-2. : e SR 3.4.14.2APPLICABILITY: MODES 1, 2, 3 and .rot

a. With any PRESSURE BOUNDARY LEAKAGE, or with primary to secondary See ITS leakage not within limit, be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in 3.4.13 COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

b. With any Reactor Coolant System operational leakage greater than any one of the ACTION A above limits, excluding PRESSURE BOUNDARY LEAKAGE or primary to secondary leakage, reduce the leakage rate to within limits within 44ours or e in-at ACTION B e taT STAND within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within Nthe following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months except as permitted by paragraph c below.

c. In the event that integrity of any pressure isolation valve specified in Table 3.4-2 Add ACTION A cannot be demonstrated, POWER OPERATION may continue, provided that at P proposed least two valves in each high pressure line having a non-functional valve are in and Required remain in, the mode corresponding to the isolated condition(a) AActions A.1 and A.2 Note SR 3.4.14.2 d. The provisions of Section 3.0.4 are not applicable for entry into MODES 3 and 4 for Note the purpose of testing the isolation valves in Table 3.4-2. Mot ACTION A --- ' Motor operated valves shall be placed in the closed position and power supplies deenergized.

DAVIS-BESSE, UNIT 1 3/44-15 Order dtd-4'208 I-Amendment No. l-35'80-2" 276 Page 1 of 8 Attachment 1, Volume 9, Rev. 0, Page 298 of 415

Attachment 1, Volume 9, Rev. 0, Page 299 of 415 ITS 3.4.14 ITS REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS 4.4.6.2.1 Reactor Coolant System operational leakages shall be demonstrated to be within each of the above limits by:

a. Monitoring the containment atmosphere gaseous or particulate radioactivity at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

b. Monitoring the containment sump level and flow indication at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

.See ITS 3.4 .13

c. Measurement of the CONTROLLED LEAKAGE from the reactor coolant pump seals to the makeup system when the Reactor Coolant System pressure is 2185 +/- 20 psig at least once per 31 days.

d. Performance of a Reactor Coolant System 1 water inventory balance at least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months during steady state operation.., )(2)

e. Verifying that primary to secondary leakage is 5 150 gallons per day through any one steam generator, at least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . (2)

SR 3.4.14.2 4.4.6.2.2 Each Reactor Coolant System Pressure Isolation Valvelspecified i.wTable 3.4-21shall be individually demonstrated OPERABLE by verifying leakage testing (or the equivalent) to be within its limit prior to entering MODE 2:

a. After each refueling outage,

b. Whenever the plant has been in COLD SHUTDOWN for 7 days, or more, and if leakage testing has not been performed in the previous 9 months, and

c. /Prior to returning the ve to service following maintenance, rpair or replacementi @

work on the valve. 7'-L003 SR 3.4.14.2 d. The provisions of Specification 4.0.4 are not applicable for entry into MODES 3 or 4.

Note 4.4.6.2.3 Whenever the intgrityofa pressure isoation val e liste in Tayle 3.4-- cannot be demonstrated, determine ard record the integrity of the high pressure fnb ath on a daily basis.

Integrity shall be determin d by performing either a leakage test of the r aining pressure isolation valve, or a combred leakage test of the remaining pressure isol tion valve in a series -L04 with the closed motor-opoted containment isolation valve. In addition, ecord the position of the closed motor-operat containment isolation valve located in the hi pressure piping on a daily basis. / t Not applicable to primary to secondary leakage. - See ITS (2)

Not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after establishment of steady state operation.- I , 3.413 J DAVIS-BESSE, UNIT I 3/4 4-16 -Order-dated-4A20/&l-Amendment No. -4 rl-96i-220, 276 Page 2 of 8 Attachment 1, Volume 9, Rev. 0, Page 299 of 415

Attachment 1, Volume 9, Rev. 0, Page 300 of 415 ITS 3.4.14 ITS TABLE 3.4-2 REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVES SR 3.4.14.2 Notes:

SR 3.4.14.2 1) 1. Leakage rates less than or equal to 1.0 gpm are considered acceptable.

2. Leakage rates greater than 1.0 gpm but less than or equal to 5.0 gpm are considered acceptable if the latest measured rate has not exceeded the rate deteralied by the previous test by an amount that reduces the margin between measured leakage rate and the maximum permissible rate of 5O gpm by SOS or greater.

3. Leakage rates greater than 1.0 gpm out less than or equal to 5.0 gpe are considered unacceptable if the latest measured rate exceeded the rate determined by the previou" test by an amount that reduces the margin between measured leakage rate and the maximum permissible rate of 5.0 gpm by 50% or greater.

4. Leakage rates greater than 5.0 gpm are considered unacceptable.

(b) Valves CF-30 nd CF-31 will be tested-with the Re ctor Coolant system presson 012O psig. Valves I]4-76 and DII- will be tested with norml ore Flooding Tank pressure which is '75 psig. Mini-mum than differe lal test pressure across each valve shall not be less 150 ps d.

(c) To satisfy requirements, leakage may be m asured indirectly -- O (as from performance of pressure indicators If accomplished In acco ce with approved procedures and sup rted by computations showi ng th t the method is capable of demnst lng valve compliance with the Ieaka criteria.

le I 4

DAVIS-BESSE. UNIT 1 3/4 -1 6 a Order dtd. 4/20/81 Page 3 of 8 Attachment 1, Volume 9, Rev. 0, Page 300 of 415

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Attachment 1, Volume 9, Rev. 0, Page 301 of 415

Attachment 1, Volume 9, Rev. 0, Page 302 of 415 ITS 3.4.14 ITS TABLE 3.3-3 (Continued)

ACTION STATEMENTS ACTION 12 - With the number of OPERABLE Units one less than the Total Number of Units, restore the inoperable functional unit to OPERABLE SeeITS]

status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in at least HOT STANDBY within the 3.3.6 next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

ACTION C ACTION 13 - a. With less than the Minimum Units OPERABLE and indicated reactor coolant pressure > 328 sig, both Decay Heat *LA03 Isolation Valves I(DHII Od DHl2)J`shall be verified closed. LS_

withinT4 hours

b. With Less than the Minimum Units OPERABLE and indicated reactor coolant pressure < 328 psig operation may continue; however, the functional unit shall be OPERABLE prior to increasing indicated reactor coolant pressure above 328 psig.

ACTION 14 - With less than the Minimum Units OPERABLE and indicated reactor coolant pressure < 328 psig, operation may continue; however, the functional unit shall be OPERABLE prior to increasing indicated See3IS]

reactor coolant pressure above 328 psig, or the inoperable functional unit shall be placed in the tripped state.

ACTION 15 - a. With the number of OPERABLE units one less thanthe Minimum Units Operable per Bus, place the inoperable unit in the tripped condition within one hour. For functional unit 4.a the sequencer shall be placed in the tripped condition by physical removal of the sequencer module. The inoperable See ITS 3.3.8 functional unit may be bypassed for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for and ITS 3.8.1 J surveillance testing per Specification 4.3.2.1.1'

b. With the number of OPERABLE units two less than the Minimum Units Operable per Bus, declare inoperable the Emergency Diesel Generator associated with the functional units not meeting the required minimum units OPERABLE and take the ACTION required of Specification 3.8.1.1.

DAVIS-BESSE, UNIT I 3/4 3-12a Amendment No. 28,52,1.2,136,211, 218 Page 5 of 8 Attachment 1, Volume 9, Rev. 0, Page 302 of 415

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Attachment 1, Volume 9, Rev. 0, Page 305 of 415 ITS 3.4.14 ITS Revised by RK Letter Date fMIRVEiIM 0.w P.rna

-A June 6, 1995

=VAl _" C~~iS L SUt AWL

b. At least once each REFUELING INTERVAL, or prior to operation after ECCS piping has been drained by verifying that the ECCS piping is fall of water by venting the ECCS pump casings and dischare piping high points.

C. Dy a visual inspection which verifies that no loose debris (ras, trash, clothing, etc.) Is present In the containment which colad be transported to the containment emergency sump and cause restriction of the puap suction during LOCA conditions. This See ITS 1 visual Inspection sealf be performed: 3.5.2 ]

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LCO 3.4.14 part 2

d. At least once each REFUELING INTERVAL by:

11. Verifying that the interlocls:

Se_ T I

a) Clseb~fl* and )-12L nlenrz the pressurize* LA0 SR 3.4.14.4 SR 3.4.4.4 - ea stMuatod eithercoolant system -pressmre.w*rich ifreactor Men 9 AIs J Igreater than the Allowable Value Oft I nLalied. . e interlock to closeis SR 3.4.14.3 Note, no required if the valve Is closed and 480 V AC power SR SR 3.4.14.

3.4.14.4 Note, is disconnected from its motor operators.

Note b ev h pn*o u ve b) Prevent the openin ofR I1 ~ I-26n SR 3.4.14.3 simulated or actual reactor oolant system pressure

]vhld is greater " than 4L*isp;Alie" "*

the Allowable Value (d1S psig)

Add proposed SR; 3.:4.14.5 -0

2. a) A visual inspection of the containment "mergem.cy st which verifiesthat the subsystem suctioninlets are not restricted by debris and that the smV oomonents.

(trash racks, sereen etc.) show no evidence of structural distress or corrosion.

b) Verifying that on a Borated Water Storage Tank (BWST)

Low-Low Level interlock trip, with the motor operators See ITS]

for the MWST outlet isolation valves and the containment emergency sucp recirculation valves energized, the BWST Outlet Valve HV-0117A (NV-OhT7) automatically close in 97S seconds aftir the operator manually pushes the control switch to open the Containment Emergency Sump Valve HV-DH9"(1NV-IM9B) which should be verified to open in g5 seconds.

.3. Deleted DAVIS-BESSE, UNIT I 3/4 5-4 Amendment No.

Page 8 of 8 Attachment 1, Volume 9, Rev. 0, Page 305 of 415

Attachment 1, Volume 9, Rev. 0, Page 306 of 415 DISCUSSION OF CHANGES ITS 3.4.14, RCS PRESSURE ISOLATION VALVE (PIV) LEAKAGE 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 3.4.6.2 Actions b and c specify the compensatory actions to take when the leakage through any RCS PIV(s) is greater than the specified limit. ITS 3.4.14 ACTIONS A and B also state the appropriate compensatory actions under the same condition; however, ITS 3.4.14 ACTIONS Note 1 has been added.

ITS 3.4.14 ACTIONS Note 1 allows separate Condition entry for each RCS PIV flow path. This changes the CTS by explicitly stating that the Actions are to be taken separately for each inoperable RCS PIV flow path.

The purpose of the Note is to provide explicit instructions for proper application of the Action for Technical Specification compliance. In conjunction with proposed Specification 1.3, "Completion Times," this Note provides direction consistent with the intent of the existing Action for inoperable PIVs. This change is designated as administrative because it does not result in technical changes to the CTS.

A03 CTS 3.4.6.2 Actions b and c specify the compensatory actions to take when the leakage through any RCS PIV(s) is greater than the specified limit. ITS 3.4.14 ACTIONS A and B also state the appropriate compensatory actions under the same condition; however, ITS 3.4.14 ACTIONS Note 2 has been added.

ITS 3.4.14 ACTIONS Note 2 states "Enter applicable Conditions and Required Actions for systems made inoperable by an inoperable RCS PIV." This changes the CTS by explicitly stating that the Conditions and Required Actions for systems made inoperable by an inoperable RCS PIV must be entered.

The purpose of the Note is to provide explicit instructions for proper application of the ACTION for Technical Specification compliance. This Note facilitates the use and understanding of the intent to consider any system affected by inoperable RCS PIVs, which is to have its ACTIONS also apply if it is determined to be inoperable. With the addition of ITS LCO 3.0.6, this intent would not be necessarily applied. This clarification is consistent with the intent and interpretation of the existing Technical Specifications, and is therefore considered an administrative presentation preference. This change is designated as administrative because it does not result in technical changes to the CTS.

A04 CTS Table 3.3-3 requires one channel of the decay heat isolation valve interlock to be OPERABLE. This channel is the channel common to the Safety Features Actuation System (SFAS) instrumentation, and it provides a interlock signal to one of the two isolation valves. The other channel that provides an interlock signal to the decay heat isolation valve is not common to SFAS instrumentation.

This channel is covered by CTS 4.5.2.d.1, which requires interlock testing for the Davis-Besse Page 1 of 7 Attachment 1, Volume 9, Rev. 0, Page 306 of 415

Attachment 1, Volume 9, Rev. 0, Page 307 of 415 DISCUSSION OF CHANGES ITS 3.4.14, RCS PRESSURE ISOLATION VALVE (PIV) LEAKAGE two decay heat isolation valves (DH-11 and DH-12). ITS 3.4.14 is combining these two requirements into a single LCO. ITS LCO 3.4.14 part 2 requires the Decay Heat Removal (DHR) System interlock function to be OPERABLE. This changes the CTS by combining the requirements for the interlock function into a single LCO.

This change is acceptable since the requirements are not being changed, except as justified in other Discussion of Changes. The requirements are simply being combined into a single LCO, consistent with NUREG-1430. This change is designated as administrative because it does not result in technical changes to the CTS.

MORE RESTRICTIVE CHANGES M01 CTS 3.4.6.2 Actions b and c specify the compensatory actions to take when the leakage through any RCS PIV(s) is greater than the specified limit. The compensatory action is to isolate the high pressure portion of the affected system from the low pressure portion of the affected system by use of a combination of at least two closed valves. The CTS does not include any leakage restrictions that may be used to satisfy the isolation requirement of this action. ITS 3.4.14 ACTION A is consistent with the requirement in CTS 3.4.6.2 Action c, however, a Note has been added to the Required Actions (ITS 3.4.14 Required Actions A.1 and A.2 Note) which specifies that each valve used to satisfy ITS 3.4.14 Required Actions A.1 and A.2 must have been verified to meet SR 3.4.14.2.a, the RCS PIV maximum leakage limit Surveillance Requirement, and either be in the RCS pressure boundary or the high pressure portion of the system. This changes the CTS by providing a Note which explicitly states that the valves used to satisfy Required Action must satisfy the same leakage requirements of the RCS PIVs and provides an option for them to be in the RCS pressure boundary.

The purpose of CTS 3.4.6.2 Action c is to isolate the flow path in order to minimize the leakage from the high pressure portion of the RCS to the low pressure piping. The ITS 3.4.14 Required Actions A.1 and A.2 Note requires the valves used to provide isolation between the high pressure and low pressure portions of the affected system to have been verified to meet the RCS PIV maximum leakage limits within the required Surveillance Frequency. The addition of the Note represents an additional restriction on unit operation necessary to help ensure the valves used to isolate the high pressure portion from the low pressure portion of the affected system are capable of preventing the overpressurization of the low pressure portion of the system. The ITS 3.4.14 Required Actions A.1 and A.2 Note also provides the option for the valves to be in the RCS pressure boundary. However, if it is in the RCS pressure boundary, it is in the high pressure portion of the system. This change is designated as more restrictive because it adds a new requirement to the CTS.

M02 The CTS does not require a CHANNEL CALIBRATION of the decay heat isolation valve interlock channel that is not common to SFAS instrumentation.

ITS SR 3.4.14.5 requires a CHANNEL CALIBRATION every 24 months. This changes the CTS by adding a specific CHANNEL CALIBRATION requirement for this channel.

Davis-Besse Page 2 of 7 Attachment 1, Volume 9, Rev. 0, Page 307 of 415

Attachment 1, Volume 9, Rev. 0, Page 308 of 415 DISCUSSION OF CHANGES ITS 3.4.14, RCS PRESSURE ISOLATION VALVE (PIV) LEAKAGE The purpose of the CHANNEL CALIBRATION is to ensure the channel can perform as required. Currently, the CTS only requires a functional test of the channel (CTS 4.5.2.d.1). The addition of the CHANNEL CALIBRATION requirement will help ensure the accuracy of the instrument string, therefore the change is acceptable. The proposed 24 month Frequency is consistent with the CHANNEL CALIBRATION Frequency for the other channel (the channel common to SFAS instrumentation) and with the Frequency of CTS 4.5.2.d.1.

This change is designated as more restrictive because it adds a new 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 3.4.6.2.f requires the leakage from each RCS PIV specified in Table 3.4-2 to be < 5 gpm. CTS 4.4.6.2.2, the Surveillance which checks the RCS PIV leakage, also references Table 3.4-2. CTS Table 3.4-2 contains a list of the RCS PIVs and their associated valve numbers. ITS 3.4.14 does not contain a list of the RCS PIVs or their associated valve numbers. This changes the CTS by relocating the list of RCS PIVs and their associated valve numbers 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. ITS 3.4.14 still requires the RCS PIVs to be OPERABLE, and ITS SR 3.4.14.2 requires periodic Surveillances to determine RCS PIV leakage. It is not necessary for the list of RCS PIVs to be in the Technical Specifications in order to ensure that the RCS PIVs are OPERABLE. Other lists of components, such as containment isolation valves and equipment response time, have been relocated from the Technical Specification to licensee-controlled documents while retaining the requirements on these components in Technical Specifications. Also, this change is acceptable because these types of procedural details will be adequately controlled in the ITS Bases. Changes to the Bases are controlled by the Technical Specification Bases Control Program in Chapter 5. This program provides for the evaluation of changes to ensure the Bases are properly controlled. This change is designated as a less restrictive removal of detail change because information relating to system design is being removed from the Technical Specifications.

