Text

Improved Technical Specification Conversion License Amendment Request, Volume 9, (Revision 1), Section 3.4 - Reactor Coolant System (RCS)
	ML081480468
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Site:	Davis Besse
Issue date:	05/16/2008
From:	; FirstEnergy Nuclear Operating Co
To:	; Office of Nuclear Reactor Regulation
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'ookalýRV qt'm DAVIS-BESSE NUCLEAR POWER STATION UNIT 1 IMPROVED TECHNICAL SPECIFICATION CONVERSION LICENSE AMENDMENT REQUEST I-.

VOLUME 9 (Rev. 1)

SECTION 3.4 - REACTOR COOLANT SYSTEM (RCS)

Attachment 1, Volume 9, Rev. 1, Page i of Hi Summary of Changes

.ITS Section 3.4 Change Description Affected Pages The changes described in the Davis-Besse Pages 5, 9, 13, 14, and 21 response to Question 200712031021 have been made. This changes the time allowed to stabilize conditions and perform SIR 3.4.1.4 (24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months from 7 days).

The changes described in the Davis-Besse Pages 43, 49, and 51 response to Question 200710041523 have been made. This reclassifies an administrative DOC to a less restrictive DOC.

The changes described in the Davis-Besse Pages 70, 74, 75, 76, 81, and 82 response to Question 200710011404 have been made. This change adds a default Condition for ITS 3.4.4 Condition A inadvertently left out.

The changes described in the Davis-Besse Pages 89, 110, 118, 119, 120, 122, 123, 134, 138, response to Question 200803111017 have been 142, 143, 145, 149, and 150 made. This change adds back in the LCO NOTE for ITS 3.4.6 and 3.4.7.

The changes described in the Davis-Besse Pages 189 and 190 response to Question 200710231028 have been made. An editorial correction has been made to a JFD.

The changes described in the Davis-Besse Pages 184, 185, 188, 189, 193, and 194 response to Question 200801140737 have been made. The minimum pressurizer heater power has been changed from 150 kW to 85 kW.

The changes described in the Davis-Besse Page 220 response to Question 200710231302 have been made. An editorial change has been made to a CTS Markup.

The changes described in the Davis-Besse Page 230 response to Question 200712271037 have been made. An editorial change to a Bases Insert has been made.

The changes described in the Davis-Besse Pages 240, 243, and 244 response to Question 200710231615 have been made. This change adds a new M DOG to ITS 3.4.12 to justify the additional of a default compensatory Action for the DHR System isolation valves The changes described in the Davis-Besse Pages 302, 303, and 309 response to Question 200712181530 have been made. A new Administrative DOC has been added to ITS 3.4.14 to justify the deletion of a CTS 3.0.4 exception not needed in the ITS.

The changes described in the Davis-Besse Pages 316, 324, and 331 response to Question 200712261451 have been made. The Applicability of ITS 3.4.14 has been clarified to exclude the DHR System interlock Function in MODE 4, consistent with CTS.

Page 1 of 2 Attachment 1, Volume 9, Rev. 1, Page i of ii

Attachment 1, Volume 9, Rev. 1, Page ii of ii Summary of Changes ITS Section 3.4 Change Description Affected Pages The changes described in the Davis-Besse Pages 406 and 407 response to Question 200712271033 have been made. The ITS 3.4.17 Bases have been changed to delete an unused Reference.

0 Page 2 of 2 Attachment 1, Volume 9, Rev. 1, Page ii of ii

Attachment 1, Volume 9, Rev. 1, Page 1 of 418 ATTACHMENT 1 VOLUME 9 DAVIS-BESSE IMPROVED TECHNICAL SPECIFICATIONS CONVERSION ITS SECTION 3.4 REACTOR COOLANT SYSTEM (RCS)

Revision 1 Attachment 1, Volume 9, Rev. 1, Page 1 of 418

Attachment 1, Volume 9, Rev. 1, Page 2 of 418 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. 1, Page 2 of 418

Attachment 1, Volume 9, Rev. 1, Page 3 of 418

ATTACHMENT 1 ITS 3.4.1, RCS PRESSURE, TEMPERATURE, AND FLOW DEPARTURE FROM NUCLEATE BOILING (DNB) LIMITS Attachment 1, Volume 9, Rev. 1, Page 3 of 418

Attachment 1, Volume 9, Rev. 1, Page 4 of 418 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) 0 Attachment 1, Volume 9, Rev. 1, Page 4 of 418

Attachment 1, Volume 9, Rev. 1, Page 5 of 418 ITS 3.4.1 ITS POWER DISTRIBUTION IRMITS DNB PARAMETERS LIMITING CONDITION FOR OPERATION LCO 3.4.1 3.2.5 The following ONB related parameters shall be maintained within the limits shown on Table 3.2-2.

a. Reactor Coolant Hot Leg Temperature.

b. Reactor Coolant Pres~sure

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

ACTION A-- If any parameter above exceeds its limit, restore the parameter to within its imit within 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months sor reduce THERMAL POWER to less than 5% of RATED THERMAL ACTION B --- ER within the next hours. A02 SURVEILLANCE REOUIREMENTS 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 .

SIR3.4.1.4 4... The Reactor Coolant System total flow rate shall be det~ermfined" to} be within its limit by measurement at least once per. 18..months..

Notrqie ob efre nil124 ho~urs after stable l thermal conditions are established at >_70% RTP. .

L0S2 DAVIS-BESSE, UNIT I 3/4 2-13 Amendment No. 64 71_2- 222 Page 1 of 2 Attachment 1, Volume 9, Rev. 1, Page 5 of 418

Attachment 1, Volume 9, Rev. 1, Page 6 of 418 iTS*1 ITS TABLE 3.2-2 a,)

(4

(.1 D"tB MARGIN LCO 3.4.1.a LCO 3.4.1.b Required Measured Required Measured Parameters with Parameters with Four Reactor Three Reactor Coolant Pumps Coolant Pumps Parameter Operat ing Opera t Inc Reactor Coolant Bot Leg LCO 3.4.1 Temperature TBSF <610 <610(1)

RLeateor Coolant ?ressare, psis.(2)

XpuI(31 2060.8 Reactor CooL~azd Flow ;*ate. >389,50 >290,957

= ::::-M01 NOTE to SR 3.4.1.1 and -[(1l)

SR 3.4.1.2a Applicable to the loop vith 2 Reactor Coolant Pumps Operating.

LCO 3.4.1 . i(2) Limit not applicable during either a THERMAL POVM ramp increase In excess of 5Z of RATED THERMAL Applicability g POZ per minute or a THERMAL POUM step increase of greater than lOZ of RATED THERMAL PlER. LA01 NoTE T s s.l - uIred measured flow, includao flow rate uncertainty of 2.51 - .tare ,e o aa nimor 52umpi Wrn epolon r as eag 2 of 2core.

A03 Page 2 of 2 Attachment 1, Volume 9, Rev. 1, Page 6 of 418

Attachment 1, Volume 9, Rev. 1, Page 7 of 418 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 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 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. 1, Page 7 of 418

Attachment 1, Volume 9, Rev. 1, Page 8 of 418 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 1 - 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. 1, Page 8 of 418

Attachment 1, Volume 9, Rev. 1, Page 9 of 418 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 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 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 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 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. 1, Page 9 of 418

Attachment 1, Volume 9, Rev. 1, Page 10 of 418 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. 1, Page 10 of 418

Attachment 1, Volume 9, Rev. 1, Page 11 of 418 Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs) 0 Attachment 1, Volume 9, Rev. 1, Page 11 of 418

Attachment 1, Volume 9, Rev. 1, Page 12 of 418 CTS RCS Pressure, Temperature, and FlowDNB 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 3.4A1 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 pumps (RCPs) operating:

2064.8 RCS loop pressure shall be > psig, RCS hot leg temperature shall be <*64.6]IF, and RCS total flow rate shall be

_____ ____ >:[139. /EG] lbfh and61 389500 p~

32

b. With three RCPs operating:

RCS loop pressure shall be " 2psig, RCS hot leg temperature shall be _ 6g4.6 F, and RCS total flow rate shall be

>1[10o4.,/E6] lb/h . .. *r *-

APPLICABILITY: MODE 1.

Table 3.2-2

,4W

--l'tJIE I----..........--------.--.

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, 03/31104 Attachment 1, Volume 9, Rev. 1, Page 12 of 418

Attachment 1, Volume 9, Rev. 1, Page 13 of 418 CTS RCS Pressure, Temperature, and Flow DNB Limits 3A4..1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.1.1 -----... -- NOTE ------.-------

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

4.2.5.1 Verify RCS loop.pressure > 1 psig with 12.hours 0 0

four, RCPs operating or > 2 psig with three RCPs operating.2 SR 3.4.1.2 -- - --.-----.--- 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 hot leg temperature 6 F. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months 0

4.2.51 SR 3.4.1.3 Verify, RCS total flow> 1139.: E6] lb/h with four 389,500 gpm 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months 0

RCPs operating or 2 1104. *E6] lb/hý with three RCPJ operating. 290,957 gpm 4.2.5.2 SR 3.4.1.4 ---------------..............------

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

Only requir d to be perfor ed when staille thermal

-~-conditions/are of MODE/1.

establishe4 in the higher ower range

/ /

0 Verifym RCStotal flow rate is within limit by P1 qmonths 0

measurement.

--. Not required to be performed until 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 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. 1, Page 13 of 418

Attachment 1, Volume 9, Rev. 1, Page 14 of 418 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 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after stable thermal conditions are established at >_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. Furthermore, the 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months allowance is consistent with both the WOG ISTS (NUREG-1431) and the CEOG ISTS (NUREG-1432).

Davis-Besse Page 1 of I Attachment 1, Volume 9, Rev. 1, Page 14 of 418

Attachment 1, Volume 9, Rev. 1, Page 15 of 418 Improved Standard Technical Specifications (ISTS) Bases Markup and Justification for Deviations (JFDs)

Attachment 1, Volume 9, Rev. 1, Page 15 of 418

Attachment 1, Volume 9, Rev. 1, Page 16 of 418 Q-I)

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

B 3.4.1 RCS PressureTemperature,.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 (DN BR) 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 andtlis abovethat used a the initial pressure in the analyses. A pressure greater than the minimum specified corresponds too will producea. higher minimum DNBR. A pressure lower than the minimum specified will cause the plant to approach the DN B limit.

The full with LCOpower for maximum operationRCS coolant within hot leg operating the nominal temperature is consistent envelope and Mj o an the 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 hot 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 flowwill cause the plant to approach the DNB 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)j LCO will meet the DNBR criterion o This is the acceptance limit for the RCS DNBR parameters. Changes to the facility that could impact 0 these parameters must be assessed for their impact on the DNBR BVVOG STS B 3.4.1-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 1, Page 16 of 418

Attachment 1, Volume 9, Rev. 1, Page 17 of 418 All changes are I 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 stuckcontrol rod events. Akey assumption for the analysis of these events is that the core power distribution is within the limits of LCO 3.2;1, "Regulatin'Rod ion Limits," LCO 3.2.3, "AXIAL POWER IMBALANCEOcPP Li" and LCO 3.2.4, "QUADRANT POWER TILTFR 10 QPT."

nominal The core outlet pressure assumed in the safety anal ses isE1jI*pia.

The minimum pressure specified in LCO 3.4.1 is the Ii it alue in the (?

reactor oolant loop as measured at'the hotleg pressure tap. corresponding '

2 The safety analyses are performed with an assumed RCS coolant average temperature 0 FF(579°F plus 2" ilwance for aulational to limit the range of allowable, steady

/state operation, consistent with the -- unc rtaint. The corre ondin ca culated by a.tuming an RCS core hot leg temperature outlet pressure of 6 Ei4.is .andn of 21/35 psia inta coniton .asu e ." in th RCS flow rate Of 374,880 gpm.I The maximum temperature specified is

.*-related accident analyses. J 'the limit value at the hot leg resistance temperature detector.

The safety analyses are performed with an assumed RCS flow rate of 3 8 gpm. The minimum flow rate specified in LCO 3.4.1 is the equivalent ].minimum mass flow rate* including a 2.5% uncertainty 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 RCSDNB limits satisfy Criterion 2 of 10 CFR 50.36(c)(2)(ii).

LCO This LCO specifieslimits.on the monitored process variables: RCS loop (hot leg) pressure, RCS hot leg temperature, and RCS total flow rate to ensure thatthe 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.

BWOG STS B 3.4.1-2 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 1, Page 17 of 418

Attachment 1, Volume 9, Rev. 1, Page 18 of 418 I All changes are .

unless otherwise noted RCS Pre~sure, 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 rateare'ý given for the measurement locationdbu ave not been adjuste for instrument error/. Plant specific limitnof~instrument error are ýtablished by the plant stiff to meet the oper tional requirements oftlg LC 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 MODESthe 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 parameterand 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. 1, Page 18 of 418

Attachment 1, Volume 9, Rev. 1, Page 19 of 418 All changes are unless otherwise noted RCS Pressure, Temperature,.and Flow. DN B 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 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 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 sa--sfety analysis i'-2s35 psi . The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months interval has~been shown by (nomial) 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 appliedto 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 hourSurveillance 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.

0 BWOG STS B 3.4.1-4 Rev. 3.0, 03131/04 Attachment 1, Volume 9, Rev. 1, Page 19 of 418

Attachment 1, Volume 9, Rev. 1, Page 20 of 418 r All changes are unless otherwise noted 9 RCS Pressure, Temperature, and Flow DNB.Limits-B 3.4.11 BASES SURVEILLANCE REQUIREMENTS (continued)

A Note has.been added toý indicate the temperature limits are to be applied to the loop with.two pumps in operation for the three pump operating condition. I 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 every1j 8]months 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. INSERT 1 The Frequency of M1 8months refle s the importance of v ifying flow Iafter a refuelin outagewh en hetre has been altered oV RCS flow 0

(INSERT 2 ý- characteristfs may have beennrdified, which may havycaused chanpe J ,INSERT 3 The Surveillance is modifiipd by a Note that indicates the SR o s not o be/prformed until stable t~ermal conditions are esta ~ishead at I ned 0

/higer~werlevels. ]TheNote 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 poweror in MODE 2 or below because at low power the AT across the core will be too small to provide valid results.

REFERENCES FSAR

2. UFSAR, Appendix 4B.

Section 000 BWOG STS B 3.4.1-5 Rev, 3.0, 03131/04, 0

Attachment 1, Volume 9, RIev. 1, Page 20 of 418

Attachment 1, Volume 9, Rev. 1, Page 21 of 418 B 3.4.1 0I 0 These minimum required measured flows include a flow rate uncertainty of 2.5%.

O 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 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after stable thermal conditions are established at >_70% RTP.

Insert Page B 3.4.1-5 Attachment 1, Volume 9, Rev. 1, Page 21 of 418

Attachment 1, Volume 9, Rev. 1, Page 22 of 418 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. 1, Page 22 of 418

Attachment 1, Volume 9, Rev. 1, Page 23 of 418 Specific No Significant Hazards Considerations (NSHCs)

Attachment 1, Volume 9, Rev. 1, Page 23 of 418

Attachment 1, Volume 9, Rev. 1, Page 24 of 418 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. 1, Page 24 of 418

Attachment 1, Volume 9, Rev. 1, Page 25 of 418 ATTACHMENT 2 ITS 3.4.2, RCS MINIMUM TEMPERATURE FOR CRITICALITY Attachment 1, Volume 9, Rev. 1, Page 25 of 418

, Volume 9, Rev. 1, Page 26 of 418 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) , Volume 9, Rev. 1, Page 26 of 418

Attachment 1, Volume 9, Rev. 1, Page 27 of 418 ITS 3.4.2 ITS REACTIVITY 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 shall

ýbe 525-F. avg APPLICABILITY: MODES I and 2*.

ACTION:

ACTION A Vith a Reactor Coolant System loop temperatu (T <, 52OF, Ei\

iTt in its lmit within e.inu es be IN T DR within SURVEILLANCE REOUIREMENTS SR 3.4.2.1 4.1.1.4 The. RCS temperature (Ta 9 ) shall be determined to be

525?F:

a. Within 15 minqu es prior to achieving actor criticality,, a d

b. At least onc per 30 minutes when t e reactor is critical and e~

the Reactor/Coolant System Tavq is ess than 5300F.