LA02 (Type 3 - Removing ProceduralDetails for Meeting TS Requirements or Reporting Requirements) CTS Table 3.4-2 is modified by Notes (b) and (c).

Note (b) describes the pressure at which the RCS PIVs are to be tested. Note (c) explains an alternative method of testing the PIVs to satisfy the ALARA requirements. ITS 3.4.14 does not retain these Notes. This changes the CTS by relocating the information in the Notes to the Bases.

Davis-Besse Page 3 of 7 Attachment 1, Volume 9, Rev. 0, Page 308 of 415

Attachment 1, Volume 9, Rev. 0, Page 309 of 415 DISCUSSION OF CHANGES ITS 3.4.14, RCS PRESSURE ISOLATION VALVE (PIV) LEAKAGE The removal of these details for performing Surveillance Requirements from the Technical Specification is acceptable because this type of information is not necessary to be included in the Technical Specifications to provide adequate protection of public health and safety. ITS 3.4.14 still retains the requirements that RCS PIV leakage must be within limit and provides the appropriate Surveillance that includes the leakage limit. Also, this change is acceptable because these types of procedural details will be adequately controlled in the ITS Bases. Changes to the Bases are controlled by the Technical Specification Bases Control Program in Chapter 5. This program provides for the evaluation of changes to ensure the Bases are properly controlled. This change is designated as a less restrictive removal of detail change because procedural details for meeting Technical Specification requirements are being removed from the Technical Specifications.

LA03 (Type 1 - Removing Details of System Design and System Description,Including Design Limits) CTS Table 3.3-3 Action 13 and CTS 4.5.2.d.1 provide the specific valve numbers for the decay heat removal isolation valves. CTS Table 3.3-4 footnote

states that the Decay Heat Removal System interlock function Allowable Value is referenced to the RCS pressure instrumentation tap.

ITS 3.4.14 does not include these details. This changes the CTS by moving the valve numbers and information concerning the Allowable Value reference point 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 the Decay Heat Removal System interlock function to be OPERABLE, and provides Surveillances to ensure the interlock operates at the proper setpoint.

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 (Category2 - Relaxation of Applicability) CTS 3.4.6.2.f is applicable in MODES 1, 2, 3, and 4. ITS 3.4.14 is applicable in MODES 1,2, and 3, and in MODE 4, except valves in the decay heat removal (DHR) flow path when in, or the transition to or from, the DHR mode of operation. This changes the CTS by exempting the DHR flow path PIVs (CF-30, CF-31, DH-76, and DH-77) from the leakage requirements when in or during the transition to or from the DHR mode of operation.

The purpose of CTS 3.4.6.2.f is to ensure the RCS PIVs are within leakage limits.

This change is acceptable because the LCO requirements continue to ensure that the components are maintained consistent with the safety analyses and Davis-Besse Page 4 of 7 Attachment 1, Volume 9, Rev. 0, Page 309 of 415

Attachment 1, Volume 9, Rev. 0, Page 310 of 415 DISCUSSION OF CHANGES ITS 3.4.14, RCS PRESSURE ISOLATION VALVE (PIV) LEAKAGE licensing basis. It is not necessary for the DHR PIVs to meet the leakage limits when in or during transition to or from the DHR mode of operation. These check valves cannot open until the DHR System is placed in service, which is not until RCS pressure is less than the test pressure of the DHR system. Thus overpressurization of the DHR piping is not a concern. This change is designated as less restrictive because less stringent LCO requirements are being applied in the ITS than are being applied in the CTS.

L02 (Category 3 - Relaxation of Completion Time) CTS 3.4.6.2 Action b requires, in part, that if the RCS PIV leakage is not within limit, it must be restored within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . If RCS PIV leakage is not restored, either a unit shutdown is required or the requirements of CTS 3.4.6.2 Action c must be met. CTS 3.4.6.2 Action c states, in part, that with the integrity of any pressure isolation valve specified in Table 3.4-2 not demonstrated, power operation may continue provided at least two valves in each high pressure line that has a non-functional valve are in and remain in, the mode corresponding to the isolated condition. Therefore, the two CTS Actions result in requiring the two valves to be in the isolated condition within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . ITS 3.4.14 ACTION A contains this same requirements, but allows 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months to isolate the first valve and 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months to isolate the second valve.

This changes the CTS by extending the time requirement to close the second valve from 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

The purpose of CTS 3.4.6.2 Actions b and c is to allow time to reduce leakage before isolating the pathway. This change is acceptable because the Completion Time is consistent with safe operation under the specified Condition, considering the OPERABLE status of the redundant systems or features. This includes the capacity and capability of remaining systems or features, a reasonable time for repairs or replacement, and the low probability of a DBA occurring during the allowed Completion Time. The time to close the first valve remains the same and the time to close the second valve has been changed from 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . The 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months Completion Time to close the first valve ensures leakage in excess of the allowable limit is reduced. The 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months time allows time for these actions and restricts the time of operation with leaking valves. The 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Completion Time to close the second valve considers the time required to complete the Required Action and the low probability of the first valve failing during this period. This change is designated as less restrictive because additional time is allowed to restore parameters to within the LCO limits than was allowed in the CTS.

L03 (Category5 - Deletion of Surveillance Requirement) CTS 4.4.6.2.2.c requires testing of RCS PIVs following maintenance, repair, or replacement work on the valve. ITS 3.4.14 does not include this requirement. This changes the CTS by eliminating a post-maintenance Surveillance Requirement.

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 safety function. Whenever, the OPERABILITY of a system or component has been affected by repair, maintenance, modification, or replacement of a component, post maintenance testing is required to Davis-Besse Page 5 of 7 Attachment 1, Volume 9, Rev. 0, Page 310 of 415

Attachment 1, Volume 9, Rev. 0, Page 311 of 415 DISCUSSION OF CHANGES ITS 3.4.14, RCS PRESSURE ISOLATION VALVE (PIV) LEAKAGE demonstrate the OPERABILITY of a system or component. This is described in the Bases for ITS SR 3.0.1 and required under SR 3.0.1. In addition, the requirements of 10 CFR 50, Appendix B, Section XI (Test Control), provide adequate controls for test programs to ensure that testing incorporates applicable acceptance criteria. Compliance with 10 CFR 50, Appendix B is required under the unit operating license. As a result, post-maintenance testing will continue to be performed and an explicit requirement in the Technical Specifications is not necessary. This change is designated as less restrictive because Surveillances which are required in the CTS will not be required in the ITS.

L04 (Category 5 - Deletion of Surveillance Requirement) CTS 4.4.6.2.3 provides additional compensatory measures to take, above those required by CTS 3.6.4.2 Action c, when leakage through an RCS PIV is not within limit. The CTS requires a daily leakage test of the remaining OPERABLE RCS PIV in the flow path or a combined leakage test of the two valves used to comply with CTS 3.6.4.2 Action c. In addition, the position of the second, non-RCS PIV valve is required to be recorded on a daily basis. ITS 3.4.14 does not include these additional compensatory measures. This changes the CTS by deleting the additional compensatory measures taken when leakage through an RCS PIV is not within limit.

The purpose of CTS 4.4.6.2.3 is to help ensure that the leakage through the valves used to isolate the penetration with an inoperable RCS PIV is minimized so that an overpressurization event of the downstream piping cannot occur. The change is acceptable since the requirements to ensure the leakage through the two closed valves is within the RCS PIV leakage limit and to ensure closure of the valves are maintained in the ITS. The RCS PIV leakage is ensured prior to using each of the valves as an isolation boundary, as required by the ITS 3.4.14 Required Actions Note. Once leakage is checked, it is not expected to change since the valve cannot be manipulated (ITS 3.4.14 ACTION A requires the valves to be isolated - thus they must remain isolated to comply with the ACTION).

Manipulation of manual valves that have been closed and automatic valves that have de-activated to comply with Technical Specification Actions is a controlled evolution and the valves are not expected to be inadvertently moved from the isolated condition. Furthermore, these valves will be verified to be in the correct position when first isolated to comply with ITS 3.4.14 ACTION A. This change is designated as less restrictive because a Surveillance required by the CTS will not be required in the ITS.

L05 (Category 3 - Relaxation of Completion Time) CTS Table 3.3-3 Action 13.a states, in part, that with the decay heat isolation valve interlock channel inoperable, both Decay Heat Removal Isolation Valves shall be verified closed.

While no specific time is provided, the term "verified closed" implies this is an immediate action. ITS 3.4.14 ACTION C states, in part, that with the Decay Heat Removal (DHR) System interlock function inoperable, isolate the affected penetration by use of two closed deactivated automatic valves within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

This changes the CTS by allowing 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months to complete the Required Action instead of the current immediate time.

The purpose of CTS 3.3-3 Action 13.a is to isolate the DHR isolation valves ifthe DHR valve interlock is inoperable. This change is acceptable because the Davis-Besse Page 6 of 7 Attachment 1, Volume 9, Rev. 0, Page 311 of 415

Attachment 1, Volume 9, Rev. 0, Page 312 of 415 DISCUSSION OF CHANGES ITS 3.4.14, RCS PRESSURE ISOLATION VALVE (PIV) LEAKAGE Completion Time is consistent with safe operation under the specified Condition, considering the operability status of the redundant systems of required features, the capacity and capability of remaining features, a reasonable time for repairs or replacement of required features, and the low probability of a overpressurization event occurring during the allowed Completion Time. The four hour Completion Time will provide the operator sufficient time to reposition the valves. This change is designated as less restrictive because the Completion Time specified in CTS has been extended in the ITS.

Davis-Besse Page 7 of 7 Attachment 1, Volume 9, Rev. 0, Page 312 of 415

Attachment 1, Volume 9, Rev. 0, Page 313 of 415 Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs)

Attachment 1, Volume 9, Rev. 0, Page 313 of 415

Attachment 1, Volume 9, Rev. 0, Page 314 of 415 CTS RCS PIV Leakage 3.4.14 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.14 RCS Pressure Isolation Valve (PMV) Leakage 3.4.6.2.f LCO 3.4.14 Leakage from each RCS PIV shall be within limits.

0 APPLICABILITY: MODES 1, 2, and 3, MODE 4, except valves in the decay heat removal (DHR) flow path when in, or during the transition to or from, the DHR mode of operation.

ACTIONS DOC A02 1. Separate Condition entry is allowed for each flow path.

DOC A03 2. Enter applicable Conditions and Required Actions for systems made inoperable by an inoperable PIV.

CONDITION REQUIRED ACTION COMPLETION TIME Actions b A. One or more flow paths -------------- NOTE ----------

and c with leakage from one or Each valve used to satisfy Required more RCS PIVs not Action A.1 and Required Action A.2 within limit, must have been verified to meet 03 05 2a SR 3.4.14] and be on the RCS pressure boundarypr the high 0 pressure portion ofthe systerm A.1 Isolate the high pressure 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months portion of the affected system from the low pressure portion by use of one closed manual, deactivated automatic, or check valve.

AND BWOG STS 3.4.14-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 0, Page 314 of 415

Attachment 1, Volume 9, Rev. 0, Page 315 of 415 3.4.14 CTS (DINSERT 1 AND Table 3.3-3 The Decay Heat Removal (DHR) System interlock function shall be OPERABLE.

Functional Unit 5.a, 4.5.2.d.1 Insert Page 3.4.14-1 Attachment 1, Volume 9, Rev. 0, Page 315 of 415

Attachment 1, Volume 9, Rev. 0, Page 316 of 415 CTS RCS PIV Leakage 3.4.14 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME A.2 Jlsolate the high pressure portion of the affected 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months 0 system from the low pressure portion .by use of a second closed manual, deactivated automatic, or check valve.

[or] /

Restore RC PIV to within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months limits.

Action b B. Required Action and associated Completion B.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months 00 Time for Condition A not AND met.

B.2 Be in MODE 5. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months 0 Table 3.3-3 C. M]Decay Heat Removal C.1 Isolate the affected Action 13 XDHR) System ainterlock function inoperable.

eetion line* closed man by use of [o or deactivated automatic 0@

valve. -------E a

BWOG STS 3.4.14-2 Rev. 3.0, 03/31104 Attachment 1, Volume 9, Rev. 0, Page 316 of 415

Attachment 1, Volume 9, Rev. 0, Page 317 of 415 3.4.14 CTS 0INSERT 2 Table 3.3-3 Action 13 OR C.2 -------- NOTE------

Only applicable if RCS pressure < 328 psig.

Restore the interlock Prior to increasing function to OPERABLE RCS pressure status. > 328 psig Insert Page 3.4.14-2 Attachment 1, Volume 9, Rev. 0, Page 317 of 415

Attachment 1, Volume 9, Rev. 0, Page 318 of 415 CTS RCS PIV Leakage 3-4.14 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY Action c footnote (a),

SR 3_t4.4 Ol

[ --................. NOT% -----

required to be performed in MO[DE 0

4.4.6.2.2.d

2. Not requirel to be performed on Ihe RCS PIVsI located in t~e DHR flow path wh n in the DHR mode of o&eration. /A 0

3. RCS PIVs atuated during the pe/rformance of this Surveil=nce are not require4 to be tested more than nce if a repetitive te ting loop cannot be /voided. /

4.4.6.2.2.a, tVeriileakage from each RCS PIV is equivalent to In accor nce 4.4.6.2.2.b. ~gpm~por nominal inch of valve/size up to aI with the nservice Table 3.4-2 Imaximum of5gpmlatan RCS pressure Testing rogram 21 ps [and<[2 5llpsia. or [18] months AND Prior to entering 0 MODE 2 INSERT 4 whenever the unit has been in 0

MODE 5 for 7 days or more, if leakage testing has not been performed in the previous 9 months AND /

[ Within hours following valve actuatio due to automatic or 0

manual otion or flow thr ugh the valve]

BWOG STS 3.4.14-3 Rev. 3.0, 03/31104 Attachment 1, Volume 9, Rev. 0, Page 318 of 415

Attachment 1, Volume 9, Rev. 0, Page 319 of 415 3.4.14 CTS (DINSERT 3 Table 4.3-2 SR 3.4.14.1 Perform CHANNEL CHECK on the DHR 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Functional Unit 5.a System interlock channel common to Safety Features Actuation System (SFAS) instrumentation.

(DINSERT 4 Table 3.4-2 b. When current measured rate is > lgpm, the current measured rate has Note (a) not exceeded the rate determined by the previous test by an amount that reduces the margin between measured leakage rate and 5.0 gpm by 50%.

Insert Page 3.4.14-3 Attachment 1, Volume 9, Rev. 0, Page 319 of 415

Attachment 1, Volume 9, Rev. 0, Page 320 of 415 CTS RCS PIV Leakage 34.14 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY Table 4.3-2 SR 3.4.14.[ ..................... .NO TE -

Functional Unit 5.a, MNot required to be met when the DHR System 0auclour~e interlocl*s disabled in accordance with 00 4.5.2.d.1 .b) LCO 3.4.12. L f--Lunrtion__

Verify DHR System oc sure interlock revents F1_*monthsM the valves from being opened with a simulated or actual RCS pressure signal _> ý328 Table 4.3-2 SR 3.4.14 Functional MNot required to be met when the DHR System Unit 5.a, lautc>u interloc*s disabled in accordance with 4.5.2.d.1 .a) LCO 3.4.12. funtio 0

Verify DHR System utocsure interloclcauses the F11months R valves to close automatically with a simulated or actual RCS pressure signal BWOG STS 3.4.14-4 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 0, Page 320 of 415

Attachment 1, Volume 9, Rev. 0, Page 321 of 415 3.4.14 CTS (INSERT 5 Table 4.3-2 SR 3.4.14.5 Perform CHANNEL CALIBRATION on the 24 months Functional DHR System interlock channels.

Unit 5.a Insert Page 3.4.14-4 Attachment 1, Volume 9, Rev. 0, Page 321 of 415

Attachment 1, Volume 9, Rev. 0, Page 322 of 415 JUSTIFICATION FOR DEVIATIONS ITS 3.4.14, RCS PRESSURE ISOLATION VALVE (PIV) LEAKAGE

1. The second part of the LCO has been added to ensure consistency between the LCO, ACTIONS, and Surveillance Requirements. The ISTS LCO, ACTIONS, and Surveillances do not match up since there is no explicit statement in the LCO requiring the DHR System interlock function to be OPERABLE. LCO 3.0.1 requires LCOs to be met during the MODES or other specified conditions in the Applicability.

LCO 3.0.2 states that upon discovery of a failure to meet an LCO, the Required Actions of the associated Conditions shall be met. Currently, if the DHR System interlock function is inoperable, the LCO is still met. Thus, ACTION C is not required to be entered since the LCO is still met. Therefore, the inclusion of the second portion of the LCO ensures consistency between the LCO, ACTIONS, and Surveillance Requirements. In addition, due to the addition of the term "DHR" into the LCO statement, the use of the term "decay heat removal (DHR)" in the Applicability has been changed to "DHR."

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

3. ISTS 3.4.14 has been modified to reflect the Davis-Besse current licensing basis requirements for the DHR System interlock function, with the exception of the Completion Time provided in ISTS 3.4.14 Required Action C.1. The Current Technical Specifications (CTS) requires the DHR System line to be isolated by closing both of the automatic valves in the flow path. This is reflected in ITS 3.4.14 Required Action C.1. The CTS also allows the interlock function to be inoperable with RCS pressure below 328 psig provided the interlock function is restored to OPERABLE status prior to increasing RCS pressure to > 328 psig. This is reflected in ITS 3.4.14 Required Action C.2. In addition, both a 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months CHANNEL CHECK and a 24 month CHANNEL CALIBRATION are required by the CTS. These Surveillances are reflected in ITS SR 3.4.14.1 and ITS SR 3.4.14.5. Due to the addition of these Surveillances, the remaining Surveillances have been renumbered.

4. Editorial changes have been made to be consistent with the Writers Guide for the Improved Standard Technical Specifications, TSTF-GG-05-01, Section 4.1.7.g.

5. The Davis-Besse RCS PIV leakage limits have been provided, consistent with current licensing basis. In addition, since ITS SR 3.4.14.2 includes two limits, only the first limit (a maximum limit) is applicable for the Required Actions A.1 and A.2 Note.

6. Note 2 to ISTS SR 3.4.14.1 has been deleted since it is not necessary. The ISTS 3.4.14 Applicability does not require leakage to be met for DHR valves in the flow path when in MODE 4 and when in, or during the transition to or from, the DHR mode of operation.

7. The third Frequency of ISTS SR 3.4.14.1 has been deleted since it is not required by the current licensing basis. The first two Frequencies are adequate to ensure the RCS PIV leakage is within the limit. In addition, due to this deletion, Note 3 has also been deleted.

8. Due to the deletion of ISTS SR 3.4.14.1 Notes 2 and 3, the remaining Note has not been numbered and the word "NOTES" has been changed to "NOTE."

Davis-Besse Page 1 of 1 Attachment 1, Volume 9, Rev. 0, Page 322 of 415

Attachment 1, Volume 9, Rev. 0, Page 323 of 415 Improved Standard Technical Specifications (ISTS) Bases Markup and Justification for Deviations (JFDs)

Attachment 1, Volume 9, Rev. 0, Page 323 of 415

Attachment 1, Volume 9, Rev. 0, Page 324 of 415 RCS PIV'Leakage B 3.4.14 B 3.4 REACTOR COOLANT SYSTEM (RCS)

B 3.4.14 RCS Pressure Isolation Valve (PLY) Leakage BASES BACKGROUND 10 CFR 50.2, 1:0 CFR 50.55a(c), and GDC 55 of 10 CFR 50, Appendix A discuss reactor coolant (Refs. 1, 2, and 3 define RCS P s as an two normally closed valves in pressure boundary series within the ý ressure boundary that separate the high pressure K valves, which are , RCS from an attached low pressure system. During their lives, these coolanI"-_

valves can produce varying amounts of reactor coolant leakage through INSERT 1 either normal operational wear or mechanical deterioration. The RCS PIV Leakage LCO allows RCS high pressure operation when leakage through these valves exists in amounts that do not compromise safety.

The PIV leakage limit applies to each individual valve. Leakage through both series PIVs in a line must be included as part of the identified LEAKAGE, governed by LCO 3.4.13, "RCS Operational LEAKAGE." This is true during operation only when the loss of RCS mass through two series valves is determined by a water inventory balance (SR 3.4.13.1).

A known component of the identified LEAKAGE before operation begins is the least of the two individual leakage rates determined for leaking series PIVs during the required surveillance testing; leakage measured through one PIV in a line is not RCS operational LEAKAGE if the other is leaktight.

Although this specification provides a limit on allowable PIV leakage rate, its main purpose is to prevent overpressure failure of the low pressure portions of connecting systems. The leakage limit is an indication that the PIVs between the RCS and the connecting systems are degraded or degrading. PIV leakage could lead to overpressure of the low pressure piping or components. Failure consequences could be a loss of coolant accident (LOCA) outside of containment, an unanalyzed accident that could degrade the ability for low pressure injection.

The basis for this LC is the 1975 NRC "Reactor Safe Study" (Ref. 4) that identified poten ial intersystem LOCAs as a signif nt contributor to the risk of core m A subsequent s dy (Ref. 5) evaluated various PI configurations to determine the robability of intersystem LOCAs.

PIVs are provided to isolate the RCS from the lfollowinypicallyl connectesyems:

[ Decay Heat Removal (DHR) System.

BWOG STS B 3.4.14-1 Rev. 3.1, 12/01/05 Attachment 1, Volume 9, Rev. 0, Page 324 of 415

Attachment 1, Volume 9, Rev. 0, Page 325 of 415 B 3.4.14 (O INSERT 1 The 1975 Reactor Safety Study, WASH-1400, (Ref. 4) identified intersystem loss of coolant accidents (LOCAs) as a significant contributor to the risk of core melt. The study considered designs containing two in-series check valves and two check valves in series with a motor operated valve that isolated the high pressure RCS from the low pressure safety injection system. The scenario considered is a failure of the two check valves leading to overpressurization and rupture of the low pressure injection piping which results in a LOCA that bypasses containment. A letter was issued (Ref. 5) by the NRC requiring plants to describe the PIV configuration of the plant. On April 20, 1981, the NRC issued an Order modifying the Davis-Besse Technical Specifications to include testing requirements on PIVs and to specify the PIVs to be tested (Ref. 6).

Insert Page B 3.4.14-1 Attachment 1, Volume 9, Rev. 0, Page 325 of 415

Attachment 1, Volume 9, Rev. 0, Page 326 of 415 RCS PIV Leakage B 3.4.14 BASES BACKGROUND (continued)

Ia. Decay Heat Remlal (DHR) System, 1b. Emergency Core Coig System (ECCS), and0 Ic. Makeup and iPtf*fication System. CF-30, CF-31, OH-76, and DR-77 The PIVs are Ilisted in [FSAR op-trb-n]

Reference 6 o Violation of this LCO could result in continued degradation of a PIV, which could lead to overpressurization of a low pressure system and the loss of the integrity of a fission product barrier APPLICABLE Reference 4 identified potential intersystem LOCAs as a significant SAFETY contributor to the risk of core melt. The dominant accident sequence in ANALYSES the intersystem LOCA category is the failure of the low pressure portion of the DHR System outside of containment. The accident is the result of a postulated failure of the PIVs, which are part of the reactor coolant pressure boundary (RCPB), and the subsequent pressurization of the DHR System downstream of the PIVs from the RCS. Becauselte ow lpressure gortion ot the DHR System isRTypi7'W designed or sig, to handle normal RCS pressures. overpressurization failure of the DHR low pressure line would result ina LOCA outside containment and subsequent risk of core melt.

Reference 5 evaluatr various PIV configurations, leaka e testing of the valves, and operati al changes to determine the effe on the probability of intersystem LO/-0As. This study concluded that perodic leakage testing of the PIVs can ubstantially reduce the probability an intersystem LOCA.

RCS PIV leakage satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii).

LCO RCS PIV leakage is identified LEAKAGE into closed systems connected to the RCS. Isolation valve leakage is usually on the order of drops per minute. Leakage that increases significantly suggests that something is operationally wrong and corrective action must be taken.

The LCO PIV leakage limit is"0.5 gpm per nominal inch of/valve size with a maximum limit of ,5 gpm. The previous criterion of 1 gpn for all valve sizes imposed i n unjustified penalty on the larger valveswithout INET2providing infor ation on potential valve degradation an resulted in higher person el radiation exposures. A study conclud d a leakage rate Simit based on/valve size was superior to a sRngle allo 3ble va0ue.

BWOG STS B 3.4.14-2 Rev. 3.1, 12/01/05 Attachment 1, Volume 9, Rev. 0, Page 326 of 415

Attachment 1, Volume 9, Rev. 0, Page 327 of 415 B 3.4.14 INSERT JA Two motor operated valves (which are not PIVs) are included in series in the suction piping of the DHR System to isolate the high pressure RCS from the low pressure piping of the DHR System when the RCS pressure is above the design pressure of the DHR System piping and components. Ensuring the DHR System interlock function that closes the valves and prevents the valves from being opened is OPERABLE ensures that RCS pressure will not pressurize the DHR System beyond its test pressure.

O INSERT 2

< 5.0 gpm. However, when the current measured rate is > 1.0 gpm, the current measured rate shall not exceed the rate determined by the previous test by an amount that reduces the margin between measured leakage rate and the maximum permissible rate (5.0 gpm) by 50%.

Insert Page B 3.4.14-2 Attachment 1, Volume 9, Rev. 0, Page 327 of 415

Attachment 1, Volume 9, Rev. 0, Page 328 of 415 RCS PIV Leakage B 3.4.14 BASES LCO (continued)

Reference 7 permits leakage testing at a lower pressure differential than between the specified maximum RCS pressure and the normal pressure of the connected system during RCS operation (the maximum pressure differential) in those types of valves in which the .higher service pressure will tend to diminish the overall leakage channel opening. In such cases, the observed rate may be adjusted to the maximum pressure differential by assuming leakage is directly proportional to the pressure differential to Ie one half power. 0 APPLICABILITY In MODES 1, 2, 3, and 4, this LCO applies because the PIV leakage potential is greatest when the RCS is pressurized. In MODE 4, valves in the DHR flow path are not required to meet the requirements of this LCO when in, or during the transition to or from, the DHR mode of operation.

In MODES 5 and 6, leakage limits are not provided because the lower reactor coolant pressure results in a reduced potential for leakage and for a LOCA outside the containment.

ACTIONS The ACTIONS are modified by two Notes. Note 1 is added to provide clarification that each flow path allows separate entry into a Condition.

This is allowed based upon the functional independence of the flow path.

Note 2 requires an evaluation of affected systems if a PIV is inoperable.

The leakage may have affected system operability, or isolation of a leaking flow path with an alternate valve may have degraded the ability of the interconnected system to perform its safety function.

A.1 and A.2 0

If the leakage from one or more RCS PIVs is -*T -e flow path must be isolated by two valves. Required Actions A.1 not within limit, the and A.2 are modified by a Note that the valves used for isolation must meet the same leakage requirements as the PIVs and must be on the RCS pressure boundary[r the high pressure portion of the systerrm Required Action A.1 requires that the isolation with one valve must be performed within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . Four hours provides time to reduce leakage in excess of the allowable limit and to isolate the affected system if leakage cannot be reduced. The 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months allows the actions and restricts the operation with leaking isolation valves.

BWOG STS B 3.4.14-3 Rev. 3.1, 12/01/05 Attachment 1, Volume 9, Rev. 0, Page 328 of 415

Attachment 1, Volume 9, Rev. 0, Page 329 of 415 B 3.4.14 0 INSERT 3 Ensuring the DHR System interlock function that closes the valves and prevents the valves from being opened is OPERABLE ensures that RCS pressure will not pressurize the DHR System beyond its test pressure.

Insert Page B 3.4.14-3 Attachment 1, Volume 9, Rev. 0, Page 329 of 415

Attachment 1, Volume 9, Rev. 0, Page 330 of 415 RCS PIV Leakage B 3.4.14 BASES ACTIONS (continued)

MRequired Action A.2 specifies that the double isolation .barrier of two valves be restored by closing some other valve qualified for 'isolation or 0

restoring one leaking PIV. -he 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months time after exceeding the limit considers the time require to complete the Action and the low probability of a second valve failing during this time period.

0 or The 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months time a er exceeding the limit allows for t e restoration of the leaking PIV to OP ABLE status. This timeframe co siders the time required to comple e this Action and the low probabily of a second valve failing during this eriod. ]

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

Two options are proided for Required Action A.2. Th* second option (72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months restorationh is appropriate if isolation of a se ond valve would place the unit in an nanalyzed condition.

0 B.1 and B.2 If leakage cannot be reduced, *he system isolateda or other Required Actions accomplished, the plan must be brought to a MODE in which the 0

requirement does not apply. To achieve this status, the plant must be brought to MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and to MODE 5 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . This Required Action may reduce the leakage and also reduces the potential for a LOCA outside the containment. The allowed Completion Times are reasonable based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

C. i funcion The inoperability of the DHR Fautoc sure interlocVrenders the DHR suction isolation valves incapable of isolating in response to a high pressure condition and preventing inadvertent opening of the valves at RCS pressures in excess of the DHR systems design pressure. If the DHR autoc sure interlock is inoperable, operation may continue as long :3 as the DH R suctiocee~aln is ' byo e a atne closed 5-*

tdis --

cfrldeactivated a-utomnatfic valv !h'M obu~rs.'tThis action accomplishes the purpose of the auto sure fn t won' BWOG STS B 3.4.14-4 Rev. 3.1, 12/01/05 Attachment 1, Volume 9, Rev. 0, Page 330 of 415

Attachment 1, Volume 9, Rev. 0, Page 331 of 415 B 3.4.14 0* INSERT 4 Alternately, if the RCS pressure is < 328 psig, isolating the associated DHR penetration is not required. In this case, the DHR System interlock function must be restored to OPERABLE status prior to increasing RCS pressure > 328 psig. Since RCS pressure is below the setpoint, there is no need to isolate the associated penetration.

Insert Page B 3.4.14-4 Attachment.1, Volume 9, Rev. 0, Page 331 of 415

Attachment 1, Volume 9, Rev. 0, Page 332 of 415 RCS PIV Leakage B 3.4.14 BASES SUR VEILLANCE REC *UIREMENTS SRT 3.4.14.7N 0 2

Performance of leakage testing on each RCS PIV or-isolation valve used to satisfy Required Action A.1 or A.2 is required to verify that leakage is

....... 6 below the specified limit and to identify each leaking valve- FT-he leakag imit of 0.5 gpm pq/r inch of no mi~nalvalve diameter uA) to 5 gpm m ai l__ u 0 J-apphies to each volve.1 Leakage testing requires :a stable pressure condition.,

0 I E 7 For the two PIVs in series, the leakage requirement applies to each valve individually and not to the combined leakage across both valves. Ifthe PIVs are not individually leakage tested, one valve may have failed completely and not detected if the other valve in series meets the leakage requirement. In this situation, the protection provided by redundant valves would be lost.

Testing is to be performed every[I-j] months, a typical refueling cycle, if 24 the plant does not go into MODE 5 for at least 7 days. The Jinontl t

Frequency is consistent with 10 CFR 50.55a(g) (Ref. 8) as contained in the Inservice Testing Program, is within frequency allowed by the American Society of Mechanical Engineers (ASME) Code (Ref. 7), and is based on the need to perform such surveillances under conditions that 2 apply during an outage and the potential for an unplanned transient if the Surveillance were performed with the plant at power.

[In addition, testing nust be performed once after the /alve has been opened by flow or .ercised to ensure tight reseatin . PIVs disturbed in the performance ol this Surveillance should also be :sted unless documentation sh.ws that an infinite testing loop ca not practically be avoided. Testing fnust be performed within 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , after the valve has been reseated. 1V ithin 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months is a reasonable an practical time limit for performing thiý test after opening or reseating a valve. ]

0 I eromed The leakage limit is to be rrrtat the ROS pressure associated with MODES 1 and 2. This permits leakage testing at high differential pressures with stable conditions not possible in the MODES with lower pressures.t

/ Entry into MODES 3 and 4 is allowed to establish the necessary differential pressures and stable conditions to allow for performance of this Surveillance. The Note that allows this provision is complimentary to the Frequency of prior to entry into MODE 2 whenever the unit has been BWOG STS B 3.4.14-5 Rev. 3.1, 12/01/05 Attachment 1, Volume 9, Rev. 0, Page 332 of 415

Attachment 1, Volume 9, Rev. 0, Page 333 of 415 B 3.4.14 0 INSERT 5

SR 3.4.14.1 SR 3.4.14.1 is the performance of the CHANNEL CHECK of the decay heat isolation valve interlock channel that is common to the Safety Features Actuation System (SFAS) instrumentation. The check provides reasonable confidence that the channel is operating properly. The Frequency of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is based on instrument reliability and is reasonable for detecting off normal conditions.

O INSERT 6 The RCS PIV leakage limit is < 5.0 gpm. However, RCS PIV leakage is also limited when the current measured rate is > 1.0 gpm, such that the current measured rate shall not exceed the rate determined by the previous test by an amount that reduces the margin between measured leakage rate and 5.0 gpm by 50%.

O INSERT 7 Valves CF-30 and CF-31 will be tested with the RCS pressure > 1200 psig and valves DH-76 and DH-77 will be tested at > 575 psig (i.e., the normal core flooding tank pressure). Minimum differential test pressure across each valve shall be > 150 psid.

Additionally, to satisfy ALARA requirements, leakage may be measured indirectly (as from the performance of pressure indicators) if accomplished in accordance with approved procedures and supported by computations showing that the method is capable of demonstrating valve compliance with the leakage criteria.