~~I

~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 I Attachment 1, Volume 9, Rev. 1, Page 27 of 418

Attachment 1, Volume 9, Rev. 1, Page 28 of 418 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 0 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 bein 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 0F, 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 keff < 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. 1, Page 28 of 418

Attachment 1, Volume 9, Rev. 1, Page 29 of 418 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 keff - 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 keff < 1.0. As a result, changing the ACTION to "be in MODE 2 with kef < 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 > 525 0 F within 15 minutes prior to achieving reactor criticality, and every 30 minutes when the reactor is critical and the RCS Tavg < 5300 F. ITS SR 3.4.2.1 requires RCS Tavg in each loop to be verified > 525°F 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 > 525 0 F, 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. 1, Page 29 of 418

Attachment 1, Volume 9, Rev. 1, Page 30 of 418 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. 1, Page 30 of 418

Attachment 1, Volume 9, Rev. 1, Page 31 of 418 Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs)

Attachment 1, Volume 9, Rev. 1, Page 31 of 418

Attachment 1, Volume 9, Rev. 1, Page 32 of 418 CTS 0 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,> 5251F.

APPLICABILITY: MODE 1, MODE 2 with keff > 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 withukff

< 1.0.

30 minutes 0

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY 4.1.1.4 SR 3.4.2.1 Verify RCS Tav, in each loop > 5251F. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months BW\/OG STS 3.4.2-1 Rev. 3.0 03/31/04 Attachment 1,'Volume 9, Rev. 1, Page 32 of 418

Attachment 1, Volume 9, Rev. 1, Page 33 of 418 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. 1, Page 33 of 418

Attachment 1, Volume 9, Rev. 1, Page 34 of 418 Improved Standard Technical Specifications (ISTS) Bases Markup and Justification for Deviations (JFDs)

Attachment 1, Volume 9, Rev. 1, Page 34 of 418

Attachment 1, Volume 9, Rev. 1, Page 35 of 418 K All changes are 1 unless otherwise noted RCS Minimum Temperature for Crit cality B 3'4.2 B 3.4 REACTOR COOLANT SYSTEM (RCS)

B 314:2 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 0

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

The reactor coolant moderator temperaturecefficient used in core operating and accident analysis is typically defined for the normal 582oeratin temperature range (532°F t A[ý7£F). 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 (Tvg) using inputs of the same range. Nominal T.,, for making the reactor critical is 532TF. ISafety and loperating analyses for lower temperatures have not been made.

_______________ .I APPLICABLE There are no accident analyses that dictate the minimum ter SAFETY for criticality ball low power sa ety a lyses assume initia ANALYSES Itemperatur near the 525*F limit (R . 11.+.*

I T he RCSrminimum temperature for criticality satisfies Criterion 2 of 10 CdFR 50.36(c)(2)(ii).

I much t--

4 LCO The purpose of the LCO is to preent ýclityoutside' the norma operating regime (532°F to 9* and to prevent operation in an unanalyzed condition.

The LCO limit of 525'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 keff ý 1.0.

BWOG STS B 3.4.2-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 1, Page 35 of 418

Attachment 1, Volume 9, Rev. 1, Page 36 of 418 B 3.4.2 0 INSERT I 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. 1, Page 36 of 418

Attachment 1, Volume 9, Rev. 1, Page 37 of 418 RCS Minimum Temperature for Criticality B 3.42 BASES ACTIONS A.j With Tavg below 525°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 ýction and maintain the plant 0

within the analyzed range. If Tavg 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 CoI1. FSAR,hSection 15.2.1 00 BWOG STS B 3.4.2-2 Rev.. 3,0, 03/31/04 Attachment 1, Volume 9, Rev. 1, Page 37 of 418

Attachment 1, Volume 9, Rev. 1, Page 38 of 418 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 I Attachment 1, Volume 9, Rev. 1, Page 38 of 418

Attachment 1, Volume 9, Rev. 1, Page 39 of 418 Specific No Significant Hazards Considerations (NSHCs)

Attachment 1, Volume 9, Rev. 1, Page 39 of 418

Attachment 1, Volume 9, Rev. 1, Page 40 of 418 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. 1, Page 40 of 418

Attachment 1, Volume 9, Rev. 1, Page 41 of 418 0 ATTACHMENT 3 ITS 3.4.3, RCS PRESSURE AND TEMPERATURE (P/T) LIMITS Attachment 1, Volume 9, Rev. 1, Page 41 of 418

, Volume 9, Rev. 1, Page 42 of 418 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) , Volume 9, Rev. 1, Page 42 of 418

Attachment 1, Volume 9, Rev. 1, Page 43 of 418 AITS 3.4.3 ITS REACTOR COOLANT SYSTEM 3/4.4.9 PRESSURE/TEMPERATURE LIMITS R1EACTOR COOLANT SYSTEM LIMITING CONDITION FOR OPERATION LCO 3.4.3 3.4.9.1 The Reactor Coolant system (except- th ressurizer) temperature LA302on and pressure shall be limnited.in accordance with t~he :limtlines lFijtures 3.4-2,:.4-3 and 3.&4,ý-durin healiup,, ooldown, criticality, and]

F, nserviceý leak and hydrostatic testing wih

a. A maximum hestu~ of 500F in,any pue hour. eriod t a~nd

b. A maximum cool 4 own of loo4F in any oneh urperiodwith cold 0 leg temperaturz 27 00 F and a maximum co 1down of 509F in any one hour perio; with cold leg temperatu* <270-F.

APPLICABILITY: At all times. A03 ACTION: C Notes A04 With .any of the above limits exceeded, restore the temeratUre and/or ACTIONSA pressure to within the limits within q30 minute pertorm n engineering and C -*levaluation to determine the e ects o out-a -imt c ndition on the

int~egrity of Cýe Reactor :Coolant .System;]: determine that the Reactor Actin A an2 .* M01 Coolant System remains acceptable for continued operation o-r be inat Completion Times N B least HOT STANDBY within the .next 6 .hours a d b ACTI I .C L *T ON w t i I Lhe following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQIJIREHEMT SR 3.4.3.1 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 urveillance specimens representative of the essel materials shall be r ve4 and examined, to determine changes in terial properties,.at the i tervals defined in rAW [

1543A. The results a these examinations shall be used to update Figures 3.4-2, 3.4-3 and.3.4 DAVIS-BESSE, UNIT 1 3/4 4-24 Amendment No. 01, 116 Page 1 of 5 Attachment 1, Volume 9, Rev. 1, Page 43 of 418

Attachment 1, Volume Rev. 1, Page 44 of 418 3.4.3 Figure 3.4-m2 React r Coolant System Pressure-Temperature Limits for H atup and Core Criticality for the First 21 EFPY Page 2 of 5 Attachment 1, Volume 9, Rev. 1, Page 44 of 418

Attachment 1, VolumI Rev. 1, Page 45 of 418 GD *3.4.3 0

C,,

0 m

Vt Figure 3.4-3 React r Coolant System Pressure-Temperature Limit

-4 2600 for oldown for the First 21 EFPY Notes.

01:1 2400 I. Allowable oooldown rate a 270 F Is 100 FAnr (Ramp), limited by .............. . .. ...

a 1S F step change follow by an 9 minute hold. *~~~~........ .-.- - . .. .,.. .. . . . . . .. . .. . . .........

2200

4) 2. Allowable oxldown rate low 270 F Is 50 F/hr (Ramp), limitodby (.

zC,, a 15 F stop change folio ad by an 18 minute hold. ..... 4...

2000

4) 3. A maximum stop temper uro change of IS F Is allowable when mmovlng: Point TemD Press 1800 all RC pumps from opera Ion with the DHR system operating. The stop --.----................ .................

E temperature change Is d fined as PC tamp minus the OHR return lamp "0A 156

4) to the reactor coolant sy tam prior to stopping ail PC pumps. 2 120. .08 U,

1600

4. When the decay heat to oval system (DH) Is operating without any , 1 C 170 477 1400 PC pumps operating. caled OH return temperature to the ... T .. " .......... D 195 477 LAO2 iZ reactor vesselshalbe add. E 198 640 Co 1200

5. The acceptablo pressur and tomporature combnnadons are ' F 270.141.

0 below and to the right o the limit curve. G 315 2219, C5) 0 1000 6. Instrument error Is not a untod fot" In these limits,

- - - ... ......... H I

405 483 2326 ...

2500 U

4) 800

. ..... i....... .i...... ., ..... .... a .. -......;..................... .. . ......... .. . ............... *... ...*......... i........*...

4) 600 CO. .... t ...... ........ . ' ..... ... .... '.. ..... . ... ;. ...... :. .... .. ....

... : . ...... . ..... .,....... . .... .......... ....... "........ i......

U 400 A 200 4-0 50 100 150 200 250 300 310 400 450 500 Indicated Reactor Coolant Inlet Tempera Lre, F Page 3 of 5 Attachment 1, Volume 9, Rev. 1, Page 45 of 418

Attachment 1, Volum Rev. 1, Page 46 of 418 3.4.3 Fiqure 3.4-4 Reactor Coolant System Pressure-Temperature I eatup and

("I Cooldown Limits for InserviceLeak and 1-Hydro! tatc Tests 2600 for the First 21 EFPY 2400 Notes:

I. Allowable hoetup rate 1 50 Fihr (Ramp), limilod by a 15 F slop change

1 , !

fotowed by an IS min to hold.

2200

2. Allowable cooldown ra above 270 F Is 100 F/hr:(Ramp), lImIted by a IS F stop change foll wed by an 0 minute hold, 2000 0I _ _ _ __

3. Allowable cooldown ra o bolow 270 F Is 50 F/hr (Ramp). limited by 1800 - a IS F stop cheange f. lowed by an 18 minute hold. 4 4 E 4. A maximum stop tom eraturo change of 15 F Is allowable when removin 1600 al RO pumps from o radon with the DHR system operating. The step (D

temperature change I defined as OCtomp minus the DHR return tomp Point Temo Press 1400 "tq the reactor coolant system prior to stoppfng all SO pumps. A I

5. When the decay host removal system (OH) Is operating wIthoul anyi A 70 254 1200 SC pumps operaling Indocalod OH return temperature to the i

90 .364 - LAO2 U

1000 reactor vessel shall a used.

F -

i -

C D

125 205 477 477

6. The acceptable pros uro end temperature combinations are E 215 935 below and to the rig I of the limit curve.

800 * . ° F

G

G 250 1242 Ma. Cd 7. Instrument error Is n I accounled for In those limits. 318 2500 600 -

T 400 200 AJ~I__ _

0 5C) 100 150 200 250 300 350 400 Indicated Reactor Coolant Inlet Tempe ature, F Page 4 of 5 Attachment 1, Volume 9, Rev. 1, Page 46 of 418

Attachment 1, Volume 9, Rev. 1, Page 47 of 418 ITS 3.4.3 Table 4. -5 Reactor Vessel hat rial Irradiation Surveillan e Schedule D3 LETED DAVIS-BESSE, UNIT 1 3/4 4-28 Amendment No. $1, 116 Page 5 of 5 Attachment 1, Volume 9, Rev. 1, Page 47 of 418

Attachment 1, Volume 9, Rev. 1, Page 48 of 418 DISCUSSION OF CHANGES ITS 3.4.3, RCS PRESSURE AND TEMPERATURE (P/T) 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. 1, Page 48 of 418

Attachment 1, Volume 9, Rev. 1, Page 49 of 418 DISCUSSION OF CHANGES ITS 3.4.3, RCS PRESSURE AND TEMPERATURE (PIT) 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.

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 PIT 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 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 1 - 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 Davis-Besse Page 2 of 4 Attachment 1, Volume 9, Rev. 1, Page 49 of 418

Attachment 1, Volume 9, Rev. 1, Page 50 of 418 DISCUSSION OF CHANGES ITS 3.4.3, RCS PRESSURE AND TEMPERATURE (P/T) LIMITS 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, /ST 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 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 Davis-Besse Page 3 of 4 Attachment 1, Volume 9, Rev. 1, Page 50 of 418

Attachment 1, Volume 9, Rev. 1, Page 51 of 418 DISCUSSION OF CHANGES ITS 3.4.3, RCS PRESSURE AND TEMPERATURE (P/T) LIMITS 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 L01 (Category5 - Deletion of Surveillance Requirement) 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-1 543A. 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.

The purpose of CTS 4.4.9.1.2 is to ensure the material irradiation surveillance specimens are removed and examined as required by 10 CFR 50, Appendix H.

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. The proposed change is designated as less restrictive because a surveillance that is required in the CTS will not be required in the ITS.

Davis-Besse Page 4 of 4 Attachment 1, Volume 9, Rev. 1, Page 51 of 418

Attachment 1, Volume 9, Rev. 1, Page 52 of 418 Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs)

Attachment 1, Volume 9, Rev. 1, Page 52 of 418

Attachment 1, Volume 9, Rev. 1, Page 53 of 418 CTS

,RCS P/T 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 shallbe 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.

BXWAOG STS 3.4.3-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 1, Page 53 of 418

Attachment 1, Volume 9, Rev. 1, Page 54 of 418 CTS RCS P/T 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.

BWDG STS 3.4.3-2 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 1, Page 54 of 418

Attachment 1, Volume 9, Rev. 1, Page 55 of 418 JUSTIFICATION FOR DEVIATIONS ITS 3.4.3, RCS PRESSURE AND TEMPERATURE (P/T) LIMITS None Davis-Besse Page 1 of I Attachment 1, Volume 9, Rev. 1, Page 55 of 418

Attachment 1, Volume 9, Rev. 1, Page 56 of 418 Improved Standard Technical Specifications (ISTS) Bases Markup and Justification for Deviations (JFDs)

Attachment 1, Volume 9, Rev. 1, Page 56 of 418

Attachment 1, Volume 9, Rev. 1, Page 57 of 418 RCS P/T 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, and inservice ctiy leak and hydrostatic (ISLH) testing, and data for the maximum rate of change of reactor coolant temperature (Ref. 1).

Each P/T 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 P/T 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).

BVOG STS B 3.4.3-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 1, Page 57 of 418

Attachment 1, Volume 9, Rev. 1, Page 58 of 418 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 (RTNoT) 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 P/T 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 (9 The P/T 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.

BXAOG STS B 3.4.3-2 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 1, Page 58 of 418

Attachment 1, Volume 9, Rev. 1, Page 59 of 418 RCS PIT Limits B 3.4.3 BASES BAC:KGROUND (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 XI, 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 conditionsthat mightcause undetected flaws to propagate and cause nonductile failure of the RCPB, an unanalyzed condition. Reference 1 establishes the methodology for determining the PIT limits. Since the PIT limits are not derived from any DBA analysis, there are no acceptance limits related to the P/T limits. Rather, the PIT limits are acceptance limits themselves since they preclude operation in an unanalyzed condition.

RCS P/T 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 testin gand cality, in

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 PIT 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 P/T 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:

BWOG STS B 3.4.3-3 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 1, Page 59 of 418

Attachment 1, Volume 9, Rev. 1, Page 60 of 418 RCS P/T Limits B 3.4.3 BASES LCO (continued)

a. The severity of the departure from the allowable operating PIT regime or the severity of the rate,of change of temperatureT

b. The length oftime the, limits were violated (longer violations allow the 0

temperature gradient in the thick vessel walls to become more pronouncedand 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 PIT 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 nonductilefailure. 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 P/T 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 I 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 sothat the RCPB is returned to a condition that has been verified by stress analyses.

The 30 minute CompletionTime 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.

BWOG STS B 3.4.3-4 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 1, Page 60 of 418

Attachment 1, Volume 9, Rev. 1, Page 61 of 418 RCS PIT Limits B 3.4.3 BASES ACTIONS, (continued) engineeringQ Besides restoring operation towithin limits; anv,'aluation is required to determine if RCS operation can continue, The evaluationmust verify the RCPB integrity remains.,acceptable and must be completed ere within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months ]

1cniuW*prto. Several methods, may be 'use'd, including 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 XI, 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 "e The 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Completion Time is reasfa to accomplish the evaluation.

The evaluation for a mild violationS possible within this time, but more 0 severe violations may require special, event specific-stress analyses or inspections. JA favorable eval ation must be complet d before continuing 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 oftthe 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 L 2 ii -aRequiredAction and associated Completion Time of Condition A aa 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.

BWOG STS B 3.4.3-5 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 1, Page 61 of 418

Attachment 1, Volume 9, Rev. 1, Page 62 of 418

.RCS PiT Limits

,B3.14.3 BASES ACTIONS (continued)

If the required evaluation for continued operation cannot beaccomplished within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , or the resultsare 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 operation may be considered without completing Required Actio Pressure and temperature are reduced by.bringing theplant to MODE3 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 PIT limits at times other than MODE 1, 2, 3, or 4, so that the RCPB is returned to a condiftion that has been verified acceptable by stress

analysis.

The immediate Completion Time reflects the urgency of initiatingaction to restore the parameters to within the analyzed range. Most violations will not be severe, and the activity can be accomplished within this time ina 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 XA, Appendix E (Ref. 8), may also be used to support the evaluation. However, its use is restricted to evaluation of the vessel beltline.

BVWOG STS B 3.4.3-6 Rev. 3.0, 03/31/04 S

Attachment 1, Volume 9, Rev. 1, Page 62 of 418

Attachment 1, Volume 9, Rev. 1, Page 63 of 418 RCS P/T Limits 8 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 reasonableltime.

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

BVWOG STS B 3.4.3-7 Rev. 3.0, 03/31/04 0

Attachment 1, Volume 9, Rev. 1, Page 63 of 418

Attachment 1, Volume 9, Rev. 1, Page 64 of 418 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. 1, Page 64 of 418

Attachment 1, Volume 9, Rev. 1, Page 65 of 418 Specific No Significant Hazards Considerations (NSHCs)

Attachment 1, Volume 9, Rev. 1, Page 65 of 418

Attachment 1, Volume 9, Rev. 1, Page 66 of 418 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.4.3, RCS PRESSURE AND TEMPERATURE (P/T) LIMITS There are no specific NSHC discussions for this Specification.

Davis-Besse Page 1 of 1 Attachment 1, Volume 9, Rev. 1, Page 66 of 418

, Volume 9, Rev. 1, Page 67 of 418 ATTACHMENT 4 ITS 3.4.4, RCS LOOPS - MODES 1 AND 2 , Volume 9, Rev. 1, Page 67 of 418

, Volume 9, Rev. 1, Page 68 of 418 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) , Volume 9, Rev. 1, Page 68 of 418

Attachment 1, Volume 9, Rev. 1, Page 69 of 418 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 A03 APPLICABILITY: MODES 1 and 2A ACTION:

a. With one reactor coolant pump not in operation, STARTUP and POWER LCO 3.4.4.b OPERATION may be Initiated and may proceed provided THERMAL POWER isr@ tricted to less than 80.6% of RATED THERMAL POWER rand lthln]

1-01 10J- hours setpointsfor the following trips have been reduced in accordance with-Specification2.2.1 for operation with three reactor ACTION A coolant pumps operating:

1. High Flux LCO 3.4.4.b

2. Flux-aFlux-Flow 4Add proposed ACTION B M0l SURVEILLANCE REQUIREMENTS SR 3.4.4.1 4.4.1.1.1 The above reouired reactor coolant loops shall be verified to be in operation land circulating reactot coolant at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . LA01 ACTION A reactor coolant Rumpsoperating etther:F ACTION A -4

a. Within [ ours after switching to a three pump combination if the switch is made while operating, or Prior to react rcriticality if the switch s made while shut- A02

b. down. , I I *See Sp/cial Test Excep~on 3.10.3.

DAVIS-BESSE, UNIT 1 3/4 4-1 Amendment No. MA, 00,,

IZ;, 135 Page 1 of 1 Attachment 1, Volume 9, Rev. 1, Page 69 of 418

Attachment 1, Volume 9, Rev. 1, Page 70 of 418 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 if the 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 l.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 B Davis-Besse Page 1 of 3 Attachment 1, Volume 9, Rev. 1, Page 70 of 418

Attachment 1, Volume 9, Rev. 1, Page 71 of 418 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 LAO1 (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 (Category 3- 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 .

0 Davis-Besse Page 2 of 3 Attachment 1, Volume 9, Rev. 1, Page 71 of 418

Attachment 1, Volume 9, Rev. 1, Page 72 of 418 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. 1, Page 72 of 418

Attachment 1, Volume 9, Rev. 1, Page 73 of 418 Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs)

Attachment 1, Volume 9, Rev. 1, Page 73 of 418

Attachment 1, Volume 9, Rev. 1, Page 74 of 418 CTS RCS Loops - MODES I and 2 3,4.4 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.4 RCS Loops - MODES 1 and 2 LCO 3.4.1.1 LCO 3.4.4 Two RCS Loops shallbe-in operation, with:

a.. Four reactor coolant pumps (RCPs) operating M.

-0 Action a b. Three RCPs operating anýHER POWER stricted to IN-.RT 1 APPLICABILITY: MODES 1 and 2.

DOC M01 SURVEILLANCE REQUIREMENTS I

SURVEILLANCE FREQUENCY 44.1.1.1 SR 3.4.4.1 Verify required RCS loops are in operation. 12:hours, BVAOG STS 3.4.4-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 1, Page 74 of 418

Attachment 1, Volume 9, Rev. 1, Page 75 of 418 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.

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.

2 INSERT 3 Required Action and associated Completion Time of Condition A not met.

OR Insert Page 3.4.4-1 Attachment 1, Volume 9, Rev. 1, Page 75 of 418

Attachment 1, Volume 9, Rev. 1, Page 76 of 418 JUSTIFICATION FOR DEVIATIONS ITS 3.4.4, RCS LOOPS - MODES 1 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 apply if any Required Action and associated Completion Time of Condition A is not met or 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. 1, Page 76 of 418

Attachment 1, Volume 9, Rev. 1, Page 77 of 418 Improved Standard Technical Specifications (ISTS) Bases Markup and Justification for Deviations (JFDs)

Attachment 1, Volume 9, Rev. 1, Page 77 of 418

Attachment 1, Volume 9, Rev. 1, Page 78 of 418 L unless otherwise notedj RCS Loops-ý MODES 1 and 2 B 3.4.4 B3.4 REACTOR COOLANT SYSTEM (RCS)

B 3.44 RCS Loops - MODES 1 and 2 BASES BACKGROUND The primary function of the RCS. isremoval of the heat generated in the fuel due to the fission process, and transfer of this heat, via the steam ge nerators (SGs), to the secondary plant..

The secondary functions of the RCS include:

a. Moderating the neutron energy levelIto'thethermal state, to increase the probability of fissi'orn o

b. Improving the neutron economy, by acting asa reflector,

c. Carrying the soluble neutron poison, boriaci 0

d. Providing a second barrier against fission product release to the environmrent*and e, Removing the heat generated in the fuel due tofissionproduct decay following a unitshutdown.

The RCS configuration.for heat transport uses two RCS loops. Each RCS loop contains anSG:and two reactor coolant pumps (RCPs). An RCP is, located in each. of thetwo SG cold legs. The pump flow rate has been sized to provide core-heat removal with appropriate margin to departure from nucleate boiling (DN B) 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 ihnoperation the reactor power level is retrced /,6RTP to: preserve the core power to flow relationship,

ýthus maintaining the margin to DNB. The intent of the Specification is to require core heat removal with forcedlfldW during poweroperation.

Specifying the minimum numberof 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 co6pS also provides the minimum necessary paths (two SGs) for heat removal.

Flux- AFlux - Flow (Table 3.3.1-1 Function 8))

The Reactor Protection System (RPS) nuiear verpowe 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.

BWAOG STS B 3.4.4-1 Rev. 3.0, 03131/04 Attachment 1, Volume 9, Rev. 1, Page 78 of 418

Attachment 1, Volume 9, Rev. 1, Page 79 of 418 RCS Loops&- MODES 1 and 2

B3.4.4 BASES APPLICABLE Safety analyses contain various assumptins, 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 nu~mber of 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. planthas been performed assuming either three or four pumps5 ar: in operation. The. majority of theplant.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 pumnp operation, the steady state DNB.

analysis, which generates the pressure and temperature SL(iLe., the:

110.2% of 2817 MWt departuro from nucleate boiling ratio (DNBR).limit),.assumes a maximum mnJpoýw~e'r 'o' 112 , .RT Tleeveloperation

four pump The . 1This is the design value oveIrpower is the acciden condition t analysis for0 setpoint of the nuclear overpower (high flux):trip and is basedhon an analysis assumption that bounds possible. instrumentation errors,. The: DN BR. limit, defines a locus of pressure and temperature points that resuIt'in a minimum DNBR greater than or equal.to the critical heat flux correlation:

limit.

The threepump pressure. temperature limit is tied to the steady state DNB.

analysis, which is evaluated each cycle. The. flow used isthe 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 power to flow fatio of theiRPS INSERT I Inuclear overpower V;ased on RCS~flowand AjXIALPOWER'MBALANCp~E} *k.

The.maximum power level for three.pump operationis "setoint.

<80f.6" 0/%RTP and is based on.the three pump flow as'afraction of the four pump flow at full power.

0 Although~the Specification limits operation to'a minimum of three pumps total, existing design analyses show that operation with one pumpin 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).

BVWOG STS B 3.44-2 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 1, Page 79 of 418

Attachment 1, Volume 9, Rev. 1, Page 80 of 418 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. 1, Page 80 of 418

Attachment 1, Volume 9, Rev. 1, Page 81 of 418 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 I and certain acceptance criteria for DN B, four pumps are required at rated power; if RPS setpoints only three pumps are available, power must be. reduced.# Imust be reset; APPLICABILITY In-MODES land 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 Iloops 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, theforced 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

'LCO 3.4.5; 314.6, "RCS Loops wMODE "RCS Loops- MODE, "

0 LCO 3141, "RCS Loops.- MODE 5, Loops Fille LCO 3.4.8, "RCS Loops - MODE 5,. Loops Not Fille

.LCO 3.9.9, "Decay Heat Removal (DHR) and Coolant Circulation -

High Water Level" M E6; and

LCO 3.9:5, "Decay Heat Removal (DHR) and Coolant Circulation -

Low Water LevelsM E6 0

ACT*ONS DJ -for reasons other than Condition A Ifthe requirermentsof'the LCO are not me , the Required Action is to reduce powerand bring the plant to MODE 3. This lowers power level.

and thus reduces the core heat removal needs and minimizes the possibility of violating. DN B 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.

BWOG STS B 3.4.4-3 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 1, Page 81 of 418

Attachment 1, Volume 9, Rev. 1, Page 82 of 418 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 l.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.

O INSERT 3 If any Required Action and associated Completion Time of Condition A is not met or Insert Page B 3.4.4-3 Attachment 1, Volume 9, Rev. 1, Page 82 of 418

Attachment 1, Volume 9, Rev. 1, Page 83 of 418 RCS Loops - MODES 1 and 2 B 3144 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 requi~red,-number of loops in operation. Verification includes flow rate, tem perature, 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 ha.s been shown by operating practice to be sufficient to regularly assess:

degradation and verify operation within safety analyses assqmptions. In addition, control room indication and alarms will normally indicate loop status.

REFERENCES 1.hI_-FSAR, 00 BWOG STS B 3.4.4-4 " . Rev. 3.0, 03/31104 Attachment 1, Volume 9, Rev. 1, Page 83 of 418

Attachment 1, Volume 9, Rev. 1, Page 84 of 418 JUSTIFICATION FOR DEVIATIONS ITS 3.4.4 BASES, RCS LOOPS - MODES I 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.

0 Davis-Besse Page 1 of 1 Attachment 1, Volume 9, Rev. 1, Page 84 of 418

Attachment 1, Volume 9, Rev. 1, Page 85 of 418 Specific No Significant Hazards Considerations (NSHCs)

Attachment 1, Volume 9, Rev. 1, Page 85 of 418

Attachment 1, Volume 9, Rev. 1, Page 86 of 418 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.4.4, RCS LOOPS - MODES 1 AND 2 There are no specific NSHC discussions for this Specification.

Davis-Besse Page 1 of 1 Attachment 1, Volume 9, Rev. 1, Page 86 of 418

, Volume 9, Rev. 1, Page 87 of 418 ATTACHMENT 5 ITS 3.4.5, RCS LOOPS - MODE 3 , Volume 9, Rev. 1, Page 87 of 418

, Volume 9, Rev. 1, Page 88 of 418 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) , Volume 9, Rev. 1, Page 88 of 418

Attachment 1, Volume 9, Rev. 1, Page 89 of 418 ITS 3.4.5 ITS 3/I4.4 MIACTOR COOLANT STSTEM 19 S'RUZDOWWNMA MarT S.TPNDBIT L1reTIC COVDITION FOR OFEP.10?1 3..12 A. At least tva coolint loops salb LCO 3.4.5

2. Retactor *)lent Loop 2 and its asso/ Itad teams

3. Decay Beat ]Oe & ll*oop I,* See ITS 3.4.6, ITS 3.4.7, and

4. Decay zff--,t Ree a Loop 2.* ITS 3.4:8

b. At least one t a ve coolant loops .shall be in Not more tban one decay beetr~ova1 p=p nay be operatei 1th suction path

,1ath the solepower cou~opaertor, has the control0 has been through DE-1l and D-11 rzo d fro the been re-aved fro tITS 12 valve operator, or manual valves DE-21 opened.

),.

and

D-12 unless M-423

-and Me axe See ITS 3.4.6, ITS 3.4.7, and 3.4.8 I A02

d. The p Ion. of Spectffitl@US 3.0).3 And /1.4.4 are not eppl ,ble. / -LSee ITS 3.4.7 AY?,tIC,*_I'17: MO*t 3 an, Aand ITS 3.4.8 MI See ITS 3.4.6 ACTION A,- than the above required COoa1n, loops O otLZ../

Required Action C.2 Wit ssac I'q corectA03 Jta ACTION B ACTION C l'henol r meoecooer sot c itheabein wbe 0 OO a'

o b l.oop

)loop---------- to opSat,.,O.-{3uls/h eeth"nMD rmror two required RCS 12

'cetmeau*ao**-t-loops inoperable See ITS 3.4.7 ebThe decay heat removal ptnps may be de-energized for up too i houra.

provided (1) no operations are permitted that would cause dilutionrSee of ITS 3.4.6o reactor coolant lhe system boron concentration, and (2) core outlet ITS347, andl

( maintained at least 1OaF below saturation temperature..d

'I~t~ ITS3.4.8 9 DAVIS-BESSE UNIT I 3/4 4-2 Amendment No. A,i 4,

19. 92 Page 1 of 2 Attachment 1, Volume 9, Rev. 1, Page 89 of 418

Attachment 1, Volume 9, Rev. 1, Page 90 of 418 ITS 3.4.5 ITS 3/4.4 REACTOR COOLANT'SYSTEM

~TmVWTT.T.ANCR R~flh1TRMflI74 SURVELLANC ..... "......

S 3.4.6, See ITS ITS 3.4.7, and

.4.1.2.1 The required decay heat removal loop(s) shall be determined OPERABLE-er Specification 4.0.5. I_ ITS 3.4.8 ]

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) 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 3.4.5.1 4.4.1.2.3 At least one coolant loop shall be verified to be in operation -i Icircu atinm riactor 9olantLat least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

~~Add proposed SR 3.4.5.3 @

DAVIS-BESSE, UNIT 1 3/4 4-2a Amendment No. 3p,135 Page 2 of 2 Attachment 1, Volume 9, Rev. 1, Page 90 of 418

Attachment 1, Volume 9, Rev. 1, Page 91 of 418 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. 1, Page 91 of 418

Attachment 1, Volume 9, Rev. 1, Page 92 of 418 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. 1, Page 92 of 418

Attachment 1, Volume 9, Rev. 1, Page 93 of 418 DISCUSSION OF CHANGES ITS 3.4.5, RCS LOOPS - MODE 3 REMOVED DETAIL CHANGES LA01 (Type 1 - 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 I - 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. 1, Page 93 of 418

Attachment 1, Volume 9, Rev. 1, Page 94 of 418 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. 1, Page 94 of 418

Attachment 1, Volume 9, Rev. 1, Page 95 of 418 Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs)

Attachment 1, Volume 9, Rev. 1, Page 95 of 418

Attachment 1, Volume 9, Rev. 1, Page 96 of 418 CTS RCS Lo6ps,- MODE. 3 314.5 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.

.---NOE 0 ----------------..... - -- --;.........-----

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

<8 hoursper 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months period for thei-tr nsition to or fromt Decay Heat Removal Systen, 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 permittedth t would cause intro, uction of coolant into the CS with boron conce tration less.than re uired to meet the SID of LCO 3.1.1 and 0

b. Core ou let temperature is ma tained at leastjl O01F 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. 1, Page 96 of 418

Attachment 1, Volume 9, Rev. 1, Page 97 of 418 CTS RCS Loops - MODE 3' 3.4.5 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME C. Two RCS loops 4 .4.