Insert Page B 3.4.14-5 Attachment 1, Volume 9, Rev. 0, Page 333 of 415

Attachment 1, Volume 9, Rev. 0, Page 334 of 415 RCS PIV Leakage B 3.4.14 BASES SURVEILLANCE REQUIREMENTS (continued) in MODE 5 for 7 days or more, if leakage testing has not been performed in the previous 9 months. In addition, this Surveillan e is not required to be performed on the/ DHR System when the DHR Sy tem is aligned to the RCS in the de heat removal mode of operati n. PIVs contained ir the DHR flow path ust be leakage rate tested afte DHR is secured and 0

stable unit conditio s and the necessary differential pressures are established.J

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

The "24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ..." Fre uency of performance for Surveill nce Requirement 3.4.14. is not required for B&W Owner's Group plants licensed prior to 198 . These plants were licensed pri r to the NRC establishing formal echnical Specification controls fo pressure isolation valves. Subsequen ly, these earlier plants had their Ii nses modified by NRC Order to requi e certain PIV testing Frequencies (excluding the "24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ..." Freque cy) be included in that plant's Tec nical Specifications. Ba d upon the information available o the Staff at the (D time, the content of those Orders was considered acc ptable. Since 1980, the NRC Sta has determined an additional PI leakage rate determination is re uired within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months following a uation of the valve and flow through t valve. This is necessary in ord to ensure the PIVs ability to support th integrity of the reactor coolant p essure boundary.

The Revised Stan ard Technical Specifications inclu e the "24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ..."

Frequency to refle current NRC Staff position on th need to include this test requirement thin Technical Specifications.

[SR 3.4.14,and SR 3.4.14. 430 psig, the pressure at which this section of DHR piping was tested 00 Verifying that the DHR auto osure interlocks are OPERABLE ensures at the RCS pressure instrumentation tap that RCS pressure will not pressurize the DHR system beyond _

Its design pres tele of [600] psil. The interlock setpoint that prevents the 0 valves from being opened is set so the actual RCS pressure must be 0

allows DH-1-1 and DH-12 to be opened by the operator prior to the point where net positive 328 < psi to open the valves. This setpoint ensures thg DHR desig ressure wi no e exceede an t R relief Pe valveswill not lifit. he 0D0 suction pressure is lost to the reactor coolant pumps 24 1 month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage and the potential for 0D (D

24 an unplanned transient if the Surveillance was performed with the reactor at power. Th1 month Frequency is also acceptable based on consideration of the design reliability (and confirming operating experience) of the equipment.

BWOG STS B 3.4.14-6 Rev. 3.1, 12/01/05 Attachment 1, Volume 9, Rev. 0, Page 334 of 415

Attachment 1, Volume 9, Rev. 0, Page 335 of 415 RCS PIV Leakage B 3.4.14 BASES SURVEILLANCE REQUIREMENTS (continued) System interlock These SRs are modified by Notes allowing the DHR Iu losurel function to be disabled when using the DHR System suction relief valve for cold ci overpressure protection in accordance with .LCO 3.4.12.m.-. 2 REFERENCES 1. 10 CFR'50.2.

2. 10 CFR 55a(c).

3. 10 CFR 50, Appendix A, Section V, GDC 55.

4. NUREG-75/014, Appendix V, October 1975.

5. NUREG-067 -ý,RC, May 1980. I

16. [Document co,tg-ining list of PIVs.]

7. ASME Code for Operation and Maintenance of Nuclear Power Plants[ 1995 Edition with 1996 Addenda.

8. 10 CFR 50.55a(g).

'5. Letter from D.G. Eisenhut, NRC, to all LWR Licenses, LWR Primary Coolant System 0

Pressure Isolation Valves. February 23, 1980.

6. Letter from J.F. Stoltz, NRC, to R.P. Crouse, Order for Modification of License Concerning]

Primary Coolant System Pressure Isolation Valves, April 20, 1981.

BWOG STS B 3.4.14-7 Rev. 3.1, 12/01/05 Attachment 1, Volume 9, Rev. 0, Page 335 of 415

Attachment 1, Volume 9, Rev. 0, Page 336 of 415 B 3.4.14 O INSERT 8 This allowance is necessary since opening and removing control power to the DHR System isolation valves (as required by LCO 3.4.12) disables the interlock.

O INSERT 9 SR 3.4.14.5 SR 3.4.14.5 requires the performance of a CHANNEL CALIBRATION of the DHR System interlock channels (both the channel common to the SFAS instrumentation and the channel not common to the SFAS instrumentation). The calibration verifies the accuracy of the instrument string. The Frequency of 24 months is a typical refueling cycle and considers channel reliability. Operating experience has proven this Frequency is acceptable.

Insert Page B 3.4.14-7 Attachment 1, Volume 9, Rev. 0, Page 336 of 415

Attachment 1, Volume 9, Rev. 0, Page 337 of 415 JUSTIFICATION FOR DEVIATIONS ITS 3.4.14 BASES, RCS PRESSURE ISOLATION VALVE (PIV) LEAKAGE

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 are made to reflect changes made to the Specification.

4. The Reviewer's Note is deleted because it is not intended to be included in the plant specific ITS submittal.

5. Changes made to be consistent with the Specification.

Davis-Besse Page 1 of 1 Attachment 1, Volume 9, Rev. 0, Page 337 of 415

Attachment 1, Volume 9, Rev. 0, Page 338 of 415 Specific No Significant Hazards Considerations (NSHCs)

Attachment 1, Volume 9, Rev. 0, Page 338 of 415

Attachment 1, Volume 9, Rev. 0, Page 339 of 415 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.4.14, RCS PRESSURE ISOLATION VALVE (PIV) LEAKAGE There are no specific NSHC discussions for this Specification.

Davis-Besse Page 1 of 1 Attachment 1, Volume 9, Rev. 0, Page 339 of 415

Attachment 1, Volume 9, Rev. 0, Page 340 of 415 ATTACHMENT 15 ITS 3.4.15, RCS LEAKAGE DETECTION INSTRUMENTATION Attachment 1, Volume 9, Rev. 0, Page 340 of 415

, Volume 9, Rev. 0, Page 341 of 415 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) , Volume 9, Rev. 0, Page 341 of 415

Attachment 1, Volume 9, Rev. 0, Page 342 of 415 ITS 3.4.15 ITS REACTOR COOLANT SYSTEM 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE LEAKAGE DETECTION SYSTEMS LIMITING CONDITION FOR OPERATION LCO 3.4.15 3.4.6.1 The following Reactor Coolant System leakage detection systems shall be OPERABLE:

a. The containment sump[ ley61 anfow onitoring system, and

b. One containment atmosphere radioactivity monitor (gaseous or particulate).

APPLICABILITY: MODES 1,2,3 and 4.

ACTION:1f- Add proposed Required Action A.1 Noe ACTION A a. ý With the required containment sumpl leytl anowmonitoring system inoperable, operation may continue up to 30 days provided Surveillance Requirement 4.4.6.2.1 .d is ýAV performed at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

ACTION B b. With the required containment atmosphere radioactivity monitor inoperable, operation may continue up to 30 days provided:

1. Containment atmosphere grab samples are obtained and analyzed at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , or

2. Surveillance Requirement 4.4.6.2.1 .d is performed at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

\ Add proposed Required Action B.1.2 NoteL0 ACTION C c. With the above required ACTION and associated completion time not met, be in at east HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

ACTION D d. With the required containment atmosphere radioactivity monitor and the containment sump level and flow monitoring system inoperable, enter TS 3.0.3 immediately.

SURVEILLANCE REQUIREMENTS 4.4.6.1 The leakage detection systems shall be demonstrated OPERABLE by:

SR 3.4.15.1, a. Containment atmosphere particulate monitoring system-performance of CHANNEL SR 3.4.15.2, CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST at the SR 3.4.15.3 frequencies specified in Table 4.3-3.

DAVIS-BESSE, UNIT 1 3/4 4-13 Amendment No. 234 Attachment 1, Volume 9, Rev. 0, Page 342 of 415 Page 1 of 5

Attachment 1, Volume 9, Rev. 0, Page 343 of 415 ITS 3.4.15 ITS REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued)

b. Containment sump leyl an Iow monitoring system-performance SR 3.4.15.4 of CHANNEL CALIBRATION at least once each REFUELING INTERVAL.

SR 3,4,15.1, c. Containment atmosphere gaseous monitoring system-performance of CHANNEL CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST SR 3.4.15.2, at the frequencies specified in Table 4.3-3.

SR 3.4.15.3 DAVIS-BESSE, UNIT I 3/4 4-14 Amendment No. 218 Page 2 of 5 Attachment 1, Volume 9, Rev. 0, Page 343 of 41 5

Attachment 1, Volume 9, Rev. 0, Page 344 of 415 LO ItO

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Attachment 1, Volume 9, Rev. 0, Page 345 of 415 ITS 3.4.15 ITS TABLE 3.3-6 (Continued)

TABLE NOTATION ACTION B ACTION 21 - With the number of channels OPERABLE less than required by I the Ninimum Channels OPERABLE requirement, comply vith the ACTION requirements of Specification 3.4.6.1.

ACTION 22 - With the number of channels OPERABLE less than required by the Minimum Channels- OPERABLE-.requirement, comply vith the See ITS 3.3.14 ACTION requirement: of Specification 3.9.12.

DAVIS-BESSS, UNIT 1 3/4 3-33 Amendment No. 135 Page 4 of 5 Attachment 1, Volume 9, Rev. 0, Page 345 of 415

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tj c-i Amendment No. 218, 234 DAVIS-BESSE. UNIT I 3/4 3-34 Attachment 1, Volume 9, Rev. 0, Page 346 of 415

Attachment 1, Volume 9, Rev. 0, Page 347 of 415 DISCUSSION OF CHANGES ITS 3.4.15, RCS LEAKAGE DETECTION 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.

MORE RESTRICTIVE CHANGES None RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES LA01 (Type I - Removing Details of System Design and System Description, Including Design Limits) CTS 3.4.6.1 .a states that the containment sump monitoring system includes both "level and flow." In addition, CTS 3.4.6.1 Action a and CTS 4.4.6.1 .b both include "level and flow" when referring to the containment sump monitoring system. ITS 3.4.15 requires the containment sump monitor to be OPERABLE, but the details of what constitutes an OPERABLE monitor are moved to the Bases. This changes the CTS by moving the details of what constitutes an OPERABLE containment sump monitor 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. 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 CTS.

LA02 (Type 1 - Removing Details of System Design and System Description, Including Design Limits) CTS Table 3.3-6 provides the measurement range for the gaseous and particulate containment atmosphere radioactivity monitors.

ITS 3.4.15 requires either the gaseous or particulate containment atmosphere radioactivity monitor to be OPERABLE, but the details concerning their measurement range are not included. This changes the CTS by moving the Davis-Besse Page 1 of 2 Attachment 1, Volume 9, Rev. 0, Page 347 of 415

Attachment 1, Volume 9, Rev. 0, Page 348 of 415 DISCUSSION OF CHANGES ITS 3.4.15, RCS LEAKAGE DETECTION INSTRUMENTATION details of the measurement ranges for the gaseous and particulate containment atmosphere radioactivity monitors to the UFSAR, where it currently exists.

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. Also, this change is acceptable because the removed information will be adequately controlled in the UFSAR. Changes to the UFSAR 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 information relating to system design is being removed from the CTS.

LESS RESTRICTIVE CHANGES L01 (Category 4 - Relaxation of Required Action) CTS 3.4.6.1 Actions a and b.2 do not include an exclusion allowing a delay in performing an RCS water inventory balance. ITS 3.4.15 Required Action A.1 and Required Action B.1.2 include a Note that states "Not required until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after establishment of steady state operation." This changes the CTS by allowing 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after establishment of steady state operation before the RCS water inventory balance must be performed.

The purpose of CTS 3.4.6.1 Actions a and b.2 to perform an RCS water inventory balance is to provide another means of leakage detection. This change is acceptable because the Required Actions are used to establish remedial measures that must be taken in response to 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 of remaining features, a reasonable time for repairs or replacement of required feature, and the low probability of a DBA occurring during the repair period. The RCS water inventory balance is still performed, but the delay in performing it allows unit conditions to provide an accurate indication. 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 2 of 2 Attachment 1, Volume 9, Rev. 0, Page 348 of 415

Attachment 1, Volume 9, Rev. 0, Page 349 of 415 Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs)

Attachment 1, Volume 9, Rev. 0, Page 349 of 415

Attachment 1, Volume 9, Rev. 0, Page 350 of 415 CTS RCS Leakage Detection Instrumentation 3.4.15 3.4 REACTORCOOLANT SYSTEM (RCS) 3.4.15 RCS Leakage Detection Instrumentation 3.4.6.1, LCO 3.4.15 The following RCS leakage detection instrumentation shall be Table 3.3-6 OPERABLE:

Instruments 2.a and 2.b

a. One containment sump monito and 0

b. One containment atmosphere radioactivity monitor (gaseous or particulate).

APPLICABILITY: MODES 1, 2, 3, and 4.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME Action a A. Required containment A.1 -------- NOTE-----

sump monitor Not required until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months inoperable, after establishment of steady state operation.

Perform SR 3.4.13.1. Once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months AND A.2 Restore required 30 days containment sump monitor to OPERABLE status.

Action b B. Required containment B.1.1 Analyze grab samples of Once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months atmosphere radioactivity the containment monitor inoperable, atmosphere.

OR BWOG STS 3.4.15-1 Rev. 3.0. 03/31/04 Attachment 1, Volume 9, Rev. 0, Page 350 of 415

Attachment 1, Volume 9, Rev. 0, Page 351 of 415 CTS RCS Leakage Detection Instrumentation 3.4.15 ACTIONS (continued)

CONDITION { REQUIRED ACTION COMPLETION TIME B.1.2 ..... .....- NOTE ----------

Not required until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after establishment of steady state operation.

Perform SR 3.4.13.1. Once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months AND B.2 Restore.required 30 days containment atmosphere radioactivity monitor to OPERABLE status.

Action c C. Required Action and C.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time not met. AND of Condition A or BC .2 Be in MODE 5. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months Q

Action d D. Both required monitors D.1 Enter LCO 3.0.3. Immediately inoperable.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY 4.4.6.1.a, SR 3.4.15.1 Perform CHANNEL CHECK of required containment 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months 4.4.6.1.c atmosphere radioactivity monitor.

4.4.6.1.a, SR 3.4.15.2 Perform CHANNEL FUNCTIONAL TEST of required jl~days 4.4.6.1.c containment atmosphere radioactivity monitor. 0 BWOG STS 3.4.15-2 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 0, Page 351 of 415

Attachment 1, Volume 9, Rev. 0, Page 352 of 415 CTS RCS Leakage Detection Instrumentation 3.4.15 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY 4.4.6. 1.b SR 3-4.15ýJ Perform CHANNEL CALIBRATION of required containment sump monitor.

1111 8Mj months 0 4.4.6.1.a, SR 3.4.15. Perform CHANNEL CALIBRATION of required V 8M months 0 Table 4.33 containment atmosphere radioactivity monitor.

Instruments 2.a.i and 2.a.ii BWOG STS 3.4.15-3 Rev. 3.0, 03131/04 Attachment 1, Volume 9, Rev. 0, Page 352 of 415

Attachment 1, Volume 9, Rev. 0, Page 353 of 415 JUSTIFICATION FOR DEVIATIONS ITS 3.4.15, RCS LEAKAGE DETECTION INSTRUMENTATION

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

2. The specific Conditions the ACTION applies to have been added, since there is one ACTION it does not apply to (ACTION D). This is consistent with the Writers Guide for the Improved Standard Technical Specifications, TSTF-GG-05-01, Section 4.1.6.i.5.ii.

3. The CHANNEL FUNCTIONAL TEST Frequency has been changed to be consistent with the Davis-Besse current licensing basis.

4. The brackets have been removed and the proper plant specific information/value is provided. Also, the Surveillances have been put in the correct order based on the Frequency.

Davis-Besse Page 1 of 1 Attachment 1, Volume 9, Rev. 0, Page 353 of 415

Attachment 1, Volume 9, Rev. 0, Page 354 of 415 Improved Standard Technical Specifications (ISTS) Bases Markup and Justification for Deviations (JFDs)

Attachment 1, Volume 9, Rev. 0, Page 354 of 415

Attachment 1, Volume 9, Rev. 0, Page 355 of 415 RCS Leakage Detection Instrumentation B 3.4.15 B 3.4 REACTOR COOLANT SYSTEM (RCS)

B 3.4.15 RCS Leakage Detection Instrumentation BASES BACKGROUND GDC 30 of Appendix A to 10 CFR 50 (Ref. 1) requires means for rAlthough n detectin and, to the extent practical, identifying the location of the source committed to of RCS LEAKAGE. Regulatory Guide 1.45 (Ref. 2) describes acceptable methods for selecting leakage detection systems. 0 Leakage detection systems must have the capability to detect significant reactor coolant pressure boundary (RCPB) degradation as soon after occurrence as practical to minimize the potential for propagation to a gross failure. Thus, an early indication or warning signal is necessary to permit proper evaluation of all unidentified LEAKAGE.

Industry practice has shown that water flow changes of 0.5 to 1.0 gpm can readily be detected in contained volumes by monitoring changes in water level, in flow rate, or in the operating frequency of a pump. The acontainment sump used to collect unidentified LEAKAGE is instrumented detecting to increases of 0.5 to 1.0 gpm in the normal flow rates. This sensitivity is acceptable for detecting increases in unidentified LEAKAGE.