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 3A5- ------------------------------- ---------------

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.

BVWOG STS 3.4.5-2 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 1, Page 97 of 418

Attachment 1, Volume 9, Rev. 1, Page 98 of 418 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. 1, Page 98 of 418

Attachment 1, Volume 9, Rev. 1, Page 99 of 418 Improved Standard Technical Specifications (ISTS) Bases Markup and Justification for Deviations (JFDs)

Attachment 1, Volume 9, Rev. 1, Page 99 of 418

Attachment 1, Volume 9, Rev. 1, Page 100 of 418 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 ofthe 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 if forced flow is used to meet the RCPs in operation will vary depending on operational needs, and Wso intep of thi LCis olrovi forced flovArom at least one RCP for core [ iied ] 0 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 I and boron mixing-]

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 0 to. avoid idle loop stagnation that might occur ifonly one SG were in service. One generator will provide adequate heat removal. ron reduction ' natural circulation is rohibited because mixing obtain a homoge)'eous concentration inl portions of the RCS canf ot be ensurA-I APPLICABLE No safety analyseseare performed with initial conditions in MODE 3.

SAFETY ANALYSES Failure to provide heat removal may result in challenges. to a fission related to loss of RCS ls product barrier. The RCS loopsare part of the primary success path that functions 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).

BXAOG STS B 3.4.5-1 Rev. 3.1, 12101/05 Attachment 1, Volume 9, Rev. 1, Page 100 of 418

Attachment 1, Volume 9, Rev. 1, Page 101 of 418 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 nd can be transferring heat from the reactor coolant at a controlled rate. Forced usthe ed toLCO0 meet reactor coolant flow is the re ired way to transport heat, although natural meet the LCO requirements. quirements l RCP meetsflow when forced flow is being ud to Furthermore, the requirements circulation the provides LCO requirement removaLS" adequate for in op rtoof one: running one loopAfminimum for aI lop in operation are also met whend The Note permit a limited period of o eration without RC s. All RCPs may be remove from operation for -8hours per 24 hou period for the transition to orf om the Decay Heat emoval (DHR) Sys m, and otherwise may e de-energized for _1hour per 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months p nod. This means thatrnt ral circulation has be circulation, borjn reduction with cool n established. Wh n in natural nt at boron concenrations 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 temperature s that no vapor bubble natural circula on flow obstruction.

at least [10]*F bel w the saturation may form and pos ibly cause a 05 In MODES 3, , and 5, it is someti s necessary to sto all RCP or DHR 0 pump forced irculation (e.g., chan e operation from on DHR train to the

,ri or sta tup testing, to perf rm the trnition other, to perf rm surveillance tt thet to and from D R System cooling, o to avoid operation below the RCP minimum net ositive suction head imit). The time per d is acceptable because nat al circulation is adeq ate for heat remov 1,or the reactor coolant temp rature can be mainta ned subcooled and boron stratification affecting rea tivity control is not ex ected.

ifforced flow is usedn OPERABLE RCS loop consists of at least one OPERABLE RCP and

= 1 ' an S G that is O P E R A B LE. A n R C P is O P E R A B LE if it is ca pable of being powered and is able to provide forced flow if requiredl I APPLICABILITY 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, "RCS Loops - MODE LCO 3.4.7, "RCS Loops - MODE 5, Loops Fille 0 BWOG STS B 3.4:5-2 Rev. 3.1,12/01/05 Attachment 1, Volume 9, Rev. 1, Page 101 of 418

Attachment 1, Volume 9, Rev. 1, Page 102 of 418 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. 1, Page 102 of 418

Attachment 1, Volume 9, Rev. 1, Page 103 of 418 RCS Loops - MODE 3 B 3.4.5 BASES APPLICABILITY (continued)

LCO 3.4.8, "RCS Loops - MODE 5, Loops Not Filledl 4f 0 LCO 3I9.4, "Decay Heat Removal (DHR) and Cof nt Circulation -

LCO 3.9.5, High Water Level" OMWE 6' "Decay Heat Removal (DHR) and Coolant Circulation - Low 00 Water Lev ACTIONS A.1 0 Ior natural circulation If one RCS loop is inoperable, redundancy for forceodflow 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 Action 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 placedon 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~reasonablebased on operating experience, to achieve cooldown and depressutization from the existing plant conditions and without challenging plant systems. 0 CA1 and C.2 If two RCS loops are inoperable or a required RCS loop is not in operation, lexcept as,4rovided in th"ote in the 04CO 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 withoutffoced 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.

BWOG STS B 3.4.5-3 Rev. 3.1, 12/01105 Attachment 1, Volume 9, Rev. 1, Page 103 of 418

Attachment 1, Volume 9, Rev. 1, Page 104 of 418 RCS Loops - MODE 3 B 3.4.5 BASES SURVEILLANCE SR 3A4.5.1 REQUIREMENTS This SR requires verification every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months that the requiredlu er of Ioos r is in operation. Verificationincludes flow, rate, onatura I

.temperature, or pump:status. monitoring, which helpensure. that forced cir flow is providing heat removal. The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months interval has beenshown by operating practice to be sufficientlto regularly~assess RCS loop status. In addition, control room indication and alarms will normally indicate loop status.

S0 Verification that each required RCP is OPERABLE ensures that the single failure criterion is met and that an additional RCSloop 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 beacceptable 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.

S*R 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 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.

BWOG STS B 3.4.5-4 Rev. 3.1, 12/01105 Attachment 1, Volume 9, Rev. 1, Page 104 of 418

Attachment 1, Volume 9, Rev. 1, Page 105 of 418 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. 1, Page 105 of 418

Attachment 1, Volume 9, Rev. 1, Page 106 of 418 Specific No Significant Hazards Considerations (NSHCs)

Attachment 1, Volume 9, Rev. 1, Page 106 of 418

Attachment 1, Volume 9, Rev. 1, Page 107 of 418 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. 1, Page 107 of 418

, Volume 9, Rev. 1, Page 108 of 418 ATTACHMENT 6 ITS 3.4.6, RCS LOOPS - MODE 4 , Volume 9, Rev. 1, Page 108 of 418

, Volume 9, Rev. 1, Page 109 of 418 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) , Volume 9, Rev. 1, Page 109 of 418

Attachment 1, Volume 9, Rev. 1, Page 110 of 418 ITS 3.4.6 ITS I

3/4.4 PLIXTOR :COO"A.'T STSTDI SiRtrrDoU' AN" 1HOT STASDMEY LTriMNC CONMITION, FOR O?*EWION LCO3.4.6 3.4.1.2 a. At least t;m coolant loops 11,Lited i&Atal e

b. Ac lsot o. tal e/*o* laits loop ihall be tea LA01

3. Dheca lots Zr*i m LOPf path through %Y111 atd DR-i12 .leae. A0

.v:Lc'4h the a is *uctiO 4~l Deay3 i. t cm f,5 LOOP 2 * ..- .04 aett e T ene j ACTION A, . W~t~h ec that th a o ,q olanlze oopsn loops inAZZ th os .pSeeebe rmve tts . Rlosa OAL Sd~

b.-Areauire eoutoflooto*ý an3.4.7 andj oprtlt It L02 IntAte LongOfeclve actIons to reuu0. eurdCooat ý 4Vloo t peal& t r w eurdR losioeal See ITS 4......__ l___ __ ITS 3.4.8 AI " .......... ___ __ __ ___ __ ___ __

Lt= way b: hor 24 L 3.4.6 Nemr decayIt on* de -O beat AC0TI lO ACTONded (

Rereusred'ActionthB.2thedl t:

1)n tholoopmaynot rea.tarosyte thf coolant e cay nR0 nestepiayso oould t loosh 3unles edte pimar he e sleced tI10 MOE oeraermitt i hetate r ~e ctvel sys t~e s dhIos t=oo3.

eign co eptlunad---*.SeelITS.

1n oelt, etiperureand arooredute ntu thO s.A;34 reslce See ITS ]

Ah ACTO346Not decay theat s temoan tum mayboe e deergqired for upto10hour Act1)B.o Requroiree operat*ons are emtl ted OthtciT Oldcauedtion to"t t f 0 eactrequrcoolant the syte conentratn loron $ San(2)A outlet oore tenpierature Is matintined at leastiveF below saturation temperature.

DAVIIS-BESSE UNIT 1 3/4 4-2 Amendment N~o. A, A1 *',

... ... . ... Page 1 of 2 Attachment 1, Volume 9, Rev. 1, Page 110 of 418

Attachment 1, Volume 9, Rev. 1, Page 111 of 418 ITS 3.4.6 ITS 3/4.4. REACTOR COOLANT SYSTEM StRVEILLANCE REOUIREMENTS.

.4.1.2.1 The r uired decay bheat rem7 l loop(s) shall bej 'ermnined OPERABLE A04 er Sneifieatid'n 4.0.5.

er Snecificittidn,4.M. X14 SR 3.4.6.2 m7" 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) 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 operation[iý-

circu atinaratr olant at. least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . I Add proposed SR 3.4.6.3 M02 DAVIS-BESSE, UNIT 1 3/4 4-2a Amendment No. Al,35 Page 2 of 2 Attachment 1, Volume 9, Rev. 1, Page 111 of 418

Attachment 1, Volume 9, Rev. 1, Page 112 of 418 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. 1, Page 112 of 418

Attachment 1, Volume 9, Rev. 1, Page 113 of 418 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. 1, Page 113 of 418

Attachment 1, Volume 9, Rev. 1, Page 114 of 418 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 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 3.4.6 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. 1, Page 114 of 418

Attachment 1, Volume 9, Rev. 1, Page 115 of 418 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 (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 3.4.6 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. 1, Page 115 of 418

Attachment 1, Volume 9, Rev. 1, Page 116 of 418 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. 1, Page 116 of 418

Attachment 1, Volume 9, Rev. 1, Page 117 of 418 Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs)

Attachment 1, Volume 9, Rev. 1, Page 117 of 418

Attachment 1, Volume 9, Rev. 1, Page 118 of 418 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 h-eat removal (DHR) loops shall be OPERABLE and one loop shall.be in operation.

314.1.2 --- - -- - ------- -- - ---%-N

u I r-77-- ---.

n---- -.-.. . - -----.-- r-.-.----

All reactor /*olant pumps (RCPs) maybe remove rom operationfor 0

Note **

All reactor coolant pumps (RCPs) and lE- 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months per 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months periodfor the transition to 3r fromthe! DHR System Add all CPs and DHR pumps may be de-energized for ! 1 hourIpj1 18 hourp inod for any otjr reasonj provided:

a. No operations are permittedthat would cause introduction of coolant 0

into the RCS with boronconcentration less than required to meet SU

SHUTDOWN W the SDM of LCO. 3.1.1,and MARGIN (SDM)";

b. Core outlet temperature is maintained at least' 10'F below saturation temperature.

APPLICABILITY: MODE 4.

ACTIONS CONDITION REQUIRED ACTION .COMPLETION TIME Action a A. One required loop AA 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. 1, Page 118 of 418

Attachment 1, Volume 9, Rev. 1, Page 119 of 418 CTS RCS Loops - MODE 4.

3A.6 ACTI ONS. (conti nued)

CONDITION REQUIRED ACTION COMPLETION TIME Actions a and b B,. Two required loops BJ 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 of LCO 3.1.1.

operation.

AND 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 3.4.6. - -----.......-

Not NOTE ------.--------

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 poweravailable-to each required pump.

4.4.1.2.2

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 a) ýt 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. 1, Page 119 of 418

Attachment 1, Volume 9, Rev. 1, Page 120 of 418 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 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. This was approved by the NRC as documented in the NRC Safety Evaluation for Amendment 38. The Note has been maintained, as modified by the deletion of the "may be removed from operation for < 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months per 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months period for the transition to or from the DHR System, and" and "per 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months period for any other reason,"

consistent with the CTS allowance since some time might be needed to tranisiton from an operating DHR loop to an operating RCS loop (i.e., a loop with forced or natural circulation flow) or from an operating RCS loop to an operating DHR loop.

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. 1, Page 120 of 418

Attachment 1, Volume 9, Rev. 1, Page 121 of 418 Improved Standard Technical Specifications (ISTS) Bases Markup and Justification for Deviations (JFDs)

Attachment 1, Volume 9, Rev. 1, Page 121 of 418

Attachment 1, Volume 9, Rev. 1, Page 122 of 418 RCS Loops.- MODE 4 8 -3A.6, B3.4 REACTOR COOLANTSYSTEM (ROS)

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 'pdison, boric acid.

f In MODE 4 either reactor coolant pumps (RCPs) or DHR:pumps :can be.

Ifforced flow is I used fo coolant circulation. The number of.pumps in operation can, vary used to meetjthe o suit the operational needs. 'he irnt of thisLCtorovideforce dd flovfrom at least one RCP or one DHR. pump for decay heat removal, and (

ý s rviedItransport. The flow provided by one RCP or one DHR pump:is. adequate for heat removal. The other intent of this LCO is to requirethat two paths SINSERT 1 (loops)be availabe to provideiredundancyfor heat rernoval.

Irelated to loss/

of RCS loops APPLICABLE SAFETY No safety analyses"are performed with initial condition in MODE-4. 0 ANALYSES RCSLoops -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 DH R, 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 oneloop in

'operation provides enough flowto remove the decay heat fkromr the'ore o with forcedcirculation. The second loop that is required to be 5C

'OPERABLE providesredundant pathN for heat removal.

7 RC s I and. DHR pumps The Note permits a limited period of operation without RCP. ARCPS and DHR pumps "} imay be removed from operation Ifor

8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months s§.-er24 hour period for.thel

/transi tooor from the DH R System andherwise maybe.de-energiz provided certain for s 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months per 8- .ur period. This means thataturalcirCulationha requirements are met. been established using the SGs) The Note prohibits boron dilution with coolant at boron concentrations less than required to assure the SDM of "SHUTDOWN LCO.3.1.* is maintained when forced flow is stopped because an even MARGIN (SDM),' concentration distribution cannot be ensured. Core outlet temperature is

'to be maintained at least 10°F below saturation temperature so that no 5 vapor bubble may form and possibly cause a natural circulation flow.

obstruction.

INSERT 1A BWUG STS B 3-4.6-1 Rev. 3.1, 12/01/05 Attachment 1, Volume 9, Rev. 1, Page 122 of 418

Attachment 1, Volume 9, Rev. 1, Page 123 of 418 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.

5 INSERT IA In MODE 4, it is sometimes necessary to stop all RCP and DHR pump forced circulation. This is permitted to change operation from one DHR loop to the other or perform the transition to and from the DHR System. The transition is between DHR loop operation and RCS loop operation (either forced or natural circulation flow). The time period is acceptable because, while natural circulation is not yet established, the conditions for natural circulation exist, (i.e., secondary side water level is within limits),

the time period is short, the reactor coolant temperature can be maintained subcooled, and boron stratification affecting reactivity control is not expected.

Insert Page B 3.4.6-1 Attachment 1, Volume 9, Rev. 1, Page 123 of 418

Attachment 1, Volume 9, Rev. 1, Page 124 of 418 RCS Loops - MODE4 B 3.4.6 BASES LCO (continued)

The Note also p rmits the DHR pump tobe stopped 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. en the DHR pumps are stopped, no alt mate heat removal path e ists, unlessthe RCSInd SGs have been laced in service in force or naturaltcirculation The response of t e RCS without the DHR Syste depends on the cor decay heat load a d.the length of time that the D R pumps are stoppe As decay heat di inishes, the effects on RCS temperature and pre sure diminish. Wit ut cooling by DHR, higher h at loads will cause th reactor coolant te perature and pressure to in 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 o 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 (P/T and LTOP) nd 10°F subcooli g limits or

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

Ifforced flow is used, n OPERABLE RCS loop consists of at least one OPERABLE RCP and an SG that is OPERABLE 0

0)

ISrT 2 Similarly for the DHR System, an OPERABLE DHR loop is comprised of th OPERABLE DHR pump(s) capable of providing forced flow to the he aners). 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 3.4A4, "RCS Loops - MODES 1 and LCO 3.4.5, "RCS Loops - MODE LCO LCO LCO 3.4.7, 3.4.8, 3.9.4, "RCS Loops - MODE 5, Loops Filled--"

"RCS Loops - MODE 5, Loops Not Fille "Decay Heat Removal (DHR) and Coolant Circulation -

0 High Water Level" M--E-6and 0*

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

Low Water L F(M -E 6). 0)

BWOG STS B 3.4.6-2 Rev. 3.1, 12/01/05 Attachment 1, Volume 9, Rev. 1, Page 124 of 418

Attachment 1, Volume 9, Rev. 1, Page 125 of 418 B 3.4.6 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 Ž 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-12.