0 The reactor coolant contains radioactivity that, when released to the containment, can be detected by radiation monitoring instrumentation.

Reactor coolant radioactivity levels will be low during initial reactor startup and for a few weeks thereafter until activated corrosion products have been formed and fission products appear from fuel element cladding contamination or cladding defects. Instrument sensitivities of 10-9 PCicc radioactivity for particulate monitoring and of 10-6 pCi/cc radioactivity for gaseous monitoring are practical for these leakage detection systems.

Radioactivity detection systems are included for monitoring both particulate and gaseous activities because of their sensitivities and rapid responses to RCS LEAKAGE.

An increase in hunidity of the containmer* atmosphere would ndicate release of waterpor to the containmet. Dew point temperature measurements n thus be used to rnitor humidity leve of the containment mesphere as an inditor of potential RC LEAKAGE. A 0

1°F increa in dew point is well thin the sensitivity* nge of available instrume s.

BWOG STS B 3.4.15-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 0, Page 355 of 415

Attachment 1, Volume 9, Rev. 0, Page 356 of 415 RCS Leakage Detection Instrumentation B 3.4.15 BASES BACKGROUND (continued)

Since the humidityovel evaluation of an i, dicatedis leakage influenced rateby ,IWthis severalmeans factors, maya qu b/e titative questionable a should be comparedo observed increa s in liquid flowlinto or froj the containment su p [and condensate owfrom air coolers]. Hnidity level monitorin is considered most seful as an indirect al m or indication to ale the operator to a p tential problem.

0 Humidity monitors are not .requ ed for this LCO.

.Air temperature and pressure monitoring methods may also be used to infer unidentified LEAKAGE to the containment. Containment temperature and pressure fluctuate slightly during plant operation, but a rise above the normally indicated range of values may indicate RCS LEAKAGE into the containment. The relevance of temperature and pressure measurements are affected by containment free volume and, for temperature, detector location. Alarm signals from these instruments can be valuable in recognizing rapid and sizable leakage to the containment.

Temperature and pressure monitors are not.required by this LCO.

APPLICABLE The need to evaluate the severity of an alarm or an indication is important SAFETY to the operators, and the ability to compare and verifywith indications ANALYSES from other systems is necessary. tThe system response/times and I Isensitivities are detribed in the FSAR (Ref. 3). FMultiple instrument 0

locations are utilized, it needed, to ensure the transport delay time of the leakage from its source to an instrument location yields an acceptable overall response time.

The safety significance of RCS LEAKAGE varies widely depending on its source, rate, and duration. Therefore, detecting and monitoring reactor coolant LEAKAGE into the containment area are necessary. Quickly separating the identified LEAKAGE from the unidentified LEAKAGE provides quantitative information to the operators, allowing them to take corrective action should a leak occur detrimental to the safety of the unit and the public.

RCS leakage detection instrumentation satisfies Criterion 1 of 10 CFR 50.36(c)(2)(ii).

fRefer to the Bases of LCO 3.4.13, "RCS Operational LEAKAGE."

for further information regarding RCS LEAKAGE.

Q BWOG STS B 3.4.15-2 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 0, Page 356 of 415

Attachment 1, Volume 9, Rev. 0, Page 357 of 415

'RCS Leakage Detection Instrumentation B 3.4.15 BASES LCO One method of protecting against large RCS LEAKAGE derives from the ability of instruments to rapidly detect extremely small leaks. This LCO requires -instruments of diverse monitoring principles to be OPERABLE to provide a ýhigh degree of confidencethat extremely small leaksare detected in time to allow actionsto place the plant in a safe condition when RCS LEAKAGE indicates possible RCPB degradation.

The LCO requirements are satisfied when monitors of diverse measurement means are available. Thus, the containment sump monitor in combination with a particulate or gaseous radioactivity monitor, provides an acceptable minimum. (both the level and fl.

portions)

APPLICABILITY Because of elevated RCS temperature and pressure in MODES 1, 2, 3, and 4, RCS leakage detection instrumentation is required to be OPERABLE.

In MODE 5 or 6, the temperature is s- 200°F and pressure is maintained low or at atmospheric pressure. Since the temperatures and pressures are far lower than those for MODES 1,2, 3, and 4, the likelihood of leakage and crack propagation is much smaller. Therefore, the requirements of this LCO are not applicable in MODES 5 and 6.

ACTIONS A.1 and A.2 [(i.e., either level or flow or both) (

With the required containment sump monitor inoperable, nnoother form of sampling can provide the equivalent information. r However, the containment atmosphere ct' i monitor will provide indications of changes in leakage. Together with the atmosphere monitor, the periodic surveillance for RCS inventory balance, SR 3.4.13.1, water inventory balance, must be performed at an increased frequency of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to provide information that is adequate to detect leakage. A Note is added allowing that SR 3.4.13.1 is not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after establishing steady state operation (stable L-Lf-i temperature, power level, pressurizerland mna up tank level s, mioepi2 land letdown, and [RCP. jection and return flows). The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months allowance provides sufficient time to collect and process all necessary data after stable plant conditions are established. cinment Restoration of the required'sump monitor to OPERABLE status is required to regain the function in a Completion Time of 30 days after the monitor's failure. This time is acceptable considering the frequency and adequacy of the RCS water inventory balance required by Required Action A.1.

BWOG STS B 3.4.15-3 Rev. 3.0, 03131/04 Attachment 1, Volume 9, Rev. 0, Page 357 of 415

Attachment 1, Volume 9, Rev. 0, Page 358 of 415 RCS Leakage Detection Instrumentation B 3.4.15 BASES ACTIONS (continued)

B.1.1, B.1.2, and B.2 With required gaseous or particulate containment atmosphere radioactivity monitoring instrumentation channels inoperable, alternative action is required. Either grab samples of the containment atmosphere must be taken and analyzed or water inventory balances, in accordance with SR 3.4.13.1, must be performed to provide alternate periodic information. With a sample obtained and analyzed or a water inventory balance performed every 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , the reactor may be operated for up to 30 days to allow restoration of at least one of the radioactivity monitors.

The 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months interval provides periodic information that is adequate to detect leakage. A Note is added allowing that SR 3.4.13.1 is not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after establishing stead state operation (stable temperature, power levelpressurizerland ma dup tanklievel,] 2 make nd letdown, and a-injection an urn flows ). The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months allowance provides sufficient time to collect and process all necessary data after stable plant conditions are established. The 30 day Completion Time recognizes at least one other form of leak detection is available.

C.1 and C.2 If a Required Action of Condition A or B cannot be met within the required Completion Time, the unit must be brought to a MODE in which the LCO does not apply. To achieve this status, the unit must be brought to at least MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and to MODE 5 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

D. 1 With both required monitors inoperable, no automatic means of monitoring leakage are available, and immediate plant shutdown in accordance with LCO 3.0.3 is required.

BWOG STS B 3.4.15-4 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 0, Page 358 of 415

Attachment 1, Volume 9, Rev. 0, Page 359 of 415 RCS Leakage Detection Instrumentation B 3.4.15 BASES SURVEILLANCE SR 3.4.15.1 REQUIREMENTS SR 3.4.15.1 requires the performance of a CHANNEL CHECK of the required containment atmosphere radioactivity monitor. The check gives reasonable confidence thateach channel is operating properly. The Frequency of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is based on instrument reliability and is reasonable for detecting off normal conditions.

SR 3.4.15.2 SR 3.4.15.2 requires the performance of a CHANNEL FUNCTIONAL TEST of the required containment atmosphere radioactivity monitor.c Ait successful tesithofe required conctý(s) of a channel resla /a'y be performed tevrfcation ey of t hange of state of onteact t, 2~ge the relay. This/clarifie~s wat is / acceptable CHAN N E/FUNCTIONAL TEST of a Ly.

rel/ This sace!l because all of the ýther required Q

contacts of e relay are verified by other Technical S)*cifications and non-Techr cal Specifications/tests at least once per refueling interval with applicabi* extensions. The test ensures that the monitor can perform its function in the desired manner. The test verifies the alarm setpoint and relative accuracy of the instrument string. The Frequency of Wdays 0

considers instrument reliability, and operating experience has shown it proper for detecting degradation.

SR 3.4.15.3 and SR 3.4.15.4 These SRs require the performance of a CHANNEL CALIBRATION for or 24 months, :

each of the required RCS leakage detectioninstrument of the string, channels.

instrumentation including the as applicable,

.j The calibration verifies the accuracy instruments located inside containment. The Frequency of months X typi retu g cy and considers channel reliability and operating experience has proven this Frequency is acceptable.

REFERENCES 1. 10 CFR 50, AppendixA,Section IV,GDC 30.

2. Regulatory Guide 1.45.

3. FSAR ection 1. 0 BWOG STS B 3.4.15-5 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 0, Page 359 of 415

Attachment 1, Volume 9, Rev. 0, Page 360 of 415 JUSTIFICATION FOR DEVIATIONS ITS 3.4.15 BASES, RCS LEAKAGE DETECTION INSTRUMENTATION

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

2. Changes are made (additions, deletions, and/or changes) to the ISTS Bases which reflect the plant specific nomenclature, number, reference, system description, analysis, or licensing basis description.

3. Changes made to be consistent with the Specification.

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

Davis-Besse Page 1 of 1 Attachment 1, Volume 9, Rev. 0, Page 360 of 415

Attachment 1, Volume 9, Rev. 0, Page 361 of 415 Specific No Significant Hazards Considerations (NSHCs)

Attachment 1, Volume 9, Rev. 0, Page 361 of 415

Attachment 1, Volume 9, Rev. 0, Page 362 of 415 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.4.15, RCS LEAKAGE DETECTION INSTRUMENTATION There are no specific NSHC discussions for this Specification.

Davis-Besse Page 1 of 1 Attachment 1, Volume 9, Rev. 0, Page 362 of 415

, Volume 9, Rev. 0, Page 363 of 415 ATTACHMENT 16 ITS 3.4.16, RCS SPECIFIC ACTIVITY , Volume 9, Rev. 0, Page 363 of 415

, Volume 9, Rev. 0, Page 364 of 415 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) , Volume 9, Rev. 0, Page 364 of 415

Attachment 1, Volume 9, Rev. 0, Page 365 of 415 ITS 3.4.16 ITS REACTOR COOLAX*T SYSTE-N SPECIFIC ACTIVITY LIMITING COhDITIOF FOR OPERATION LCO 3.4.16 3.4.8 The specific activity of the primary coolant shall be limited to:

SR 3.4.16.2 a. 1.0 IjCi/gram DOSE EQUIVAILNT 1-131, and SR 3.4.16.1 b. < 1001E pCi/gram APPLICABILrT: MODES 1, 2, j LOl ACTION:

MODES 1, 2 and 3-. Add proposed ACTION A!L02-Note, ACTION A a. 'itb the specific activity of the primary coolant > 1.0 I-Ci/gram DOSE EQUIVAILNT 1-131 for more than 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months during one continuous time interval pr exceeding the limit line shou.n ACTION B on Igure . - , e in at ýeast NOT STANDBY vith T c 530OF onithin 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ,. avg ACTION B b. With the specific activity of the primary coolant > 100/E tiCi/grat, be in at least NOT STANDBY vith T c 5300F within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . avg MODES 1, 2, ~1ODES I, LOl ACTION A a. With the specific activity of the primary cooli JCi/grar DOSI £QUIVALE)' 1-131 Fcfjljj!j7/

sampling and analysis requirements of item 4 a luctil tbaspecific activitv of the ift nary coo:

SURVIIULLACE- REQUIREMENTS SR 3.4.16.1, 4.4.8 The specIfi: activity of the primary coolant shall be determined SR 3.4.16.2, to be vithin the limits by performance of tbe sampling and analysis SR 3.4.16.3 program of Table 4.4-4.

Applicability - *Wit~h T7v 5300F.

DAVIS-BFNSS, UNIT 2 3/4 4-20 Amendment No. P.5'.14 Page 1 of 3 Attachment 1, Volume 9, Rev. 0, Page 365 of 415

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Attachment 1, Volume 9, Rev. 0, Page 366 of 415

Attachment 1, Volume 9, Rev. 0, Page 367 of 415 ITS 3.4.16 ITS Figure 3.4.16-1

.21 40 so to 70 U go 10 PIRFCI'OF. RATED THERMAL POWER FIGURE 34.1 DOSE EqUIVALENT 1-131 Primwy Coolmnt Sie/flc Aivity Umit Vwm Paint of RATED THERMAL POWER with h*t Priary Coolant Specific Activty > 1.0,Ci/4rm DOSE EQUIVALENT 1-131 DAVISS4ESSE. UNIT 1 3/4 4423 .1mendrnent 'jo. 135 Page 3 of 3 Attachment 1, Volume 9, Rev. 0, Page 367 of 415

Attachment 1, Volume 9, Rev. 0, Page 368 of 415 DISCUSSION OF CHANGES ITS 3.4.16, RCS SPECIFIC ACTIVITY 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 3.4.8 Action a (MODES 1, 2, 3, 4, and 5) and CTS Table 4.4-4, Footnote #,

require the isotopic analysis for iodine to be performed until the specific activity of the primary coolant system is restored to within limits. ITS 3.4.16 Required Action A.1 requires this same analysis, however the explicit statement to perform the isotopic analysis for iodine until the limits are met has been deleted. This changes the CTS by deleting the explicit statement to perform the isotopic analysis for iodine until the limits are met.

The purpose of the CTS 3.4.8 Action a (MODES 1, 2, 3, 4, and 5) and CTS Table 4.4-4 is to ensure the Surveillance is performed to determine whether the specific activity is met. This statement is not necessary in the ITS, because ITS LCO 3.0.2 requires the Required Actions of the associated Conditions to be met upon discovery of failure to meet an LCO. If the LCO is met or is no longer applicable prior to expiration of the specified Completion Time(s), completion of the Required Action(s) is not recuired unless otherwise stated. This change is acceptable since ITS LCO 3.0.4 will require the Required Action to be performed until the LCO is met. This change is designated as administrative because it does not result in technical changes to the CTS.

A03 CTS 3.4.8 Action a (MODES 1, 2, 3, 4, and 5) provides a cross-reference to CTS 6.9.1.5.c, the Annual Operating Report. ITS 3.4.16 does not contain this cross-reference. This changes the CTS by deleting a cross-reference to another CTS requirement.

The purpose of the reference is to alert the user that a report may need to be generated due to the specific activity being outside the limit. However, CTS 6.9.1.5.c has not been included in the Davis-Besse ITS. Therefore, the cross-reference is not needed. Furthermore, it is an ITS convention to not include these types of cross-references. This change is designated as administrative because it does not result in technical changes to the CTS.

MORE RESTRICTIVE CHANGES None RELOCATED SPECIFICATIONS None Davis-Besse Page 1 of 5 Attachment 1, Volume 9, Rev. 0, Page 368 of 415

Attachment 1, Volume 9, Rev. 0, Page 369 of 415 DISCUSSION OF CHANGES ITS 3.4.16, RCS SPECIFIC ACTIVITY REMOVED DETAIL CHANGES LA01 (Type 3 - Removing ProceduralDetails for Meeting TS Requirements or Reporting Requirements) CTS Table 4.4-4 Item 2 requires an isotopic analysis to determine whether DOSE EQUIVALENT 1-131 concentration is within limit.

CTS Table 4.4-4 Item 4 requires an isotopic analysis for iodine including 1-131, 1-133, and 1-135. ITS SR 3.4.16.2 requires the verification that reactor coolant DOSE EQUIVALENT 1-131 specific activity is within limit. ITS 3.4.16 Required Action A.1 requires the verification that DOSE EQUIVALENT 1-131 is within the acceptable region of Figure 3.4.16-1. This changes the CTS by moving the detail that an "Isotopic Analysis" or "Isotopic Analysis for Iodine Including 1-131, 1-133, and 1-135" must be performed to satisfy the requirements of the Surveillances 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. ITS SR 3.4.16.2 and ITS 3.4.16 Required Action A.1 still retain the requirements to verify reactor coolant DOSE EQUIVALENT 1-131 is within limit. Also, this change is acceptable because these types of procedural details will be adequately controlled in the ITS Bases.

Changes to the Bases are controlled by the Technical Specification Bases Control Program in Chapter 5. This program provides for the evaluation of changes to ensure the Bases are properly controlled. This change is designated as a less restrictive removal of detail change because procedural details for meeting Technical Specification requirements are being removed from the Technical Specifications.

LESS RESTRICTIVE CHANGES L01 (Category2 - Relaxation of Applicability) CTS 3.4.8 is applicable in MODES 1, 2, 3, 4, and 5. In addition, the testing for gross activity determination in CTS Table 4.4-4 Item 1 is required in MODES 1, 2, 3, and 4, and the isotopic analysis for iodine requirement in CTS Table 4.4-4 Item 4.a and 4.b is required periodically in MODES 1, 2, 3, 4, and 5 and after a 15% RTP change in MODES 1,2, and 3, respectively. ITS 3.4.16, including the Surveillances, is applicable in MODES 1 and 2, and MODE 3 with RCS Tavg > 530 0 F. This changes the CTS by reducing the MODES in which the LCO is applicable, including the Surveillances, to only MODES 1 and 2, and MODES 3 with RCS Tavg > 5300F.