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. 1, Page 125 of 418

Attachment 1, Volume 9, Rev* 1, Page 126 of 418 RCS Loops - MODE 4 B 3.4.6 BASES ACTIONS A.1 Ifonly 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 duri[A conditions per.,fitted by the Npt'e in the LCOI (

Iseion 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 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 withoutlfo ed 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.

BWOG STS B 3.4.6-3 Rev. 3.1, 12/01/05 Attachment 1, Volume 9, Rev. 1, Page 126 of 418

Attachment 1, Volume 9, Rev. 1, Page 127 of 41.8 RCS Loops - MODE 4 B 314.6 BASES SURVEILLANCE SR 3.4;6.1 REQUIREMENTS This Surveillance requires verification eyeryl.2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months of the required.DHR or natural or RCS loopin operation to ensure force flow is.providing decay heat cirulation 0 removal. Verification includes flow rate, temperature, or pump.status 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 additioncontrol room indication and alarms will normally indicate loop status. 0 SINSERT4 0

SR 3,4.6.a4---ii 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.

BXAOG STS B 3.4.6-4 Rev. 3:1, 12/01/05 Attachment 1, Volume 9, Rev. 1, Page 127 of 418

Attachment 1, Volume 9, Rev. 1, Page 128 of 418 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. 1, Page 128 of 418

Attachment 1, Volume 9, Rev. 1, Page 129 of 418 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.

0 Davis-Besse Page 1 of 1 Attachment 1, Volume 9, Rev. 1, Page 129 of 418

Attachment 1, Volume 9, Rev. 1, Page 130 of 418 Specific No Significant Hazards Considerations (NSHCs) 0 Attachment 1, Volume 9, Rev. 1, Page 130 of 418

Attachment 1, Volume 9, Rev. 1, Page 131 of 418 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. 1, Page 131 of 418

Attachment 1, Volume 9, Rev. 1, Page 132 of 418 ATTACHMENT 7 ITS 3.4.7, RCS LOOPS - MODE 5, LOOPS FILLED Attachment 1, Volume 9, Rev. 1, Page 132 of 418

, Volume 9, Rev. 1, Page 133 of 418 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) , Volume 9, Rev. 1, Page 133 of 418

Attachment 1, Volume 9, Rev. 1, Page 134 of 418 ITS 3.4.7 ITS 3/4- PZ'-CTOR COOIA6'. STS-M-4 11 S5IUT¶DOUM ANM HO~T STANDBY LTYTT4C CONMIT10~ FOR OPERT~.ION LAQI 3.4.1.2 a. At least tva CE~

ooflat loops I~hiJi iiý4 shal3l be LCO 3.4.7 OPERABLZZ:

1. Reactor -Coolan Loo1P I And its associated Gtssd LAO 1

2. Reactor, Coo &atLoop 2 and Its &SoCIAt PkteaA

3. Decay I Iewvu+/- 14.op. l.

4. De4cay R at Lomoval Loop2.111

b. At least cu on. ýt. a 'e coolant loops thanl be. Lu operatiou."~

/

C. Not =ore t one derAy hbas reoval pip nay peatad.

1. suctiqa path hroug'h DR-11 and D8- 2=uless A~

with ithe the couat . pover has be renod fro. the Dl- 1 -and MI-12 va1 operator, or, valves DI-21 mad -2 at 04 r o A0 2

d. he proyisioan ofý Specifit~tious 303

__ 3ODEA07J See ITS APPLICA31 AcrloqI:

{*

J

[See3.4.5 ITS one With &so than the abovo rsquired coolmt loops 0?pA3LE, ACTION A, Required Action B.2 A.

1=aed~atcey initiate tortaCO viS 5acion to return the required '-*ant 10o28 to ?pu.L.L StatUs £.

v:% ithin 20 asOO hourso.

F See ITSl 3.4.5 and ITS 3.4.6 J A03 possib la, be inO~l ctsz ACTION B b.' With none of the suspend..,al operatim IoucOCentatlD of the lnii*Ate correti've acttU to,MTZ*Zt',u r*e%Lt- u a.u&

A-'

loop to operatio. A04, loop

Tenra/r lpesure.'are, emrgency power souree may be inoperabl in M be selected may not within in MODE the decay 3 unless the primary s e heat remval system's design Cgn i his ure and ]

See ITS 3.4.5 I

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

DAVIS-BESSE UNIT I 3/4 4-2 Amendment No. A. A, t, MA. 92 Page 1 of 2 Attachment 1, Volume 9, Rev. 1, Page 134 of 418

Attachment 1, Volume 9, Rev. 1, Page 135 of 418 ITS 3.4.7 ITS 3/4.4 REACTOR COOLANT SYSTEM SURVEILLANCE REOUIREMENTS 4.1.1.1 The r 40 iuired . decay heat remo 1 loop(s) shall be d terained OPERABLE er Secifcat ;ý6.0.5. ý SR 3.4.7.2 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) 18 inche 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 cooan hours *f no re tor coo (b) an 35pumps inchesare above peratl.the lover tube sheet once per 12 See ITS 3.4.5 and ITS 3.4.6 I

SR 3.4.7.1 4.4.1.2.3 At least one coolant loop shall be verified to be in operation and Icircu)ating r'actor at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

=6olant LAO2 Add proposed SR 3.4.7.3 M DAVIS-BESSE, UNIT 1 3/4 4-2a Amendment No. p,135 Page 2 of 2 Attachment 1, Volume 9, Rev. 1, Page 135 of 418

Attachment 1, Volume 9, Rev. 1, Page 136 of 418 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. 1, Page 136 of 418

Attachment 1, Volume 9, Rev. 1, Page 137 of 418 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. 1, Page 137 of 418

Attachment 1, Volume 9, Rev. 1, Page 138 of 418 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 Not used.

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 REMOVED DETAIL CHANGES LAO01 (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 3.4.7 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 Davis-Besse .Page 3 of 5 Attachment 1, Volume 9, Rev. 1, Page 138 of 418

Attachment 1, Volume 9, Rev. 1, Page 139 of 418 DISCUSSION OF CHANGES ITS 3.4.7, RCS LOOPS - MODE 5, LOOPS FILLED 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 (Category 4- 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 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 Davis-Besse Page 4 of 5 Attachment 1, Volume 9, Rev. 1, Page 139 of 418

Attachment 1, Volume 9, Rev. 1, Page 140 of 418 DISCUSSION OF CHANGES ITS 3.4.7, RCS LOOPS - MODE 5, LOOPS FILLED 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. 1, Page 140 of 418

Attachment 1, Volume 9, Rev. 1, Page 141 of 418 Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs)

Attachment 1, Volume 9, Rev. 1, Page 141 of 418

Attachment 1, Volume 9, Rev. 1, Page 142 of 418 CTS RCS Loops - MODE 5, Loops' Filled 3A4.7 3.4 REACTOR COOLANT SYSTEM'(RCS) and decay heat removal (DHR) loops shall be OPERABLE and one loop shall be in operation.

3.4.7 RCS Loops Two loops consisting of any combination of RCS loops

-. MODE 5, Loops Filled 3.4.1.2 LCO 3.4;7 operation, One decay and heat itherf remova I(DHR) loop hall be OPERABL and in

a. .OeadionaI DHR loop shal be OPERABLE or 0

b. The *yn I of each steam ge erator(SG) shall be [5]%.

3.4.1.2 Note **

EZZ NOTE .J-- -- ------ - --------

The DHR pump of the loop in operation may be removed from 0

operation for *_ 1 hourl per-ho- period provided; No operations are permitted thatwould cause introduction of

,"SHUTDOWN MARGIN (SDM)"; coolant into the RCS with boron concentration less than required to meet the SDM of LCO 3, 1 .1 and 0 b., Core outlet temperature is maintained at least 10°F below saturation temperature.

2. One requi d DHR loop may be noperable forupto, hours for surveillan e testing providedth t the other DHR loe and in op ration.

is OPERABLE 0

3. All DHR oops may be not in o eration during plan ed heatup to MODE. when at leastoneR Sloop is in operati n..

APPLICABILITY: MODE 5 with RCS loops filled.

BWOG STS 3.4.7-1 Rev. 3.0, 03131/04 Attachment 1, Volume 9, Rev. 1, Page 142 of 418

Attachment 1, Volume 9, Rev. 1, Page 143 of 418 CTS I

All changes are a unless otherwise noted RCS Loops - MODE 5, Loops Filled 34.7.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME Action a A. One required b R loop AA1 Initiate action to restore a Immediately inoperable. second Dfil loop to OPERABLE status.

AND OR One H.R loop OPE ABLE. A.2 Initiate actio to restore Immediately required S's secondary side wate/evels to within limits. j 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 n within 0 ERABLE status.

limit.

OR AND B,2 I itiate action to restore Immediately 0 One DHR I op OPERABL .

quired SGs seconda ide water level to with n imit.

Action a, Suspend operations that Immediately Action b would cause introductionof coolant into the RCS with OR boron concentration less than required to meet.SDM Required [ loop not of LCO 3,1,1.

in operation.

AND Initiate action to restore one Immediately D R loop to OPERABLE status and operation, BVVOG STS 3.4.7-2 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 1, Page 143 of 418

Attachment 1, Volume 9, Rev. 1, Page 144 of 418 CTS RCS Loops - MODE 5, Loops Filled 3.4.7 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY or RCS 4.4.1.2.3 ,SR: 3.4.7.1 Verify required DHR.loop. is in operation. 12hours0

, for each required RCS loop, )

0 4.4.1.2.2 SR 3.4.7.2 Verif re ired SGsecondary side water leveler, j 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 7days power available to each required DHR pump.

BWVOG STS 3.4.7-3 Rev. 3.0,W03/31/04 Attachment 1, Volume 9, Rev. 1, Page 144 of 418

Attachment 1, Volume 9, Rev. 1, Page 145 of 418 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 2 and 3 have been deleted and the ACTIONS have been modified to reflect the natural circulation option. NOTE 1 has been maintained, as modified by the deletion of the "per 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months period," consistent with the CTS allowance since some time might be needed to transition from an operating DHR loop to an operating RCS loop (i.e., a loop with forced or natural circulation flow). 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. 1, Page 145 of 418

Attachment 1, Volume 9, Rev. 1, Page 146 of 418 Improved Standard Technical Specifications (ISTS) Bases Markup and Justification for Deviations (JFDs)

Attachment 1, Volume 9, Rev. 1, Page 146 of 418

Attachment 1, Volume 9, Rev. 1, Page 147 of 418

RCS Loops - MODE 5, Loops Filled B.3.4.7 B 3.4 REACTOR COOLANT SYSTEM (RCS)

B13.4.7 RCS Loops - MODE:5, Loops.Filled BASES BACKGROUND In MODE 5 with 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 p)(chanor r. While the principal means for decay heat removalis via the DHR System, the SGs Q

are specified as a backup means-for redundancy. Although the.SGs cannot remove heat unless steaming occurs ýwhich i*ot Possible id IME 5, 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 carrierfor soluble neutron poison, boric acid.

In MODE 5 with RCS loops filled, DHR loops are the principal means for If forced flow is used to meet the heat removal. The number of loops in operation can vary to, suit the operational needs.orThe itnt of t is] LCOQ tre is provided forced flovwfrom at D 0

least one DHR loop, for-decay heat removal and transport. The flow provided by-one DH R loop is adequate for decay heat removal. The other-intent of this LCO is to require thata second path-be available to INSERT1 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)-iimitations, and the SG will not be able to provide steam for the turbine driven feed eri. l pumps. However, to ensure that the SGs-can be used as a heat sink, a---11

_ cI_ o driven feedMI4 pump is needed, because it is independent of stearmCondeo sate puml s, startul pumps, or the rotorriven IThe Startup Feed Pump *edwater pump can be used. If RCPs are available, the steam generator level need not be-adjusted. If RCPs are not available, the water level Motor Driven is must be adjusted for natural circulation. The high entry point in the

ýgeneratorlsholdb accessible from the Tee *er pump so that natural Feedwater Pump 0 frither circulation can befstimulated The SGs are primarily a backup to the i neded. DHR pumps, which are use 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.

BWOG STS B 34.7-1 Rev. 3.1, 12/01/05 Attachment 1, Volume 9, Rev. 1, Page 147 of 418

Attachment 1, Volume 9, Rev. 1, Page 148 of 418 B 3.4.7 (D 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 3.4.7-1 Attachment 1, Volume 9, Rev. 1, Page 148 of 418

Attachment 1, Volume 9, Rev. 1, Page 149 of 418 RCS Loops - MODE 5, Loops Filled B.3.4.7 BASES

.APPLICABLE. No safety analyses are performed .with initial conditions in MODE 5.

"*related to loss of RCS loops SAFETY ANALYSES RCS Loops - MODE 5.(Loops Filled) satisfies Criterion 4 of 10 CFR 50.36(c)(2)(ii). .{

0 , RCS or DHR, LCO The purpose of this LCO is to require that lat least one-of the DH loo s.4 one ofthese loops be* be OPERABLE and'in operation with an *'dditional DHR loo OPEABL Iorbot S~ wih/econdary side water lVol > [r50]o/4. ]o0ne DHR 10o1 provides Suffici t orced circulationto p rform the safety functions f the ireactor~coolant/under these coiditions. he second DHR loop is n rrmally maintained as a backup tothe operatinh D HR loop to. providered for dec~ay hea reo~val. HewAv*.r sf tl ' standby rOHR I(n i.* nit* dancy J

INSERT 2 OPERA6ZBLE/a s*icient:aternate m ehod of providing redundan -heat removal paos is to provide both SG with their secondary side ter.

pu e levels Ž: [5 %. Should the operatin DHR loop fail, the Ss3c#/uld be r I Lusled to reove the decay heat, The LCO Note permitsthe DHR m to be removed from operation for up to.

1 hourlper/8 hour 06riol. The circumstanoes for-stopping both DHR lop -61sare to be limited to situations where: (a) Pressure and temperature.

increases can be maintained well within the allowabl e pressure (P/T.an.rd low temperature overpressure protection) and 10'F subcooling limit .or (b) Alternate. heat paths through the SGs areieri .

The Note prohibits boron dilution with coolant at boron concentrations "SHUTDOWN MARGIN less than required to assure the SDM.of LCO 311.1 is maintained when (SDM),-. J DHR forced flow is stopped because an even concentration distribution cannot be ensured, Core outlet temperature is to be maintained at least, 10°F below saturationtemperature so that no vapor bubble would form and possibly cause a natural circulation flow obstructi6n. I'his MODE, the generator are used as a backup for decay heat rer vaxand, to.

ensure theirivailability, the RCS loop flow path is to .maintained'with subcoole iquid.

In MODE 5, it is sometimes necessary to stop all RCP or DHR. pump forced circulation. This is permitted to change operation from one DHR*

loop LI1M to the otheO, perforrpu-rveillance or staf'ttp testing-perform th transition to and from the DH R Ssen or to*av6id operation'ladlow te

[INSERT 2A IRCP minimum N imt, The time period is acceptable because FNSERT2B natural cir ion is acceptable f 9epat removal the reactor coolant temperature can be maintained subcooled, and boron stratification affecting reactivity control is not expected. j BWOG STS B 3.4.7-2 Rev. 3.1, 12/01105 Attachment 1, Volume 9, Rev. 1, Page 149 of 418

Attachment 1, Volume 9, Rev. 1, Page 150 of 418 B 3.4.7 (D 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.

QINSERT 2A The transition is between DHR loop operation and RCS loop operation (either forced or natural circulation flow).

QINSERT 2B while natural circulation is not yet established, the conditions for natural circulation exist (i.e., secondary side water level is within limits), the time period is short, Insert Page B 3.4.7-2 Attachment 1, Volume 9, Rev. 1, Page 150 of 418

Attachment 1, Volume 9, Rev. 1, Page 151 of 418 RCS Loops - MODE 5, Loops, Filled B 3.4.7 BASES LCO. (continued)

Note 2 allows one D loop to be inoperable for a perio of up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months provided that the othtr loop is OPERABLE and.in operion. This permits periodic surveillan tests to be performed on the inop rable Ioopduring 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 b 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 loo s.

I NET 3 I 0

An OPERABLE DHR loop is composed of an OPERABLE DHR pump and an OPERABLE DHR heat e-fthan el. 0 DHR pumps are OPERABLE if they are capable of being powered and 5are INSERT 5 able to provide flow if required./A SG can perform as a heat sink]

when it as an a uate 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, "RCSLoops - MODE 3V' "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 LCO 3.9.5, Water Level' M'. 6y-and "Decay Heat Removal (DHR) and Coolant Circulation - Low 0

Water Levely ACTIONS A.11. A.2. 81>-nd q.2 * .

foneDHR loop is OPERABLE and an yrequired SG has sec 'ndary side wtelevel <[5W*/o or one required DAFR loop inoperab~le,/redundancy for 0

heat removal is lost. Action must be initiated to restore a secondH-,H-R-loop to OPERABLE statusler initiate//ction to restore te s/ condary side.

[wate-r evel in t te SGs, and action/trhust be taken immediiely. ýEite

!Required Act' n will restore redtffidant decay heat rem~o al path . 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, 12/01/05 Attachment 1, Volume 9, Rev. 1, Page 151 of 418

Attachment 1, Volume 9, Rev. 1, Page 152 of 418 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-1 1 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. 1, Page 152 of 418

Attachment 1, Volume 9, Rev. 1, Page 153 of 418 RCS Loops - MODE 5, Loops Filled B 3.4.7 BASE;S ACTIONS (continued) FBý If no required R] loop is in operation, lexcelf as proyided in dote 1 or no required D-R 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 3.1.1 must be suspended and action to restore a 0PR 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. W'th.coolant added withouto d 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 REQUIREMENTS SR This 3.4.7.1 SR requires verification every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months that the [ RCS or require DDHR loop is in operation. Verification includes flow rate, temperature, or um status iornatural monitoring, which help ensure that force flow is providing heat removal, circulation 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 rqe Verifying the, SGs are OPERABLE by ensuring their secondary side water 05 35 inches abovethe lowertube sheet levels are7[5- % ensures that redundant heat removalpaths are availablelif the seeond DHR loop is rot OPERABLE. If both DHR loops 10 are OPERABLE, this Surveillance is not needed. The 12, hour Frequency has been shown by operating practice to be sufficient to regularly assess degradation and verify operation OMthin ,fety analyses ,ssumptionsI BWOG STS B 3.4.7-4 Rev. 3.1, 12/01/05 Attachment 1, Volume 9, Rev. 1, Page 153 of 418

Attachment 1, Volume 9, Rev. 1, Page 154 of 418 RCS Loops - MODE 5, Loops Filled 8 3.4.7 BASES SURVEILLANCE REOUIREMENTS (continued)

SR 3ý4.7.3 Verification that each required DHR pump is OPERABLE ensures that redundant paths for heat removal are available. The requirement also

ýensures 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 Y secondary side water level is 1-[5 in 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.

BXAOG STS B 3.4.7-5 Rev. 3.1, 12/01/05 Attachment 1, Volume 9, Rev. 1, Page 154 of 418

Attachment 1, Volume 9, Rev. 1, Page 155 of 418 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. 1, Page 155 of 418

Attachment 1, Volume 9, Rev. 1, Page 156 of 418 Specific No Significant Hazards Considerations (NSHCs)

Attachment 1, Volume 9, Rev. 1, Page 156 of,418

Attachment 1, Volume 9, Rev. 1, Page 157 of 418 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. 1, Page 157 of 418

Attachment 1, Volume 9, Rev. 1, Page 158 of 418 ATTACHMENT 8 ITS 3.4.8, RCS LOOPS - MODE 5, LOOPS NOT FILLED Attachment 1, Volume 9, Rev. 1, Page 158 of 418

, Volume 9, Rev. 1, Page 159 of 418 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) , Volume 9, Rev. 1, Page 159 of 418

Attachment 1, Volume 9, Rev. 1, Page 160 of 418 ITS 3.4.8 ITS 3/I4.& RtACTOR COOLA.,'T STS-7"M iIEUTDO* N A! HOT OSTONES T LfIN COND~ITION FOR 0PEWIOlN

_________________________________________________ j LA01 3.4.1.2 a. At least CV0 Coolant loops 1 shal ba \

LCO 3.4.8 OflLABIZ:

See ITS

1. Reactor Coolant Loop 1 and its associated Stea.

3.4.5, gentTato** ITS 3.4.6, and

2. R* actor Coolant Loop 2 and :Its assoclated Ptam ITS 3.4.7 tenetato?.

ACTION A ACTION B LCO 3.4.8 NOTE 1 I no operations are permitted that-would cause dilution C coolant system boron concentration'*,and (2) core outlet is maintained/at least 10F below saturation temperaturt Page 1 of 2 Attachment 1, Volume 9, Rev. 1, Page 160 of 418

Attachment 1, Volume 9, Rev. 1, Page 161 of 418 ITS 3.4.8 ITS 3/4.4 REACTOR COOLANT SYSTEM SURVEILLANCE REOUIREMENTS ./ -

lp(s) shall be dIter ained OPERABLE. 3.45, I t .4.1.2.1 The re~ui red decay heat re moa Oer Specificati 4.0.5. 7 an, 4.4.1.-*2.2 Th reuie sta generator(s) shall be determined OPERMABLE by IT 3.4.

verifinj secondary -side level to be greater than or equal t (a) 18 inch SeeTS 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 3.4.5 and pupis..(b) 35 inches above the lover tube sheet once per 12" ITS3.4.6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months it no reactor coolant pumps are operating. .See ITS

.. ..... 3.4.7 SR 3.4.8.1 4.4.1.2.3 At least one coolant loop shall be verified to be in operation and 34 Icircu ating r actor oant:at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . per 12 hours

~~Add proposed SR 3.48.2]

DAVIS-BESSE, UNIT 1 3/4 4-2a Amendment No. ,3,135 Page 2 of 2 Attachment 1, Volume 9, Rev. 1, Page 161 of 418

Attachment 1, Volume 9, Rev. 1, Page 162 of 418 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. 1, Page 162 of 418

Attachment 1, Volume 9, Rev. 1, Page 163 of 418 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. 1, Page 163 of 418

Attachment 1, Volume 9, Rev. 1, Page 164 of 418 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 10OF 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 100 F below saturation temperature is 2021F. This change is acceptable because the proposed change increases the margin to the boiling point since it requires the maximum RCS temperature be < 190 0 F. 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. 1, Page 164 of 418

Attachment 1, Volume 9, Rev. 1, Page 165 of 418 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. 1, Page 165 of 418

Attachment 1, Volume 9, Rev. 1, Page 166 of 418 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 1 - 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 Required Action) 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. 1, Page 166 of 418

Attachment 1, Volume 9, Rev. 1, Page 167 of 418 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. 1, Page 167 of 418

Attachment 1, Volume 9, Rev. 1, Page 168 of 418 Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs)

Attachment 1, Volume 9, Rev. 1, Page 168 of 418

Attachment 1, Volume 9, Rev. 1, Page 169 of 418 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) loopsshall be OPERABLE and one DHR loop shall be in operation.

-- .................................------ NOTES --------------- .........-----------------

3.4.1.2 1 All DHR pumps may be removed from operation for < 15 minutes footnote when switching from one loopto another provided:

a.The maximum RCS temperature i*[Ff , 1

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 3.1.1 and '

[,SHUTDOWN MARGIN (SDM);" <"-

c, No draining operations to further reduce the RCSwater volume are permitted.

DOC L01 2. One DHR loop may be inoperable for <2 hours for Aurveillance 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 i +

Action a A. One required DHR loop A.1 Initiate action to restore Immediately inoperable. DHR loop to OPERABLE status.

BWOG STS 3.4.8-1 Rev. 3.0, 03/31104 Attachment 1, Volume 9, Rev. 1, Page 169 of 418

Attachment 1, Volume 9, Rev. 1, Page 170 of 418 CTS RCS Loops - MODE 5, Loops Not Filled 314.8 ACTIONS (continued)_

CONDITION REQUIRED ACTION COMPLETION TIME loop B. No required DHR

+

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 31_1.

in operation.

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

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY 4.4.1.2.3 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.

BV*OG STS 3.4.8-2 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 1, Page 170 of 418

Attachment 1, Volume 9, Rev. 1, Page 171 of 418 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. 1, Page 171 of 418

Attachment 1, Volume 9, Rev. 1, Page 172 of 418 Improved Standard Technical Specifications (ISTS) Bases Markup and Justification for Deviations (JFDs)

Attachment 1, Volume 9, Rev. 1, Page 172 of 418

Attachment 1, Volume 9, Rev. 1, Page 173 of 418 RCS Loops -. MODE 5,.Loops Not Filled B 3.4.8 B13.4 REACTORCOOLANTSYSTEM (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 4 decay heat and transfer of this heat to the decay heat I coolers removal (DHR) heat c hangerj.. 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. RCS draining is initiated (hot legs not completely filled).

r LAdditionally, the RCS inventory is further reduced to a water level[jl]within the 0

m Loops are not filled when the Li horizontal portion of the hot legas miglit be the case for refueling or maintenance on the reactor coo1antpumps orSGs. 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 airbound and fail resulting in a loss of forced flow for heat removal. As a consequence the waterin the~core will heat up and could boil with the possibility of core uncoverirng 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 thisLCO is to provide forced flow from at least one DHRpump 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.byone 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. Heatcannot be removed via the DHR Astem unless forced flow is used. A minimum of one running decay heat removal pump meets the LCO requirement for one loop in 0 operation. An additional DHR loop is required to be OPERABLE to provide redundancy for heat removal.

BWOG STS B3.4.8-1 Rev. 3.0, 03/31/04 Attachment .1, Volume 9, Rev. 1, Page 173 of 418

Attachment 1, Volume 9, Rev. 1, Page 174 of 418 RCS Lo6ps - MODE 5I Loops..Not -Filled 153.4.8 BASES LCO (continued) 4 Note 1 permits the DHR pumps to. be removed from operation for th- [1'5 minutes when switching from one in othe otr other. The I * (

circumstances for stopping both .DHR pumps:are.to be limited to situations, where the outage time is shortland temperature is maintained(

6Fl. The Note prohibits- boron dilution with coolant at boron ee concentrations less than required to *sure the SDMof LCO 3.1.1,i and o draining operationsl when QHR. forced flow iý'stopped.

that could reduce the RCS water volume 0-Note 2 allows one DHR loop to be inoperable for.a period of2 hours 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 DH R. pump capable of providing forced flow to an OPERABLE DHRI eat ccange.

er DHR pumps are OPERABLE if they are capable of being powered and INSERT1IV- are able to provide flow if required. INSERT 2 l 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, "ROS Loops - MODES 1 and LCO 3.4.5, "RCS Loops - MODE LCO 3.4.6, "RCS Loops - MODE LCO 3.4.7, "RCS Loops - MODE 5, Loops FilledJ LCO 3.9.4, "Decay Heat Removal (DHR) and Coolant Circulation - High Water Level" M DE 6 fand(

LCO 3.9.5, "Decay Heat Removal (DHR) and Coolant Circulation - Low Water LeveF E*M 0 ACTIONS A.1 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. 1, Page 174 of 418

Attachment 1, Volume 9, Rev. 1, Page 175 of 418 B 3.4.8 0 INSERT I 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-1 1 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. 1, Page 175 of 418

Attachment 1, Volume 9, Rev. 1, Page 176 of 418 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 inoperation, except as provided by Note 1 in the LCO, the Required Actio require) immediate suspension of all operations involving introduction of coolant into the RCS with boron concentration less than required to meet the ________

minimum $DM of LCO 3.1 .1 and ]s of action to

'initiation Iimr diatel restore one DHR loop to OPERABLE status and operation.

The Re'ed Action for resto tion does not apply ý6 the condition of both 1 ps not in operation nen the exception N (e in the LCO is in forqe<1 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 3A4.8.1 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 availabilityý 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.

BVWOG STS B 3.4.8-3 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 1, Page 176 of 418

Attachment 1, Volume 9, Rev. 1, Page 177 of 418 RCS.Loops - MODE 5, Loops Not Filled B 3.4.8 BASES REFERENCES 1. Generic: Letter.88-17, October 17, 1988.

BVAOG STS B 3.4.8-4 Rev. 3.0, 03131/04 Attachment 1, Volume 9, Rev. 1, Page 177 of 418

Attachment 1, Volume 9, Rev. 1, Page 178 of 418 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.

0 Davis-Besse Page 1 of 1 Attachment 1, Volume 9, Rev. 1, Page 178 of 418

Attachment 1, Volume 9, Rev. 1, Page 179 of 418 Specific No Significant Hazards Considerations (NSHCs)

Attachment 1, Volume 9, Rev. 1, Page. 179 of 418

Attachment 1, Volume 9, Rev. 1, Page 180 of 418 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.

0 Davis-Besse Page 1 of 1 Attachment 1, Volume 9, Rev. 1, Page 180 of 418

Attachment 1, Volume 9, Rev. 1, Page 181 of 418

ATTACHMENT 9 ITS 3.4.9, PRESSURIZER Attachment 1, Volume 9, Rev. 1, Page 181 of 418

, Volume 9, Rev. 1, Page 182 of 418 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) , Volume 9, Rev. 1, Page 182 of 418

Attachment 1, Volume 9, Rev. 1, Page 183 of 418 ITS 3.4.9 ITS REACTOR COOLANT SYSTEM PRESSURIZER UbMQNG CONDITION FOR OPERATION LCO 3.4.9 3.4.4 The pressutzer shad be OPERABLEwft A02 LCO 3.4.9.a &P b. P A water level two4 and 228 ---

inches.

Add mpropoed LCO 3.4.9.b PpLICABIUTY: MODES 1 and 1 -Add proposed MODE 3 Applicabiliy M 02 U' k4* 0 , e.ubg..,.,,VWII n~rpstnortheoressuizertoOPERABLEstatusr withi Iho!u.-o-ACTION B

{Add proposed ACTIONS C and 0J M* l SR 3.4.9.1 4A.4 The presswtnw shad be demonstrated OPERABLE by vertym presawIzer leel to be v%1hMn Uluti at least once per 12 hou(S.

I~dd ropoed S M01 Amendment No. 255 DAVIS-BESSEF UNIT 1 3/44-5 Page 1 of 1 Attachment 1, Volume 9, Rev. 1, Page 183 of 418

Attachment 1, Volume 9, Rev. 1, Page 184 of 418 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 85 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. 1, Page 184 of 418

Attachment 1, Volume 9, Rev. 1, Page 185 of 418 DISCUSSION OF CHANGES ITS 3.4.9, PRESSURIZER capacity of pressurizer heaters is less than 85 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 85 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 1 - 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. 1, Page 185 of 418

Attachment 1, Volume 9, Rev. 1, Page 186 of 418 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. 1, Page 186 of 418

Attachment 1, Volume 9, Rev. 1, Page 187 of 418 Improved Standard Technical Specifications (ISTS) Markup 0 and Justification for Deviations (JFDs)

Attachment 1, Volume 9, Rev. 1, Page 187 of 418

Attachment 1, Volume 9, Rev. 1, Page 188 of 418 CTS Pressurizer 3.4.9 3.4 'RIEACTOR COOLANT SYSTEM (RCS) 3.449 .Pressurizer LCO 3.4.4 LCO 3.4.9 The pressurizer shall be OPERABLE with-.

LCO 3.4.4.b a. Pressurizer water level kinhtes ands essential 00 DOC M01 b. A minimum of j[l kW of pressurizer eaters OPERABLE n--

capable of being powered -from an emergeny. power supply

-- - - - - - - - -W- - - - -- -- ----- - - - - NO T E - ---------

OPERABILITY req irements on pressurizer heaters o not apply in MODE 4. 0 APPLICABILITY: MODES 1, 2, and 3M IMODE-4>wth RCS tempe/ritue Ž [275]°F1.

ACTIONS'

}0 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 41wit terpature 5ý5]oF.-I- -

CSI [24 hours 0

~tal DOC M01l - C.

yD F.p

ýCapacity of pressuri~zer CA Restore pressurizer heater 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months bein powere by eme genc wer essential 0

sup ly'] less than limit.

BWVOG STS 3.4.9-1 Rev. 3.0, 03131/04 Attachment 1, Volume 9, Rev. 1, Page 188 of 418

Attachment 1, Volume 9, Rev. 1, Page 189 of 418 CTS Pressurizer 3.4.9 ACTIONS (continued)2 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 < inches. 1 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]

fe kWof pre urizer heaters aree *[1] months M capa e of being powe d from an emergency2 "tpo~pvdrysupplyý . 24.. 00

SR :34.93 [a:$ i emergency power supply for pressurize heiters is OPERABLE. /

[18] months]

01 capacity of essential pressurizer heaters is > 85 kW.