The purpose of CTS 3.4.8 is to ensure that the specific activity of the RCS is within the assumptions of the Steam Generator Tube Rupture (SGTR) 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. During operation in MODE 3 with RCS Tavg < 530 0 F, and in MODES 4 and 5, the release of radioactivity in the event of a SGTR is unlikely because the saturation pressure of the reactor coolant is below the lift pressure settings of the main steam safety valves. Furthermore, the CTS Actions for when the limits are not met only Davis-Besse Page 2 of 5 Attachment 1, Volume 9, Rev. 0, Page 369 of 415

Attachment 1, Volume 9, Rev. 0, Page 370 of 415 DISCUSSION OF CHANGES ITS 3.4.16, RCS SPECIFIC ACTIVITY require the unit to be shutdown to MODE 3 with RCS Tavg < 530 0 F. This change is designated as less restrictive because the LCO requirements are applicable in fewer operating conditions than in the CTS.

L02 (Category 9 - Addition of LCO 3.0.4 Exception) CTS 3.4.8 does not allow the unit to change MODES when the RCS specific activity is not within limits. ITS 3.4.16 ACTION A Note specifies that LCO 3.0.4.c is applicable. This changes the CTS by allowing the unit to change MODES or other specified conditions in the Applicability when the specific activity for DOSE EQUIVALENT 1-131 is

> 1.0 pCi/gm.

The purpose of CTS 3/4.4.8 is to ensure appropriate limitations are placed on reactor coolant activity. 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 that the DOSE EQUIVALENT 1-131 is still within the limits of ITS Figure 3.4.16-1. This includes the low probability of a DBA occurring during the restoration time period. This change allows the unit to change MODES or other specified conditions in the Applicability when the specific activity for DOSE EQUIVALENT 1-131 is > 1.0 pCi/gm. However, after entering the Applicability the unit must enter ACTION A and verify DOSE EQUIVALENT 1-131 is within the acceptable region of Figure 3.4.16-1 every 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . This verification will ensure that a steam generator tube rupture will not lead to a site boundary dose that exceeds the 10 CFR 100 dose guideline limits. Therefore, this change is acceptable due to the significant conservatism incorporated into the specific activity limit, the low probability of event which is limiting due to exceeding this limit, and the ability to restore transient specific activity excursions while the unit remains at, or proceeds to power operation. In addition, ITS 3.4.16 ACTION A requires DOSE EQUIVALENT 1-131 to be within limit in 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months . This change is designated as less restrictive because the Required Action Note allows entry into the MODE of Applicability when the specific activity for DOSE EQUIVALENT 1-131 is > 1.0 pCi/gm.

L03 (Category4 - Relaxation of Required Action) CTS 3.4.8 Action a (MODES 1, 2, 3, 4, and 5) and CTS Table 4.4-4 Item 4.a require isotopic analysis for iodine once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months when the specific activity exceeds 100/E pCi/gm. The ITS does not contain this Action. This changes the CTS by eliminating a conditionally performed Surveillance when gross activity exceeds 100/E pCi/gm.

The purpose of CTS 3.4.8 Action a (MODES 1, 2, 3, 4, and 5) and CTS Table 4.4-4 Item 4.a is to monitor iodine activity when the specific activity limits are exceeded. 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 that DOSE EQUIVALENT 1-131 is still being monitored and the low probability of a DBA occurring during the restoration time period. When specific Davis-Besse Page 3 of 5 Attachment 1, Volume 9, Rev. 0, Page 370 of 415

Attachment 1, Volume 9, Rev. 0, Page 371 of 415 DISCUSSION OF CHANGES ITS 3.4.16, RCS SPECIFIC ACTIVITY activity exceeds 100/f pCi/gm, ITS 3.4.16 Required Action B.1 and CTS 3.4.8 Action b (MODES 1, 2, and 3*) require the plant to be in MODE 3 with Tavg < 530'F within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . Monitoring of E is required in order to determine if the LCO is met and the ACTION can be exited. Furthermore, if the Condition is entered and the unit is in MODE 2 in 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or less, the Required Action is in conflict with the Note of ITS SR 3.4.16.2, which states that this SR is only required in MODE 1. Finally, this action is an unnecessary burden as the unit is required to be in MODE 3 with Tavg < 530'F within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , exiting the Applicability. 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 7- Relaxation Of Surveillance Frequency,Non-24 Month Type Change) CTS Table 4.4-4 Item 1 requires gross activity to be determined at least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . ITS SR 3.4.16.1 requires verification that the reactor coolant gross specific activity is < 100/F #TCi/gm every 7 days. This changes the CTS by reducing the Frequency from at least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months to 7 days.

The purpose of CTS Table 4.4-4 Item 1 is to obtain a quantitative measure of radionuclides with half lives longer than 15 minutes, excluding iodines, which provides an indication of increases in gross specific activity. This change is acceptable because the new Surveillance Frequency ensures that it provides an acceptable level of monitoring. A Frequency of 7 days provides sufficient information to trend the results in order to detect gross fuel failure, while considering the low probability of a gross fuel failure between performances.

This change is designated as less restrictive because Surveillances will be performed less frequently under the ITS than under the CTS.

L05 (Category 7- Relaxation Of Surveillance Frequency, Non-24 Month Type Change) CTS Table 4.4-4 Item 3 requires radiochemical determination of E once per 6 months. Footnote

states that the sample is to be taken after a minimum of 2 EFPD and 20 days of POWER OPERATION have elapsed since the reactor was last subcritical for 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or longer. ITS SR 3.4.16.3 requires f to be determined from a sample taken in MODE 1 after a minimum of 2 effective full power days and 20 days of MODE 1 operation have elapsed since the reactor was last subcritical for > 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months . ITS SR 3.4.16.3 is further modified by a Note which states, "Not required to be performed until 31 days after a minimum of 2 effective full power days and 20 days of MODE 1 operation have elapsed since the reactor was last subcritical for > 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months ." This changes the CTS by putting a limit, 31 days, on when the Surveillance must be performed after the requisite conditions are met.

The purpose of CTS Table 4.4-4 Item 3 is to determine the value of f when the isotopic concentrations in the core are stable. This change is acceptable because the new Surveillance Frequency has been evaluated to ensure that it provides an acceptable level of monitoring. Circumstances could arise in which the 6 month Frequency for performance of the SR has passed but the operating conditions for performance of the test have not been met. In this circumstance, the Surveillance would be immediately past due as soon as the operating conditions are met. The ITS SR 3.4.16.3 Note allows 31 days to perform the Davis-Besse Page 4 of 5 Attachment 1, Volume 9, Rev. 0, Page 371 of 415

Attachment 1, Volume 9, Rev. 0, Page 372 of 415 DISCUSSION OF CHANGES ITS 3.4.16, RCS SPECIFIC ACTIVITY Surveillance after the operating conditions are met. This change is designated as less restrictive because Surveillances will be performed less frequently under the ITS than under the CTS.

Davis-Besse Page 5 of 5 Attachment 1, Volume 9, Rev. 0, Page 372 of 415

Attachment 1, Volume 9, Rev. 0, Page 373 of 415 Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs)

Attachment 1, Volume 9, Rev. 0, Page 373 of 415

Attachment 1, Volume 9, Rev. 0, Page 374 of 415

. CTS RCS Specific Activity 3.4.16 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.16 RCS Specific Activity LCO 3.4.8 LCO 3.4.16 The specific activity of the reactor coolant shall be within limits.

APPLICABILITY: MODES 1 and 2, MODE 3 with RCS average temperature (T,,) _>WF.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME Action a A. DOSE EQUIVALENT - -NOTE (MODES 1, 2, and 3' 1-131 > 1.0 pCi/gm. LCO 3.0.4.c is applicable.

and MODES 1.2,3, 4, and 5),

Table 4.4-4 Item 4.a A.1 Verify DOSE EQUIVALENT Once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months 1-131 within the acceptable region of Figure 3.4.16-1.

AND A.2 Restore DOSE 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months EQUIVALENT 1-131 to within limit.

Action a B. Required Action and B.1 Be in MODE 3 with 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months (MODES 1, 2, and 3') associated Completion Tavg <

Time of Condition A not met.

OR Action a DOSE EQUIVALENT (MODES 1.2, 3, 4, 1-131 in unacceptable and 5) region of Figure 3.4.16-1.

Action b (MODES 1, 2, and 3*)

ross specific activity of the reactor]

coolant not within limit.

G BWOG STS 3.4.16-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 0, Page 374 of 415

Attachment 1, Volume 9, Rev. 0, Page 375 of 415

. CTS RCS Specific Activity 3.4.16 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME C. Gross spe ivity of C.1 Be in MODE 3 with 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months the coolant not within T < 500 0F. 2 limit.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY LCO 3.4.8.b, SR 3.4.16.1 Verify reactor coolant gross specific activity 7 days Table 4.4-4 Item 1 _ 100/*E PCygm.

LCO 3.4.8.a, SR 3.4.16.2 -------- -------------- NOTE ------- .-

Table 4.4-4 Item 2 Only required to be performed in MODE 1.

and Item 4.b Verify reactor coolant DOSE EQUIVALENT 1-131 14 days specific activity ::_1.0 piCilgm.

AND Between 2 and 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months after THERMAL POWER change of Ž 15% RTP within a 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months period Table 4.4-4 Item 3 SR 3.4.16.3 ----------------------- NOTE------------

Not required to be performed until 31 days after a minimum of 2 EFPD and 20 days of MODE 1 operation have elapsed since the reactor was last subcritical for -, 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months .

Determine E. 184 days BWOG STS 3.4.16-2 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 0, Page 375 of 415

Attachment 1, Volume 9, Rev. 0, Page 376 of 415

. CTS RCS Specific Activity 3.4.16 Figure 3.4-1 2 This figure for illustration 0 25\ Do not use for operation. only.

I-)

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Reactor Coolant DOSE EQUIVALENT 1-131 Specific Activity Limit Versus Percent of RATED THERMAL POWER With Reactor Coolant Specific Activity >1.0 pCi/gm DOSE EQUIVALENT 1-131 BWOG STS 3.4.16-3 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 0, Page 376 of 415

Attachment 1, Volume 9, Rev. 0, Page 377 of 415 3.4.16 INSERT 1 3D

. CTS (275,26) III 275i II I 1~.I liii, Figure 3.4-1

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Insert Page 3.4.16-3 Attachment 1, Volume 9, Rev. 0, Page 377 of 415

Attachment 1, Volume 9, Rev. 0, Page 378 of 415 JUSTIFICATION FOR DEVIATIONS ITS 3.4.16, RCS SPECIFIC ACTIVITY

1. The MODE 3 Applicability for this Specification has been changed from 500OF to 530 0 F, consistent with current licensing basis. The Davis-Besse temperature limit is 5300 F, since at this temperature the saturation pressure of the primary coolant is below the lift pressure of the main steam safety valves.

2. ISTS 3.4.16 ACTION C has been deleted and incorporated in ISTS 3.4.16 ACTION B because the Required Actions are identical (be in MODE 3 with Tavg < 500 0 F). In NUREG-1430, Rev. 1, ISTS 3.4.16 ACTION C contained an additional Required Action. This Required Action was deleted in NUREG-1430, Rev. 2, as a result of approved TSTF-28. The entire ACTION C should have been deleted as a result of the application of TSTF-28, but was not. This changes the ISTS to be consistent with other Specifications where ACTION Conditions are combined when the same Required Actions apply.

3. The Davis-Besse reactor coolant DOSE EQUIVALENT 1-131 specific power limit verses percent of RATED THERMAL POWER curve is substituted for the curve provided for illustration in the ISTS.

Davis-Besse Page 1 of 1 Attachment 1, Volume 9, Rev. 0, Page 378 of 415

Attachment 1, Volume 9, Rev. 0, Page 379 of 415 Improved Standard Technical Specifications (ISTS) Bases Markup and Justification for Deviations (JFDs)

Attachment 1, Volume 9, Rev. 0, Page 379 of 415

Attachment 1, Volume 9, Rev. 0, Page 380 of 415 RCS Specific Activity B 34.16 B 3.4 REACTOR COOLANT SYSTEM (RCS)

B 3.4.16 RCS Specific Activity BASES BACKGROUND The Code of Federal Regulations, 10 CFR 100 (Ref. 1), specifies the

,maximum dose to the whole body and the thyroid an individual at the site boundary can receive for 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months during an accident. The limits on specific activity ensure that the doses are held to a small fraction of the 10 CFR 100 limits during analyzed transients and accidents.

The RCS specific activity LCO limits the allowable concentration level of radionuclides in the reactor coolant. The LCO limits are established to minimize the offsite radioactivity dose consequences in the event of a steam generator tube rupture (SGTR) accident.

The LCO contains specific activity limits for both DOSE EQUIVALENT 1-131 and gross specific activity. The allowable levels are intended to limit the 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months dose at the site boundary to a small fraction of the 10 CFR 100 dose guideline limits. The limits in the LCO are standardized based on parametric evaluations of offsite radioactivity dose consequences for typical site locations.

The parametric evaluations showed the potential offsite dose levels for an SGTR accident were an appropriately small fraction of the 10 CFR 100 dose guideline limits (Ref. 1). Each evaluation assumes a broad range of site applicable atmospheric dispersion factors in a parametric evaluation.

APPLICABLE The LCO limits on the specific activity of the reactor coolant ensure that SAFETY the resulting 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months doses at the site boundary will not exceed a small ANALYSES fraction of the 10 CFR 100 dose guideline limits following an SGTR accident. The SGTR safety analysis (Ref. 2) assumes e specific activity f the re t at the LCO limits_ and an existing reactor coolant steam generator (SG) tube leakage rate of 1 gpm. The analysis also value equivalent assumes a reactor trip and a turbine trip at t- time as th GTR to 1%failedfuel [e -n The analysis for the SGTR accident establishes the acceptance limits for 3D RCS specific activity. Reference to this analysis is used to assess changes to the facility that could affect RCS specific activity as they relate to the acceptance limits.

[ analysis bounds the LCO limit for RCS specific activity.

The assumed RCS specific activity in the SGTR BWVOG STS B3.4.16-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 0, Page 380 of 415

Attachment 1, Volume 9, Rev. 0, Page 381 of 415 RCS Specific Activity B 3.4.16 BASES APPLICABLE SAFETY ANALYSES (continued)

The rise in pressure in the ruptured SG causes radioactively contaminated steam to discharge to the atmosphere through the atmos valves orthe main steam safety valves. The atmospheric discharge stops when the turbine bypass to the condenser removes the excess energy to rapidly reduce the RCS pressure and close the valves. The unaffected SG removes core decay heat by venting steam until the cooldown ends.

The safety analysis shows the radiological consequences of an SGTR accident are within a small fraction of the Reference 1 dose guideline limits. Operation with iodine specific activity levels greater than the LCO limit is permissible, if the activity levels do not exceed the limits shown in Figure 3.4.16-11, in e e Specification, for more than 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months . 5 The remainder of the above limit permissible iodine levels shown in Figure 3.4.16-1 are acceptable because of the low probability of an SGTR accident occurring during the established 48 hour5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months time limit. The occurrence of an SGTR accident at these permissible levels could increase the site boundary dose levels, but still be within 10 CFR 100 dose guideline limits.

RCS Specific Activity satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii).

LCO The specific iodine activity is limited to 1.0 pCi/gm DOSE EQUIVALENT 1-131, and the gross specific activity in the primary coolant is limited to the number of pCi/gm equal to 100 divided by E (average disintegration energy of the sum of the average beta and gamma energies of the coolant nuclides). The limit on DOSE EQUIVALENT 1-131 ensures the 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months thyroid dose to an individual at the site boundary during the Design Basis Accident (DBA) will be a small fraction of the allowed thyroid dose.

The limit on gross specific activity ensures the 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months whole body dose to an individual at the site boundary during the DBA will be a small fraction of the allowed whole body dose.

The SGTR accident analysis (Ref. 2) shows that the 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months site boundary dose levels are within acceptable limits. Violation of the LCO nay result 3 in reactor coolant radioactivity levels that could, in the event of an SGTR, lead to site boundary doses that exceed the 10 CFR 100 dose guideline limits. fsuch that the RCS specific activity is greater than the analysis assumptions, BWOG STS B 3.4.16-2 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 0, Page 381 of 415

Attachment 1, Volume 9, Rev. 0, Page 382 of 415 RCS Specific Activity B 3.4.16 BASES APPLICABILITY In MODES 1 and 2, and in MODE 3 with RCS average temperature F, operation within the LCO limits for DOSE EQUIVALENT 1-131 Landgross specific activity are necessary to contain the potential consequences of an SGTR to within the acceptable site boundary dose values .

For operation in MODE 3 with RCS average temperature ma F, and in 0 3 MODES 4 and 5, the release of radioactivity in the event of an SGTR is unlikely since the saturation pressure of the reactor coolant is below the lift pressure settings of the atmos safety valves.

m valves and main steam 0 ACTIONS A.1 and A.2 With the DOSE EQUIVALENT 1-131 greater than the LCO limit, samples at intervals of 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months must be taken to d t te limits of astortpicatanalysisaefams Figure 3.4.16-1 are not exceeded) The Completion Time of 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months is GIG be performed for at least required to obtain and analyze a sample. Sampling must continue for 1-131, 1-133, and 1-135. trending.

The DOSE EQUIVALENT 1-131 must be restored to limits within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months . The Completion Time of 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months is required, if the limit violation resulted from normal iodine spiking.

A Note permits the use of the provisions of LCO 3.0.4.c. This allowance permits entry into the applicable MODE(S) while relying on the ACTIONS.

This allowance is acceptable due to the significant conservatism incorporated into the specific activity limit, the low probability of an event which is limiting due to exceeding this limit, and the ability to restore transient specific activity excursions while the plant remains at, or proceeds to power operation.

B.1 If a Required Action and associated Completion Time of Condition A are

,*_ not met_] if the DOSE EQUIVALENT 1-131 is in the unacceptable region of Figure 3.4.16-1 ,the reactor must be brought to MODE 3 with RCS 0

or ifthe gross specific average temperature < ' F within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . The Completion Time of activity is not within limit, 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months is required to get to MODE 3 below °F without challenging reactor emergency systems.

530f BWOG STS B 3.4.16-3 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 0, Page 382 of 415

Attachment 1, Volume 9, Rev. 0, Page 383 of 415 RCS Specific Activity B 34.16 BASES ACTIONS (continued)

C.1 With the gross ecific activity in excess of the allowed limit, the unit must be placed in a MO in which the requirement does not apply.

The allowed Completion T-e of 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months to reach MODE 3 and RCS average temperature < 500 lowers the saturation pressure of the reactor coolant below the setpo ts of the main steam safety valves, and prevents venting the SG to the en vnment in an SGTR event. The Completion Time of 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months is require to reach MODE 3 from full power conditions in an orderly manner and with t challenging reactor emergency systems.

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

Trending the results of this Surveillance allows proper remedial action to be taken before reaching the LCO limit under normal operating conditions. The Surveillance is applicable in MODES 1 and 2, and in MODE 3 with RCS average temperature at least fZF. The 7 day Frequency considers the unlikelihood of a gross fuel failure during that time period.

SR 3.4.16.2 This Surveillance is performed in MODE 1 only to ensure the iodine This Surveillance requires the verification that the reactor coolant remains within limit during normal operation and following fast power changes when fuel failure is more apt to occur. The 14 day Frequency is 2 DOSE EQUIVALENT 1-131 specific adequate to trend changes in the iodine activity level considering gross activity is within limit. This specific activity is monitored every 7 days. The Frequency, between Surveillance is accomplished by 2 and 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months after a power change of _15% RTP within a 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months period, performing an isotopic analysis of is established because the iodine levels peak during this time following a reactor coolant sample.

fuel failure; samples at other times would provide inaccurate results.

BWOG STS B 3.4.16-4 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 0, Page 383 of 415

Attachment 1, Volume 9, Rev. 0, Page 384 of 415 RCS Specific Activity B 3.4.16 BASES SURVEILLANCE REQUIREMENTS (continued)

SR 3.4.16.3 SR 3.4-16-3 requires radiochemical analysis for E determination every 184 days with the.plant operating in MODE 1 equilibrium 0 conditions. The E determination directly relates to the LCO and is required to verify plant operation within the specific gross activity LCO limit. The analysis for E is a measurement of the average energies per disintegration for isotopes with half lives longer than 15 minutes, excluding iodines. The Frequency of.1 84 days recognizes E does not change rapidly. state is not required This SR has been modified by a Note that sampling to be L@ hperformed1 days after a minimum of 2 EFPD and 20 days of MODE 1 2 operation have elapsed since the reactor was last subcritical for at least 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months . This ensures the radioactive materials are at equilibrium so the analysis for E is representative and not skewed by a crud burst or other similar abnormal event.

REFERENCES 1. 10 CFR 100.11. 15.4.2 2.,FSAR, Section [16.3 304 BWOG STS B 3.4.16-5 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 0, Page 384 of 415

Attachment 1, Volume 9, Rev. 0, Page 385 of 415 JUSTIFICATION FOR DEVIATIONS ITS 3.4.16 BASES, RCS SPECIFIC ACTIVITY

1. Changes are made to be consistent with changes made to the Specification.

2. Changes are made to be consistent with the Specification.

3. Changes are made (additions, deletions, and/or changes) to the ISTS Bases which reflect the plant specific nomenclature, number, reference, system description, analysis, or licensing basis description.

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

5. Editorial change with no change in intent.

Davis-Besse Page 1 of 1 Attachment 1, Volume 9, Rev. 0, Page 385 of 415

Attachment 1, Volume 9, Rev. 0, Page 386 of 415 Specific No Significant Hazards Considerations (NSHCs)

Attachment 1, Volume 9, Rev. 0, Page 386 of 415

Attachment 1, Volume 9, Rev. 0, Page 387 of 415 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.4.16, RCS SPECIFIC ACTIVITY There are no specific NSHC discussions for this Specification.

Davis-Besse Page 1 of 1 Attachment 1, Volume 9, Rev. 0, Page 387 of 415

Attachment 1, Volume 9, Rev. 0, Page 388 of 415 ATTACHMENT 17 ITS 3.4.17, STEAM GENERATOR (SG) TUBE INTEGRITY Attachment 1, Volume 9, Rev. 0, Page 388 of 415

, Volume 9, Rev. 0, Page 389 of 415 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) , Volume 9, Rev. 0, Page 389 of 415

Attachment 1, Volume 9, Rev. 0, Page 390 of 415 ITS 3.4.17 ITS REACTOR COOLANT SYSTEM 3/4.4.5 STEAM GENERATOR (SG) TUBE INTEGRITY LIMITING CONDITION FOR OPERATION LCO 3.4.17 3.4.5 a. SG tube integrity shall be maintained, and

b. All SG tubes satisfying the tube repair criteria shall be plugged or repaired in accordance with the Steam Generator Program.

APPLICABILITY: MODES 1,2, 3, and 4.

ACTION:

ACTIONS NOTE Note: These ACTIONS may be entered separately for each SO tube.

ACTION A a. With one or more SG tubes satisfying the tube repair criteria and not plugged or repaired in accordance with the Steam Generator Program, ACTION A 1. [Within 7 days, verify tube integrity of the affected tube(s) is maintained until the

[Dsxt refueling outage or SG tube inspection, orr be in HOT STANDBY within the A

ACTION B Enext 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months , and W ACTION A 2. Plug or repair the affected tube(s) in accordance with the Steam Generator Program prior to entering HOT SHUTDOWN following the next refueling outage or SO tube inspection.

ACTION B b. With SO tube integrity not maintained, be in HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS SR 3.4.17.1 4.4.5.1 Verify SO tube integrity in accordance with the Steam Generator Program.

SR 3.4.17.2 4.4.5.2 Verify that each inspected SO tube that satisfies the tube repair criteria is plugged or repaired in accordance with the Steam Generator Program prior to entering HOT SHUTDOWN following a SG tube inspection.

DAVIS-BESSE, UNIT I 3/4 4-6 Amendment No.-8-2-7;62-(next page is 3/4 4-13) -- 1-t4;4-3r-tI-,-1-84492r229,-

22-2-2, -2627-276 Page 1 of 1 Attachment 1, Volume 9, Rev. 0, Page 390 of 415

Attachment 1, Volume 9, Rev. 0, Page 391 of 415 DISCUSSION OF CHANGES ITS 3.4.17, STEAM GENERATOR (SG) TUBE INTEGRITY 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.

MORE RESTRICTIVE CHANGES None RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES None LESS RESTRICTIVE CHANGES None Davis-Besse Page 1 of 1 Attachment 1, Volume 9, Rev. 0, Page 391 of 415

Attachment 1, Volume 9, Rev. 0, Page 392 of 415 Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs)

Attachment 1, Volume 9, Rev. 0, Page 392 of 415

Attachment 1, Volume 9, Rev. 0, Page 393 of 415 CTS SG Tube Integrity 3.4.17 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.17 :Steam Generator (SG) Tube Integrity 3.4.5 LCO 3.4.17 SG tube integrity shall be maintained.

AND All SG tubes satisfying the tube repair criteria shall be plugged[JPor repairedJin accordance with the Steam Generator Program. 0 APPLICABILITY: MODES 1, 2, 3, and 4.

ACTIONS

... .. . . ............ . ..... .. . ........ . .. . .... . ..- N r!- ---- -.

Separate Condition entry is allowed for each SG tube.

CONDITION REQUIRED ACTION COMPLETION TIME Actions a. 1 A. One or more SG tubes A.1 Verify tube integrity of the 7 days and a.2 satisfying the tube repair affected tube(s) is criteria and not plugged maintained until the next

Or repairedf]in accordance with the refueling outage or SG tube inspection.

0 Steam Generator Program. AND A.2 Plug or repairoJthe affected Prior to entering 0

tube (s) in accordance with MODE 4 following the the Steam Generator next refueling outage Program. or SG tube inspection Actions a.1 B. Required Action and B.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and b. associated Completion Time of Condition A not AND met.

B.2 Be in MODE 5. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months OR SG tube integrity not maintained.

BVV/OG STS 3.4.17-1 Rev. 3.1, 12/01/05 Attachment 1, Volume 9, Rev. 0, Page 393 of 415

Attachment 1, Volume 9, Rev. 0, Page 394 of 415 CTS SG Tube Integrity 3.4.17 SURVEILLANCE:REQUIREMENTS SURVEILLANCE FREQUENCY 4.4.5.1 SR 3.4.17.1 Verify SG tube integrity in accordance with the :In accordance Steam Generator Program. with the Steam Generator Program 4.4.5.2 SR 3.4.17.2 Verify that each inspected SG tube that satisfiesthe Prior to entering tube repair criteria is plugged Mor repairedflin accordance with the Steam Generator Program.

MODE 4 following a SG tube 0

inspection BWOG STS 3.4.17-2 Rev. 3.1, 12/01/05 Attachment 1, Volume 9, Rev. 0, Page 394 of 415

Attachment 1, Volume 9, Rev. 0, Page 395 of 415 JUSTIFICATION FOR DEVIATIONS ITS 3.4.17, STEAM GENERATOR (SG) TUBE INTEGRITY

1. The brackets have been removed and the proper plant specific information/value is provided.

Davis-Besse Page 1 of 1 Attachment 1, Volume 9, Rev. 0, Page 395 of 415

Attachment 1, Volume 9, Rev. 0, Page 396 of 415 Improved Standard Technical Specifications (ISTS) Bases Markup and Justification for Deviations (JFDs)

Attachment 1, Volume 9, Rev. 0, Page 396 of 415

Attachment 1, Volume 9, Rev. 0, Page 397 of 415 SG Tube Integrity B 3.4.17 B 3.4 REACTOR COOLANT SYSTEM (RCS)

B 3.4.17 Steam Generator (SG) Tube Integrity BASES BACKGROUND Steam generator (SG) tubes are small diameter, thin walled tubes that carry primary coolant through the primary to secondary heat exchangers.

The SG tubes have a number of important safety functions. Steam generator tubes are an integral part of the reactor coolant pressure boundary (RCPB) and, as such, are relied on to maintain the primary system's pressure and inventory. The SG tubes isolate the radioactive fission products in the primary coolant from the secondary system. In addition, as part of the RCPB, the SG tubes are unique in that they act as the heat transfer surface between the primary and secondary systems to remove heat from the primary system. This Specification addresses only the RCPB integrity function of the SG. The SG heat removal function is addressed by LCO 3.4.4, "RCS Loops - MODES 1 and 2," LCO 3.4.5, "RCS Loops - MODE 3," LCO 3.4.6, "RCS Loops - MODE 4," and LCO 3.4.7, "RCS Loops - MODE 5, Loops Filled."

SG tube integrity means that the tubes are capable of performing their intended RCPB safety function consistent with the licensing basis, including applicable regulatory requirements.

Steam generator tubing is subject to a variety of degradation mechanisms. Steam generator tubes may experience tube degradation related to corrosion phenomena, such as wastage, pitting, intergranular attack, and stress corrosion cracking, along with other mechanically induced phenomena such as denting and wear. These degradation mechanisms can impair tube integrity if they are not managed effectively.

The SG performance criteria are used to manage SG tube degradation.

Specification 5.5. "Steam Generator (SG) Program," requires that a program be established and implemented to ensure that SG tube integrity is maintained. Pursuant to Specification 5.5.. tube integrity is maintained when the SG performance criteria are met. There are three SG performance criteria: structural integrity, accident induced leakage, and operational LEAKAGE. The SG performance criteria are described in Specification 5.5M Meeting the SG performance criteria provides I reasonable assurance of maintaining tube integrity at normal and accident conditions.

The processes used to meet the SG performance criteria are defined by the Steam Generator Program Guidelines (Ref. 1).

BWOG STS B 3.4.17-1 Rev. 3.1, 12/01/05 Attachment 1, Volume 9, Rev. 0, Page 397 of 415

Attachment 1, Volume 9, Rev. 0, Page 398 of 415 SG Tube Integrity B 3.4.17 BASES APPLICABLE The steam generator tube rupture (SGTR) accident is the limiting design SAFETY basis event for SG tubes and avoiding an SGTR is the basis for this ANALYSES Specification. The analysis of a SGTR event assumes a bounding primary to secondary LEAKAGE rate equal to the operational LEAKAGE rate limits in LCO 3.4.13, "RCS Operational LEAKAGE," plus the leakage rate associated with a double-ended rupture of a single tube. The accident analysis for a SGTR assumes the contaminated secondary fluid is n released to the atmosphere via ,_afety valves * (1

)majority j discharged/to the mair/tondense main steam The analysis for design basis accidents and transients other than a SGTR assume the SG tubes retain their structural integrity (i.e., they are assumed not to rupture.) In these analyses, the steam discharge to the atmosphere is based on the total primary to secondary LEAKAGE from all SGs of P gallon per minute r is assyT-ed to increase to (1 gallon per minutel a/a result of accident indu d conditions, I I-or a gidents that do 0 0 not involvael damage, the primary coolant activit evl of DOSE equivalent to 1% . EQUIVALENT 1-131 is assumed to be equal to the LCO/3.4.16, "RCS failed fuelanalysest accident inthe -Specific A vity," limits. For accidents that assume f I damage, the ses) primary cgolan activity is a function of the amount ý('activity releasedl tfrom the lamaged G fuel The dose consequences of thesf 3re NrC within the limits of GDC 19 (Ref. 2), 10 CFR 100 (Ref. 3) or the NRC approved licensing basis (e.g., a small fraction of these limits).

Steam generator tube integrity satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii).

LCO The LCO requires that SG tube integrity be maintained. The LCO also requires that all SG tubes that satisfy the repair criteria be pluggedmor repaired]in accordance with the Steam Generator Program.

During an SG inspection, any inspected tube that satisfies the Steam Generator Program repair criteria isaepaired or]removed from service by 0 plugging. If a tube was determined to satisfy the repair criteria but was not plugged[or repairedE the tube may still have tube integrity. 0 In the context of this Specification, a SG tube is defined as the entire length of the tube, including the tube wall[and any repairs made to ito 0 between the tube-to-tubesheet weld at the tube inlet and the tube-to-tubesheet weld at the tube outlet. The tube-to-tubesheet weld is not considered part of the tube.

A SG tube has tube integrity when it satisfies the SG performance criteria The SG performance criteria are defined in Specification 5.5. e 0

jGeneraýProgram,"and describe acceptable SG tube performance.

The Steam Generator Program also provides the evaluation process for determining conformance with the SG performance criteria.

0 BXAOG STS B 3.4.17-2 Rev. 3.1, 12/01/05 Attachment 1, Volume 9, Rev. 0, Page 398 of 415

Attachment 1, Volume 9, Rev. 0, Page 399 of 415

SG Tube Integrity B 3.4.17 BASES LCO (continued)

There are three SG performance criteria: structural integrity, accident induced leakage, and operational LEAKAGE. Failure to meet any one of these criteria is considered failure to meet the LCO.

The structural integrity performance criterion provides a margin of safety against tube burst or collapse under normal and accident conditions, and ensures structural integrity of the SG tubes under all anticipated transients included in the design specification. Tube burst is defined as, "The gross structural failure of the tube wall. The condition typically corresponds to an unstable opening displacement (e.g., opening area increased in response to constant pressure) accompanied by ductile (plastic) tearing of the tube material at the ends of the degradation." Tube collapse is defined as, "For the load displacement curve for a given structure, collapse occurs at the top of the load versus displacement curve where the slope of the curve becomes zero." The structural integrity performance criterion provides guidance on assessing loads that have a significant effect on burst or collapse. In that context, the term "significant" is defined as "An accident loading condition other than differential pressure is considered significant when the addition of such loads in the assessment of the structural integrity performance criterion could cause a lower structural limit or limiting burst/collapse condition to be established." For tube integrity evaluations, except for circumferential degradation, axial thermal loads are classified as secondary loads. For circumferential degradation, the classification of axial thermal loads as primary or secondary loads will be evaluated on a case-by-case basis.

The division between primary and secondary classifications will be based on detailed analysis and/or testing.

Structural integrity requires that the primary membrane stress intensity in a tube not exceed the yield strength for all ASME Code,Section III, Service Level A (normal operating conditions) and Service Level B (upset or abnormal conditions) transients included in the design specification.