BWVOG STS 3.4.9-2 Rev. 3.0, 03/31/04 0

Attachment 1, Volume 9, Rev. 1, Page 189 of 418

Attachment 1, Volume 9, Rev. 1, Page 190 of 418 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 in ISTS LCO 3.4.9.b and ISTS SR 3.4.9.2 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). In addition, since the ISTS SR 3.4.9.3 Bases states that the SR is not applicable if the heaters are permanently powered by 1 E power supplies, ISTS SR 3.4.9.3 has been deleted.

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 2800 F. In the ISTS, the temperature cross-point is 330 0 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. 1, Page 190 of 418

Attachment 1, Volume 9, Rev. 1, Page 191 of 418 -

Improved Standard Technical Specifications (ISTS) Bases Markup and Justification for Deviations (JFDs)

Attachment 1, Volume 9, Rev. 1, Page 191 of 418

Attachment 1, Volume 9, Rev. 1, Page 192 of 418 Pressurizer B3.4.9, B1314 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 saturatedoconditions for pressure control purposes to. prevent bulk boiling in the remainder0of the RCS. Key.

functions include. maintainingýrequired primary system pressure, during steady. state operation and limiting the pressure changes caused by reactor coolant thermal expansion and contraction during normjal!load transients.

an The pressure control components addressed by this LCO include the, pressurizer water leve the required heatersj, and theirontrols and lemergency per supplies Pressurizer safety valves andOpressurizer the pilo por operated relief val Vefl(PORV) are addressed by. LCO'34.10, "Pressurizer Safety Valves," and LCO 3.4.11; "Pressurizerr Operated Relief Valve (PORV)" respectively.

The maximum water level limit has been established to ensure that a liquidto vapor interface exists to permit R(S 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 transporttand

b. By restricting the level, to a maximum, expected transient reactor coolant volume increases (pressurizer insurge) will not cause excessive level changes thatcould 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 levellimit 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. If the 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.

BWOG STS B 3.4.9-1 Rev. 3.0, 03/3,1104 Attachment 1, Volume 9, Rev. 1, Page 192 of 418

Attachment 1, Volume 9, Rev. 1, Page 193 of 418 Pressurizer There are two essential heater banks, R 3.4.9 BASES with each bank powered from a separate essential bus and each bank having a capacity of 126 kW.

0 BACKGROUND (continued) 0 The pressurizer heaters are used to maintain a pressure in the RCS so reactor coolant in the loops is.subcooled and thus int.he preferred state,ý for heat.transport to the steamgenerators (SGs). Thisfunction must be maintained with a loss of offsite power. Consequently, the emphasis. of this LCO is to ensure thatthe 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 [161kW ensures that the RCS 0

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 asteam bubble is reflected SAFETY implicitly in the accident analyses. INo safety analys esare performedir ANALYSES [iowerMODES. All analyses performed from a critical reactor condition assume. the existence of a steam bubble and saturated conditions in the, 0

pressurizer. In making this assumption, the analyses neglect the small:

fraction of noncondensible gases normally present.

Safety analyses presented inthe 4FSAR do not take credit forpressurizer.

heater operation; however, an implicit initial condition assumption ofthe 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 thatit 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 isnot adjusted for instrument error.

BV*OG STS B 3.4.9-2 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 1, Page 193 of 418

Attachment 1, Volume 9, Rev. 1, Page 194 of 418 Pressurizer B 3.4.9 BASES APPLICABLE SAFETY AN.ALYS ES (continued)

Evaluations performed for the design basis large break loss of coolant, accident (LOCA), which assumed abhier aximum.evel than assumed for the LOMFW event have been made. The higher pressurizer level assumed for th LA isthe basis for the volume of reactor coolant released to the, containment. The.containrient analysis performed using 0

the mass andenergy 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 maintainingthe 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 isan initial condition or coincident event assumed in many accident analyses, maintaining hot, high pressure conditions over an extended time period is. notevaluated as u

part of FSAR'accident analyses. 0 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' subcobling in the long term during loss ofoffsite power., as'indicated in NUREG-0737 (Ref. 1), is the reason:

for providing'an LCQ.

LCO The LCO requirement forthe pressurizer to be OPERABLE with a water level [20inchesensuresthat 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. tial Sic ahesnil The LCO.requires OPERABLE Iand/apable~e of [16_kW a minimumbeing poweredoffrom pressurizer heaters powerI an/emergency (

Ibank has acapability of :t1 1126 MW either essential __,Jsupply]..As suel , the LCO addresses both the heat/*rs and the power bank can be used to Isupplies.] The minimum heater capacity required is sufficient to maintain l ~eet, the LCO the system 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 ofJ 2EqJk is derived from 0(

the use of nine. heaters rated at 14 kWeach. 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. 1, Page 194 of 418

Attachment 1, Volume 9, Rev. 1, Page 195 of 418 Pressurizer B 3.4.9 BASES APPLICABILITY The need for pressure contro is. most pertinent when core heat can cause the greatest effect on RCS temperature., resulting inthe 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 MOE3and-, for yressurizer water leveljor MODEA with RC-S]

ternerature Ž [27,]1 . 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 pump startup:. The t mperature of [275];F has been: Iesignated as the cutoff forapp icabili because LCO3.4.1 2, Low To perature Overpressure Prote tion (LTOP) SyStem,' provides requirement for 0

pressurizer level btow [275]1F. The LCO'does not apply to MODE 5.with loops filled because LCO 3A4,12 applies. The LCO does not apply to MODES 5 and 6 vith partial loop operation.

In MODES 1', 2,.and 3, there is the need to maintain-the availability of pressurizer heaters capable of being powderedfrom 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 subco6ling.foran extended period. To Applicability is modified by a.Note stating that'the' OPERABILI requirements on pressurizer heaters do not apply in MODE 4.\ For 0

MODE 4, 5,.or 6, it is not necessaryto control pressure.(by heaters) to ensure loop subcooling for heatftransfer whoe:the Decay Heat Removal System is in service, and therefore the. LCOis notapplicable.

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 restoredreducing core power constrains heat input effects that drive pressurizer insurgeithat could result from an anticipated transient. Byshutting domthe reactor and r ucing reactor coolant tempaerture to at leastM9 E 3, the potential ermal energyof the reactorroolant mass~for L CA.mass and energ releases is reduced.

INSERT 1 0

BWOG STS B 3.4.9-4 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 1, Page 195 of 418

Attachment 1, Volume 9, Rev. 1, Page 196 of 418 B 3.4.9 0 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. 1, Page 196 of 418

Attachment 1, Volume 9, Rev. 1, Page 197 of 418 Pressurizer B3.4.9 BASES ACTIONS :(continued)

Six hours isa reasonab e time based upon operating e eiience.to re,.ch MODE 3 fromfull po rwithout challenging plant syst ms and operators.

Further pressure and emperature reduction to MODE 4 with RCS temperature < [275I/1 places the plant into a MODE ere.the LCO is not applicable. The [2 hour Completion Time to reach he nonapplicable MODE is reasona le based upon operating experie ce.

CA' Ifrthe [emergen power sup s to the heaters a enot capable ofl]

roviding [takW, orlthe*pressurizer heaters are inoperable, restoration 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 essential power will not occur in this period. Pressure control may be maintained during this time. using normal station powered heaters.

D1 :and D.2

-- essentialI If16 ressurzer heater capabilitycannot 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, theplant 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[t~e IJ-fo~l!ng 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 order!y 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. hour 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.

BVVOG STS B 3.4.9-5 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 1, Page 197 of 418

Attachment 1, Volume 9, Rev. 1, Page 198 of 418 Pressurizer B 3.4,9 BASES SURVEILLANCE REQUIR EMENTS (continued)

MSR, 3.4,9.2 0 SR: requires.,the power supplies are capable of producing the minimum power and the-Iasse iatedl pressurizer heaters are verifiedto 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; Frequencyo m monthsMis considered adequate 0 to detect heater degradation and has been shown by operating experience tobe acceptable., ] 0

[SR 3.4;9.3 This SR is not a plicable ifWthe heaters are per, nently powered by IE power suppliS '

This Surveillanc demonstrates that the heater can be manually 0

transferred to,* a d energized by, emergency p er supplies. The Frequency of [1 :months is based Ona typical uel cycle and is consistent with irhilartverificati6nýs of emergen y power. ]

REFERENCES 1. NU.REG.-0737, November 1980.

BWOG STS B 3.4.9-6 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 1, Page 198 of 418

Attachment 1, Volume 9, Rev. 1, Page 199 of 418 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. 1, Page 199 of 418

Attachment 1, Volume 9, Rev. 1, Page 200 of 418 Specific No Significant Hazards Considerations (NSHCs)

Attachment 1, Volume 9, Rev. 1, Page 200 of 418

Attachment 1, Volume 9, Rev. 1, Page 201 of 418 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. 1, Page 201 of 418

, Volume 9, Rev. 1, Page 202 of 418 ATTACHMENT 10 ITS 3.4.10, PRESSURIZER SAFETY VALVES , Volume 9, Rev. 1, Page 202 of 418

, Volume 9, Rev. 1, Page 203 of 418 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) , Volume 9, Rev. 1, Page 203 of 418

Attachment 1, Volume 9, Rev. 1, Page 204 of 418 ITS 3.4.10 ITS REACTOR COOLANT SYSTEM SAFETY VALVES AND PILOT OPERATED RELIEFVALVE - OPERATING LIMITING CONDITION FOR OPERATION LCO 3.4.10 3.4.3 All pressurizer code safety valves shall be OPERABLE with a lift setting of < 2525 ps.iq* JWhen not isolated, the pressurizer pilot operated 1 SeeITS relief valve shall have a trip setpoint of k.2435 psig and an allowable value*

[of >2435 psig.'**

3.4:11 J APPLICABILITY: MODES 1, 2 and 3.

ACTION:

ACTION A -- W-ith one pressurizer code safety valve inoperable, either restore the tnoperable Valve to OPERABLE status within 15 minutes o*r+be in HOT SHUTDOWN ACTION B Wi n ýours.

Add proposed Required Action B.1 Mot Add proposed ACTION B for two pressurizer safety valves inoperable. M02 SURVEILLANCE REQUIREMENTS I,->

SR3.4.10.1

" l Add 4.4.3 For the pressurizer codeproposed reset limit__

safety valves, there are no additional ýM )

Surveillance Requrements otherthan hose required b Specification 4.0.5S ITS For the pressurizer pilot operated relief valve a CHANNEL CALIBRATION check see.ITS Ishall be performed each REFUELING INTERVAL.34 LA01

The lift setting pressure shall rrespond to ambient copditions: of the valve nominal operating te rature'and pressure.

See ITS

- Allowable value for CHANNEL CALIBRATION check. 3.4.1 1J DAVIS-BESSE, UNIT I 3/4 4-4 Amendment No. , I 1.218, Page 1 of 1 Attachment 1, Volume 9, Rev. 1, Page 204 of 418

Attachment 1, Volume 9, Rev. 1, Page 205 of 418 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. 1, Page 205 of 418

Attachment 1, Volume 9, Rev. 1, Page 206 of 418 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 valvb 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. 1, Page 206 of 418

Attachment 1, Volume 9, Rev. 1, Page 207 of 418 Improved Standard Technical Specifications (ISTS) Markup 0 and Justification for Deviations (JFDs)

Attachment 1, Volume 9, Rev. 1, Page 207 of 418

Attachment 1, Volume 9, Rev. 1, Page 208 of 418 Pressurizer Safety..Valves.

CTS.

3.4.10 3.4 REACTOR COOLANT SYSTEM (RCS) 3.410 Pressurizer Safety'Valves 3.4.3 LCQ 3.4.10 Two pressurizer safety valves shall be OPERABLE with, lift settings:

1Ž4[2475 sigand<*252,!qpsig. 0 APPLICABILITY: MODES 1, 2, and f IMODE 4 with l11 RCS. cold leg temperajures > [283]°FL 0

The lift settings arpe iQt required to be within the LCO iipits for entry into MODES:3 and 4 for the purpose of setting the pressuri er safety valves under ambient (hot conditions... Thisexception is allov'edfor[36] hours 0

following entry into IODE 3 provided a preliminary col 1 setting was made:

prior.to heatup.

ACTIONS CONDITION REQU IRED ACTION COMPLETION TIME Action A. One pressurizer safety AX1 Restorevalveto 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 wit ny fl []hours 0 RCS old leg temp ratur Two pressurizer safety s-.[28]F, DOC M02 valves inoperable.

BWOG STS 3.4.10-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 1, Page 208 of 418

Attachment 1, Volume 9, Rev. 1, Page 209 of 418 CTS Pressurizer Safet"yValves 3.4.10 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY 4.4.3 SR 3.4.10.1 Verify each pressurizer safety valveis OPERABLE In accordance in accordance with the Inservice Testing Program. with the, Inservice Following testing, lift settings shall be within + 1%. Testing Program BWG STS 3.4.10-2 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 1, Page 209 of 418

Attachment 1, Volume 9, Rev. 1, Page 210 of 418 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 > 2831F 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 330 0 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. 1, Page 210 of 418

Attachment 1, Volume 9, Rev. 1, Page 211 of 418 Improved Standard Technical Specifications (ISTS) Bases Markup and Justification for Deviations (JFDs)

Attachment 1, Volume 9, Rev. 1, Page 211 of 418

Attachment 1, Volume 9, Rev. 1, Page 212 of 418 Pressurizer Safety Valves B 3.4;10 B 3.4 REACTOR COOLANT SYSTEM (RCS)

B 14-10 Pressurizer Safety Valves BASES BACKGROUND The purpose of the two spring loaded pressurizer safety valves is to provide RCS overpressure protection. Operating in conjunctioh with the Reactor Protection System (RPS), two valves are used to ensure that the p n Safety Limit (SL) of 2750 psig is: not exceeded for analyzed transients oduring operation in MODES 1 and 2. Two safety valves are used for S 4 ad"MODE 3 and portiogs-' MODE 41 For the remainder of MODE[A,---

MODE* 5, and MODE 6 with the reactor head on, overpressure protection 0 is provided by operating procedures and LCO 3.4.12, "Low Temperature Overpressure Protection (LTOP) ye 0

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 (Reft 1). The required lift pressure00 psig. 2i si i%. 0 The amet condiscaasociated withe pressur zernto requenchtank located in the c stainment. The docharge flow is indicateenby an increase in tedperature downstere of the safety valvep and by an 0

dr p s incrgse 1stablishe d.htank ttyfimerature and level, rsu directly into-a opening separate teeFlow into acontainment. through the requirement for lifting pressures above t000 psig. The lift setting is for pressurizer safety valves generates l the ambient conditions associated with MODES 1, 2, and 3.*This requires acoustic levels or vibration that is l either that the valves be set hot or that a correlation between hot and cold eetd pipe. These sensors provideI bypeolcrcsnosnthischarge settings be established. [(nominal operating temperature and pres-sure) ]-

valve position indication (open/closed) inI the control room. The pressurizer safety valves are part of the primary success path and 0

mitigate the effects of postulated accidents. OPERABILITY of the safety valves ensures that the RCS pressure will be limited to 110% of design 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. r-,

0 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, 12101105 Attachment 1, Volume 9, Rev. 1, Page 212 of 418

Attachment 1, Volume 9, Rev. 1, Page 213 of 418 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 from a startup accident establishes the minimum-safety valve capacity. The subcitic startup accident is assumed to occur at < 150 poweo. Single failure of a 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 assumptionsland to comply with Code requirement . The IupPer jd lower pressur tole nce Iimitf1 e ased on theN D1 tolerance requirements (R for lifting pressuresabove 1000 psig.

The limit protected by this Specification is the reactor coolant pressure 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 limitcould 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, and portios of MODE 4 abovoe LTOP cut in ltempe r5ature, 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 rtion

-conservatively included, although the listed accidents may not require both safety valves for protection.