This includes safety factors and applicable design basis loads based on ASME Code,Section III, Subsection NB (Ref. 4) and Draft Regulatory Guide 1.121 (Ref. 5).

The accident induced leakage performance criterion ensures that the primary to secondary LEAKAGE caused by a design basis accident, other than a SGTR, is within the accidentanalysis assumptions. The accident analysis assumes that accident induced leakage does not exceed 01 gpm prS exc/pt for Specific types o 'egradlation at spec!ýc Iocatlon_

ereheNRC has aproved r ater accident induce5leakag f. The accident induced leakage rate includes any primary to secondary LEAKAGE existing prior to the accident in addition to primary to secondary LEAKAGE induced during the accident.

BWOG STS B 3.4.17-3 Rev. 3.1, 12/01/05 Attachment 1, Volume 9, Rev. 0, Page 399 of 415

Attachment 1, Volume 9, Rev. 0, Page 400 of 415 SG Tube Integrity B 3.4.17 BASES LCO (continued)

The operational LEAKAGE performance criterion provides an observable indication of SG tube conditions during plant operation. The limit on operational LEAKAGE is contained in LCO 3.4.13, "RCS Operational LEAKAGE," and limits primary to secondary LEAKAGE through any one SG to 150 gallons per day. This limit is based on the assumption that a single crack leaking this amount would not propagate to a SGTR under the stress conditions of a LOCA or a main steam line break. If this amount of LEAKAGE is due to more than one crack, the cracks are very small, and the above assumption is conservative.

APPLICABILITY Steam generator tube integrity is challenged when the pressure differential across the tubes is large. Large differential pressures across SG tubes can only be experienced in MODE 1, 2, 3, or 4.

RCS conditions are far less challenging in MODES 5 and 6 than during MODES 1, 2, 3, and 4. In MODES 5 and 6, primary to secondary differential pressure is low, resulting in lower stresses and reduced potential for LEAKAGE.

ACTIONS The ACTIONS are modified by a Note clarifying that the Conditions may be entered independently for each SG tube. This is acceptable because the Required Actions provide appropriate compensatory actions for each affected SG tube. Complying with the Required Actions may allow for continued operation, and subsequent affected SG tubes are governed by subsequent Condition entry and application of associated Required Actions.

A.1 and A.2 Condition A applies if it is discovered that one or more SG tubes examined in an inservice inspection satisfy the tube repair criteria but were not plugged Tor repaired in accordance with the Steam Generator Program as required by SR 3.t4.17.2. An evaluation of SG tube integrity 0 of the affected tube(s) must be made. Steam generator tube integrity is based on meeting the SG performance criteria described in the Steam Generator Program. The SG repair criteria define limits on SG tube degradation that allow for flaw growth between inspections while still providing assurance that the SG performance criteria will continue to be met. In order to determine if a SG tube that should have been plugged Nor repaired]has tube integrity, an evaluation must be completed that demonstrates that the SG performance criteria will continue to be met 0

until the next refueling outage or SG tube inspection. The tube integrity BWOG STS 83.4.17-4 Rev. 3.1, 12/01/05 Attachment 1, Volume 9, Rev. 0, Page 400 of 415

Attachment 1, Volume 9, Rev. 0, Page 401 of 415 SG Tube Integrity B 3.4.17 BASES ACTIONS (continued) determination is based on the estimated condition of the tube at.the time the situation is discovered and the estimated growth of the degradation prior to the next SG tube inspection. If it is determined that tube integrity is not being maintained, Condition B applies.

A Completion Time of 7 days is sufficient to complete the evaluation while minimizing the risk of plant operation with a SG tube that may not have tube integrity.

Ifthe evaluation determines that the affected tube(s) have tube integrity, Required Action A.2 allows plant operation to continue until the next refueling outage or SG inspection provided the inspection interval continues to be supported by an operational assessment that reflects the affected tubes. However, the affected tube(s) must be plugged [Pr (->

repaired]prior to entering MODE 4 following the next refueling outage or SG inspection. This Completion Time is acceptable since operation until the next inspection is supported by the operational assessment.

B.1 and B.2 Ifthe Required Actions and associated Completion Times of Condition A are not met or if SG tube integrity is not being maintained, the reactor must be brought to MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and MODE 5 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

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

SURVEILLANCE SR 3.4.17.1 REQUIREMENTS During shutdown periods the SGs are inspected as required by this SR and the Steam Generator Program. NEI 97-06, Steam Generator Program Guidelines (Ref. 1), and its referenced EPRI Guidelines, establish the content of the Steam Generator Program. Use of the Steam Generator Program ensures that the inspection is appropriate and consistent with accepted industry practices.

During SG inspections a condition monitoring assessment of the SG tubes is performed. The condition monitoring assessment determines the "as found" condition of the SG tubes. The purpose of the condition monitoring assessment is to ensure that the SG performance criteria have been met for the previous operating period.

BWOG STS B 3.4.17-5 Rev. 3.1, 12/01/05 Attachment 1, Volume 9, Rev. 0, Page 401 of 415

Attachment 1, Volume 9, Rev. 0, Page 402 of 415 SG Tube Integrity B 3.4.17 BASES SURVEILLANCE REQUIREMENTS (continued)

The Steam Generator Program determines the scope of the inspection and the methods used to determine whetherthe tubes contain flaws satisfying the tube repair criteria. Inspection scope (i.e., which tubes or areas of tubing within the SG are to be inspected) is a function of existing and potential degradation locations. The Steam Generator Program also specifies the inspection methods to be used to find potential degradation.

Inspection methods are a function of degradation morphology, non-destructive examination (NDE) technique capabilities, and inspection locations.

The Steam Generator Program defines the Frequency of SR 3.4.17.1.

The Frequency is determined by the operational assessment and other limits in the SG examination guidelines (Ref. 6). The Steam Generator Program uses information on existing degradations and growth rates to determine an inspection Frequency that provides reasonable assurance that the tubing will meet the SG performance criteria at the next scheduled inspection. In addition, Specification 5.5.1fntns prescriptive requirements concerning inspection intervals to provide added assurance that the SG performance criteria will be met between scheduled inspections.

SR 3.4.17.2 During an SG inspection, any inspected tube that satisfies the Steam Generator Program repair criteria isgepaired orjremoved from service by plugging. The tube repair criteria delineated in Specification 5.5SraTre ( j) intended to ensure that tubes accepted for continued service satisfy the SG performance criteria with allowance for error in the flaw size measurement and for future flaw growth. In addition, the tube repair criteria, in conjunction with other elements of the Steam Generator Program, ensure that the SG performance criteria will continue to be met until the next inspection of the subject tube(s). Reference 1 provides guidance for performing operational assessments to verify that the tubes remaining in service will continue to meet the SG performance criteria.

[team generator tube repairs are only performed using approved repair methods as described in the Steam Generator Program.[]

The Frequency of prior to entering MODE 4 following a SG inspection ensures that the Surveillance has been completed and all tubes meeting the repair criteria are plugged[Tor repaired]prior to subjecting the SG tubes to significant primary to secondary pressure differential.

0 BWOG STS B 3.4.17-6 Rev. 3.1, 12/01/05 Attachment 1, Volume 9, Rev. 0, Page 402 of 415

Attachment 1, Volume 9, Rev. 0, Page 403 of 415 SG Tube Integrity B 3.4.17 BASES REFERENCES 1. NEI 97-06, "Steam Generator Program.Guidelines."

2. 10 CFR 50 Appendix A, GDC 19.

3. 10CFR 100.

4. ASME Boiler and Pressure Vessel Code, Section Il1,Subsection NB.

5. Draft Regulatory Guide 1.121, "Basis for Plugging Degraded Steam Generator Tubes," August 1976.

6. EPRI, "Pressurized Water Reactor Steam Generator Examination Guidelines."

Regu r7. Guide 1.83, "Inservice Inspection f Pressurized atorSy Water Reactor Steam Generator Tubes," July 1975. 0 BVVOG STS B 3.4.17-7 Rev. 3.1, 12/01/05 Attachment 1, Volume 9, Rev. 0, Page 403 of 415

Attachment 1, Volume 9, Rev. 0, Page 404 of 415 JUSTIFICATION FOR DEVIATIONS ITS 3.4.17 BASES, STEAM GENERATOR (SG) TUBE INTEGRITY

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

2. Changes are made (additions, deletions, and/or changes) to the ISTS Bases which reflect the plant specific nomenclature, number, reference, system description, analysis, or licensing basis description.

3. Updated references with valid and committed Regulatory Guide.

4. Editorial change. The title of the SG Program has already been defined in the Bases.

5. The correct LCO number has been provided.

Davis-Besse Page 1 of 1 Attachment 1, Volume 9, Rev. 0, Page 404 of 415

Attachment 1, Volume 9, Rev. 0, Page 405 of 415 Specific No Significant Hazards Considerations (NSHCs)

Attachment 1, Volume 9, Rev. 0, Page 405 of 415

Attachment 1, Volume 9, Rev. 0, Page 406 of 415 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.4.17, STEAM GENERATOR (SG) TUBE INTEGRITY There are no specific NSHC discussions for this Specification.

Davis-Besse Page 1 of 1 Attachment 1, Volume 9, Rev. 0, Page 406 of 415

, Volume 9, Rev. 0, Page 407 of 415 ATTACHMENT 18 Relocated Current Technical Specifications , Volume 9, Rev. 0, Page 407 of 415

Attachment 1, Volume 9, Rev. 0, Page 408 of 415 CTS 3/4.4.10.1, ASME CODE CLASS 1, 2, AND 3 COMPONENTS Attachment 1, Volume 9, Rev. 0, Page 408 of 415

, Volume 9, Rev. 0, Page 409 of 415 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) , Volume 9, Rev. 0, Page 409 of 415

Attachment 1, Volume 9, Rev. 0, Page 410 of 415 CTS 3/4.4.10.1 I

REACTOR COOLANT SYSTEM 3.4. 0 S TRUCTURAL INTEG ASME CODE CLASS 1, 2 and 3 OMPONENTS JMTINm CONDITION FO 0 RATION 3.4.10.1 The structural integrity f ASME Code Class 1, 2 and 3 components s all be maintained in accordance with Specification 4. .10.1.

APPLICABILITY: All MODES.

ACTION;

a. With the structural inte rity of any ASME Code Class I component(s) not conforming to the above requirements. re tore the structural integrity of the affected co ponent(s) to within its limit or isolate the d component(s) prior to increasing the Reac r Coolant System temperature more than150 °F the minimum temperature requi above by NDT consideratons.

b. With the structural in grity of any ASME Code Class 2 componenE( ) not conforming to the above requirements, estore the structural integrity of the affected c mponent(s) to within its --- a limit or isolate the af ected component(s) prior to increasing the R tor Coolant System temperature above 2 *F.

c. With the structural i tegrity of any ASME Code Class 3 compone (s) not conforming to the above requirements restore the structural integrity of the compon t(s) to within its limit or isolate the affected omponent(s) from service.

d. The provisions of pecification 3.0.4 are not applicable.

SURVEILLANCE RE _REMENTS 4.4.10.1 In addition to hrequirements of Specification 4.0.5:

a. Inservice inspection of each reactor coolant pump flywheel shall be performed at least once every 10 years. The inservice inspection shall be either an ultrasonic examination of the volume from the inner bore of the flywheel to the circle of one-half the outer radius, or a S.. ITS surface examination of exposed surfaces of the disassembled flywheel. The recommendations delineated in Regulatory Guide 1. 14, Revision 1, August 1975, Positions 3, 4 and 5 of Section C.4.b shall apply.

DAVIS-BESSE, UNIT I 3/4 4-30 Amendment No. 232 Page 1 of 2 Attachment 1, Volume 9, Rev. 0, Page 410 of 415

Attachment 1, Volume 9, Rev. 0, Page 411 of 415 CTS 3/4.4.10.1 REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued)

b. Each internals vent valve hall be demonstrated OPERABLE at least onc per 24 months* during shutdo 'it by:

I. Verifýing through vis al inspection that the valve body and valve di c exhibit no abnormal degradatior,

2. Verifying the valve i not stuck in an open position, and

3. Verifying through nual actuation that the valve is fully open wh n a force of

< 400 lbs. is applied vertically upward.

See ITS]

5.5]

An exception app ies for the interval following the March 2003 erification completed during the Thirte nth Refueling Outage. Under this exception, t e next performance of this surveillance equirement may be delayed until March 25, 2 6.

DAVIS-BESSE, UNIT I 314 4-31 Amendment No. 23, 95, 165, 268 Page 2 of 2 Attachment 1, Volume 9, Rev. 0, Page 411 of 415

Attachment 1, Volume 9, Rev. 0, Page 412 of 415 DISCUSSION OF CHANGES CTS 3/4.4.10.1, ASME CODE CLASS 1, 2, AND 3 COMPONENTS ADMINISTRATIVE CHANGES None MORE RESTRICTIVE CHANGES None RELOCATED SPECIFICATIONS R01 CTS 3/4.4.10.1 provides requirements for the ASME Code Class 1, 2 and 3 components to ensure their structural integrity. The inspection programs for ASME Code Class 1, 2 and 3 components ensure that the structural integrity of these components will be maintained throughout the life of the components.

ASME Code Class 1, 2, and 3 components are monitored so that the possibility of component structural failure does not degrade the safety function of the system. The monitoring activity is of a preventive nature rather than a mitigative action. Other Technical Specifications require important systems to be OPERABLE (for example, Emergency Core Cooling Systems) and in a ready state for mitigative action. This Technical Specification is more directed toward prevention of component degradation and continued long term maintenance of acceptable structural conditions. Hence, it is not necessary to retain this Specification to ensure immediate OPERABILITY of safety systems. Further, this Technical Specification prescribes inspection requirements that are performed during plant shutdown. It is, therefore, not directly important for responding to design basis accidents. This LCO does not meet the criteria for retention in the ITS; therefore, it will be retained in the Technical Requirements Manual (TRM).

This change is acceptable because CTS 3/4.4.10.1 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. The programmatic inspections stipulated by this Specification are not installed instrumentation used for, nor capable of, detecting a significant abnormal degradation of the reactor coolant pressure boundary during operations prior to a design basis accident (DBA). The ASME Code Class 1, 2 and 3 Components Specification does not satisfy criterion 1.

2. The programmatic inspections stipulated by this Specification are not a process variable, design feature, or operating restriction that is an initial assumption in a DBA or transient. The ASME Code Class 1, 2 and 3 Components Specification does not satisfy criterion 2.

3. The ASME Code Class 1, 2, and 3 components inspected per this Specification are assumed to function to mitigate a DBA. Their capability to perform this function is addressed by other Technical Specifications.

This Technical Specification only specifies programmatic inspection Davis-Besse Page 1 of 2 Attachment 1, Volume 9, Rev. 0, Page 412 of 415

Attachment 1, Volume 9, Rev. 0, Page 413 of 415 DISCUSSION OF CHANGES CTS 3/4.4.10.1, ASME CODE CLASS 1, 2, AND 3 COMPONENTS requirements for these components, and these inspections can only be performed when the plant is shutdown. Therefore, criterion 3 is not satisfied.

4. As discussed in B&W Owners Group Technical Report 47-1170689-00 (Appendix A pages A-63 and A-64), the assurance of operability of the entire system as verified in the system operability Specification dominates the risk contribution of the system. The lack of a long term assurance of structural integrity as stipulated by this Specification 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.

The ASME Code Class 1, 2 and 3 Components Specification does not meet criterion 4.

Since the 10 CFR 50.36(c)(2)(ii) criteria have not been met, the ASME Code Class 1, 2 and 3 Components LCO and associated Surveillances may be relocated out of the Technical Specifications. The ASME Code Class 1, 2 and 3 Components Specification 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. In addition, Surveillances, except for the reactor coolant pump (RCP) flywheel inspection and the internal vent valve requirements, are already required by regulations in 10 CFR 50.55a to be performed in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable addenda. The RCP flywheel inspection requirement and the internal vent valve requirements are not covered by other regulatory requirements and are needed for safe operation of the plant; therefore, these requirements will be maintained in the Davis-Besse Improved Technical Specifications. Chapter 5.0 of the Davis-Besse Improved Technical Specifications will contain a section which provides a programmatic approach to the requirements relating to the structural integrity of ASME Code Class 1, 2, and 3 components. This change is designated as relocation because the Specification did not meet the criteria in 10 CFR 50.36(c)(2)(ii) and has been relocated to the TRM.

REMOVED DETAIL CHANGES None LESS RESTRICTIVE CHANGES None Davis-Besse Page 2 of 2 Attachment 1, Volume 9, Rev. 0, Page 413 of 415

Attachment 1, Volume 9, Rev. 0, Page 414 of 415 Specific No Significant Hazards Considerations (NSHCs)

Attachment 1, Volume 9, Rev. 0, Page 414 of 415

Attachment 1, Volume 9, Rev. 0, Page 415 of 415 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS CTS 3/4.4.10.1, ASME CODE CLASS 1, 2, AND 3 COMPONENTS There are no specific NSHC discussions for this Specification.

Davis-Besse Page 1 of 1 Attachment 1, Volume 9, Rev. 0, Page 415 of 415

ML072200472

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