The LCO is not applicable in MODE 41when anRCS cold leg temperatup is !- [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 with the lift sett* gs outside (I1 the LCO limits, is permits testing an examination oft safety. valves at high pressur/ and temperature ne their normal op ating range, but only after thtalves have had a pr iminary cold sett* g. Thecold setting gives assu 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. 1, Page 213 of 418

Attachment 1, Volume 9, Rev. 1, Page 214 of 418 Pressurizer Safety Valves B 3.4.10 BASES LCO (continued) condition. Only e valve at a time will removed from rvice 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 derive-rom operating 0

experienc 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 B.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 4Fwith any RCS cold leperature _ [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 orderly manner and without challenging plant systems. Similarly, the allowed is reasonable, based on operating experience, to reacn MODE 4 ours 0 without challenging plant systems. With any RCS cold le$ temperature at or below [283]°F, 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, and thereby removps the need for overpressure protect n by two pressurizer safety ,alves.

SURVEILLANCE SR 3.4.10.1 0

REQUIREMENTS SRs are specified in the Inservice Testing Program. Pressurizer safety Fm valves are to be tested in accordance with the requirements of the ASME Code (Ref.ýj), which provides the activities and the Frequency necessary

[3 ýosafisfy'e SRs. No additional requirements are specified.

in accordance with LReference1I 00 The pressurizer safety valve setpoint is *__31°,.for OPERABILITY;. R F

however, the valves are reset to +/- 1% during the Surveillance to al wfor 0

REFERENCES ASME Code for Operation and Maintenance of Nuclear Power Plant1. 0

1. ASME Boiler and Pressure Vessel Code,Section III. ,1995 Edition with 1996 Addenda BWOGSTS B 3.4.10-3 Rev. 3.1, 12/01/05 Attachment 1, Volume 9, Rev. 1, Page 214 of 418

Attachment 1, Volume 9,, Rev. 1, Page 215 of 418 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. 1, Page 215 of 418

Attachment 1, Volume 9, Rev. 1, Page 216 of 418 Specific No Significant Hazards Considerations (NSHCs)

Attachment 1, Volume 9, Rev. 1, Page 216 of 418

Attachment 1, Volume 9, Rev. 1, Page 217 of 418 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. 1, Page 217 of 418

Attachment 1, Volume 9, Rev. 1, Page 218 of 418 ATTACHMENT 11 ITS 3.4.11, PRESSURIZER PILOT OPERATED RELIEF VALVE (PORV)

Attachment 1, Volume 9, Rev. 1, Page 218 of 418

, Volume 9, Rev. 1, Page 219 of 418 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) , Volume 9, Rev. 1, Page 219 of 418

Attachment 1, Volume 9, Rev. 1, Page 220 of 418 ITS 3.4.11 ITS REACTOR COOLANT SYSTEM

.SAFETY VALVES. AND PILOT OPERATED RELIEF VALVE - OPERATING LIMITING CONDITION FOR OPERATI.ON LCO 3.4.11 3.4.3 All pressurizer code safet -evl e s..al beOPERABLE valvessa e OEt. with a Iýift See ITS 3.94.10 Fsettin of. < 252S psi ..*ý When not isolated, the pressurizer pilot operated shalbeOPERABLE relief valv-,Sha yea rip setp -owa n o - psig-n-an a vau - A1 lot AZ435 ptg.**__ . .

APPLICABILITY: MODES 1, 2 and 3. *LA02*

AUT ON:

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

3SeT10 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

Add proposed I Add ACTIONS A, B, and C proposed ACTIONS A, B, and C MOl SURVEILLANCE REQUIREMENTS 4.4.3 For the.pressurizer code safety valves, there are no additional __See ITS Surveillance Requirements other than those required 'by Specification 4.0.5. I 3.4.10 For the pressurizer pi ot operated rellief valve a cALIBRA.O.

cHANNeL thec eapropEFUELING INT RVALnd.

bederformed Ashall 3U02 I I_[Add proposed SIR 3.4.11.1 and SIR 3.4.1 1f2 } M02*

The lift setting pressure shall correspond to ambient conditions of the valve at nominal operating temperature-and pressure. H See ITS 3.410 Allowable value for CHANNEL CALl TION check.A DAVIS-BESSE, UNIT I 3/4 4-4 Amendment No. 33,60,128,136, 218 Page 1 of I Attachment 1, Volume 9, Rev. 1, Page 220 of 418

Attachment 1, Volume 9, Rev. 1, Page 221 of 418 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. 1, Page 221 of 418

Attachment 1, Volume 9, Rev. 1, Page 222 of 418 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 requirement to 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. 1, Page 222 of 418

Attachment 1, Volume 9, Rev. 1, Page 223 of 418 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. 1, Page 223 of 418

Attachment 1, Volume 9, Rev. 1, Page 224 of 418 Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs)

Attachment 1, Volume 9, Rev. 1, Page 224 of 418

Attachment 1, Volume 9, Rev. 1, Page 225 of 418 CTS Pressurizer PORV 3.4.11 3.4. REACTOR COOLANT SYSTEM (RCS) 3.4.11 Pressurizer [Po rOperated 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 REQUIREDACTION COMPLETION TIME DOC M01 A, PORV inoperable. A.1 Close blockval e. 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months AND A.2 Rernovepower from block .1hour valve.

DOC M01 B. Block valve inoperableý BA1 Close block valve.ý 1hour 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 M01 C. Required Action and C.1 BeinMODE 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 BWAOG STS 3.4.11-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 1, Page 225 of 418

Attachment 1, Volume 9, Rev. 1, Page 226 of 418 CTS Pressurizer PORV, 3:4.11 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY DOC M02 SR 3.4.1:1.1 ----------------------------- -NOTE----7, - - -

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. mont s 0 SR 3.4.11.3 [ Verib PORV and block valve are capable ofbei g 18mrnonths.]

0 powe ed from an emergency power source.

BWVOG STS 3.4.11-2 Rev. 3:0, 03/31/04 Attachment 1, Volume 9, Rev. 1, Page 226 of 418

Attachment 1, Volume 9, Rev. 1, Page 227 of 418 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. 1, Page 227 of 418

Attachment 1, Volume 9, Rev. 1, Page 228 of 418 Improved Standard Technical Specifications (ISTS) Bases Markup and Justification for Deviations (JFDs)

Attachment 1, Volume 9, Rev. 1, Page 228 of 418

Attachment 1, Volume 9, Rev. 1, Page 229 of 418 Pressurizer PORV B3.4.11 B 3.4 REACTOR COOLANT SYSTEM RCS)o B 3.4.11 Pressurizer Pa ]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 essential buses, which are powered from either the offsite The PORV, its block valve, and their controls are powered from Fn] powe pii~esnowet1'uie l~ut are also ca.le, powered fabmrejmergency _

p or the u . Power supplies for the PORV are separate from those for the block valve. Power supply requirements are defined in NUREG-0737, Paragraph G.1 (Ref. 1). 2435 psig)

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.

INSERT 1 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 - Rev. 3.1, 12/01/05 Attachment 1, Volume 9, Rev. 1, Page 229 of 418

Attachment 1, Volume 9, Rev. 1, Page 230 of 418 B 3.4.11 OINSERT 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. BAW-1890, September 1985 (Ref. 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. 1, Page 230 of 418

Attachment 1, Volume 9, Rev. 1, Page 231 of 418 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[dlo not take credit for PORV actuation, but]do take credit for the safety valves.

The PORV also provides low temperature overprejssure protection (LTOP) during hea up and cooldown. LCO 3.4.12, "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. 1, Page 231 of 418

Attachment 1; Volume 9, Rev. 1, Page 232 of 418 i Pressurizer PORV B 3.4.11 BASES APPLICABLE SAFETY AN ALYS ES (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). FSAR safety analyses for SGTR have been performed assuming that offsite 0

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 significant.

PORV s~etpoint is/reduced for LTOP in MODES 4, 5, ar/d 6 with the reactor vessel hep*d in place. LCO 3.4.12 addresses tl~e PORV 0 requirements in jhese MODES. l ACTIONS A.1 and A.2 m With the PORV inoperable, the PORV must be restored or the flow path isolated within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . The block valve sudbe closed and power must 0

be removed from the block valve to reduce the potentialf inadvertent POR op ning and depre surization .

0 d.pssurization 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. 1, Page 232 of 418

Attachment 1, Volume 9, Rev. 1, Page 233 of 418 Pressurizer PORV B3.4.11 BASES ACTIONS (continued).

B-1 and B,2 If the block valve is inoperable,.-it must be restored to OPERABLE status within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . The prime importance for the capability to close the block Svalve 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 If the 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. M. Block valve (

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.

SAny combination of indications (e.g., acoustic, system response)

SR 3.4j11.2 may be used to confirm a complete cycle of the PORV PORV cycling demonstrates its function. The Frequency of Mmonths is based on a typical refueling cycle and industry accepted practice.

BWOG STS B 3.4.11-4 Rev. 3.1, 12/01/05 Attachment 1, Volume 9, Rev. 1, Page 233 of 418

Attachment 1, Volume 9, Rev. 1, Page 234 of 418 Pressurizer PORV

%B3..411.

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 demonstra s that emergency power can be ,rovided and is 0

performed by transf rring power from the normal sup ly to the emergency supply and cycling t e valves. The Frequency of 18 onths is based on a typical refueling c cle and industry accepted practic REFERENCES 1. NUREG-0737, Paragraphl G.11, November 1980.

0

2. NRC IE Bulletin 79-05B, April 21, 1979.

[*J-4. ASME Code for Operation and Maintenance of Nuclear Power 0 SPlants.

ý3. Septemýber1985.

ýBAW-189ý0, 0 BWOG STS B 3.4.11-5 Rev. 3.1, 12/01/05 Attachment 1, Volume 9, Rev. 1, Page 234 of 418

Attachment 1, Volume 9, Rev. 1, Page 235 of 418 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. 1, Page 235 of 418

Attachment 1, Volume 9, Rev. 1, Page 236 of 418 Specific No Significant Hazards Considerations (NSHCs)

Attachment 1, Volume 9, Rev. 1, Page 236 of 418

Attachment 1, Volume 9, Rev. 1, Page 237 of 418 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. 1, Page 237 of 418

Attachment 1, Volume 9, Rev. 1, Page 238 of 418

1i ATTACHMENT 12 ITS 3.4.12, LOW TEMPERATURE OVERPRESSURE PROTECTION (LTOP)

Attachment 1, Volume 9, Rev. 1, Page 238 of 418

, Volume 9, Rev. 1, Page 239 of 418 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) , Volume 9, Rev. 1, Page 239 of 418

Attachment 1, Volume 9, Rev. 1, Page 240 of 418 ITS 3.4.12 ITS REACTOR COOLANT ,SYSTEM SAFETY VALVES - SHUTDOON LT1~TTTU rflOrTTTn* *r~ na*A'rn*J LCO 3.4.12 374.2 Decay Beat Removal\System relief valve Dl- 4shall be OPERABLE with a lift setting of < 330 psig and isolation valves H-iI d DH.12 open and coantrol power to their valve operators removed. L APPLICABILITY: MODES 4 and 5.

ACTION: Add proposed MODE 6 App0icability A. Vith not OPERABLE:

LAO 1 ACTION C 1. Make the valve OPERABLE within eight hours; or

2. a.. Vithin next one hour, disable the capability of both high pressur injection (HPI) pumps to inject water into the reactor coolant system; and

b. Vithin next eight hours: M02 ACTION D 1I- 1. Disable the automatic transfer'of makeup pump suction to the borated water 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 within the acceptable region on Figures 3.4-2a (in MODE 4) and 3.4-2b (in MODE 5).

ACTION A B. With DH-li D9-H2 closed, open DH-2i 4 Dli-2 within one hour.

LA01 ACTION B C. With the control power not removed from 0D8-li d DH-i2,ýremove the pover to the valve Operators at the M tor Contro enters ithin one hour.

SURVEILLANCE REQUIREMENTS 4.4.2 Decay Beat Removal System relief valve Dli-* 849 shall be determined 4.44.2 Decay Beat Removal System relief Valve hl edtrie OPERABLE: LAO1 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 eithert SR 3.4.12.1 I. isolation valves ID-11 4(dD0-12 open with control power removed from their valve operators; or LAVd Required Action B.2 2. Valves IDO-21 d08 open.

I The t setting p essure shall correspond to bient conditons of theLA valve at nominal rati temperature and pre sure. "

DAVIS-BESSE, UNIT 1 3/A 4-3 Amendment No. 37,7*9,135 Page 1 of 3 Attachment 1, Volume 9, Rev. 1, Page 240 of 418

Attachment 1, Volume 9, Rev. 1, Page 241 of 418 ITS 3.4.12 ITS Figure 3.4.12-1 Figure 3.4-2a I Reactor Coolant System Pressure - Pressurizer Level

Limits for inopera'leiDecay Heat Removal System Relief':Valvein MODE 4 400 350 - - - -UNACCEPTABLE REGION

02 0.*. 300 0 MODE 4 I...

a I.. 250 0.

a

-4'

'I-0 200 - - 1' 'ftczr~ - - - ~z*

I.'

0

-4 0

0 150 I..

0 U

.4' -

100 ACCEPTABLE REGION - - - - - -

50 I~~~ __ I__I__i__

.NOTE: NOT 6RRECTED YOR INý RMEUIT E~RRO -1 4-I~ 4~ -I LAQi

-1 I I I I I I I I

I a-0 40 .80 120 160 200 240 Initial Pressurizer Level (Inches)

DAVIS-BESSE, UNIT I 3/4 4-4a Amendment No. 0*, 116 Page 2 of 3 Attachment 1, Volume 9, Rev. 1, Page 241 of 418

Attachment 1, Volume 9, Rev. 1, Page 242 of 418 ITS 3.4.12 ITS Figure 3.4.12-2 .Fi~ure 3.4-2b I Reactor. Coolant System Pressuree - Pressurze 'Level Liiits for inoperable Decay Heat Removal System Relief Value in MODE 5 MOl 400 LAO1 350 U,

0~

300

.1~

QJ.

I-0~ 250 I-UNACCEPTABLE REGION 200 a

=

0 0 150 C.

A-U U

100

- Ln-"L" -......

AMMEIABLE REGION 50

ý_] , I I .I I t. ..1 \

0 40 sU IU 160t 2uu Z.U Initial Pressurizer Level (Inches)

DAVIS-BESSE, UNIT 1 3/4 4-4b Amendment No. ýj, 116 Page 3 of 3 Attachment 1, Volume 9, Rev. 1, Page 242 of 418

Attachment 1, Volume 9, Rev. 1, Page 243 of 418 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.

M02 CTS 3.4.2 Actions B and C provide compensatory measures if DH-11 or DH-12 is not open and control power removed. DH-11 and DH-12 provide a relief path for the DHR System relief valve. However, the CTS provides no default action for non compliance with either CTS 3.4.2 Action B or C. Furthermore, since the unit is already in a shutdown condition, CTS LCO 3.0.3 would not require anything other than putting the unit in MODE 5 (if the unit were in MODE 4), and this would not compensate for the inoperable flow path. ITS 3.4.12 ACTIONS A and B provide similar compensatory measures as CTS 3.4.2 Actions B and C when DH-i 1 or DH-12 is not open and control power removed. In addition, if ITS 3.4.12 ACTION A or B is not met, ITS 3.4.12 ACTION D will require similar compensatory actions as is required by CTS 3.4.2 Action A.2. This changes the CTS by providing clear, specific Actions if CTS 3.4.2 Action B or C is not met.

The purpose of CTS 3.4.2 is to ensure that there is a sufficient low temperature protection during shutdown conditions. While noncompliance with CTS 3.4.2 Action B or C is not a realistic occurrence, for consistency with the format of the ITS, a default action should be provided if CTS 3.4.2 Action B or C is not met.

Since the relief path is required to be OPERABLE to support the DHR System relief valve, compensatory actions equivalent to CTS 3.4.2 Action A.2 is applied Davis-Besse Page 1 of 3 Attachment 1, Volume 9, Rev. 1, Page 243 of 418

Attachment 1, Volume 9, Rev. 1, Page 244 of 418 DISCUSSION OF CHANGES ITS 3.4.12, LOW TEMPERATURE OVERPRESSURE PROTECTION (LTOP) for the situation where a noncompliance occurred with either CTS 3.4.2 Action B or C. The proposed actions are acceptable since they provide additional compensatory measures if the relief path is not restored to OPERABLE status and are consistent with the actions when the DHR System relief valve has not been restored within the associated Completion Time. This change is designated as more restrictive because it adds additional compensatory measures to be taken if the relief path is not restored to OPERABE status.

RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES LA01 (Type 1 - 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 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.

Davis-Besse Page 2 of 3 Attachment 1, Volume 9, Rev. 1, Page 244 of 418

Attachment 1, Volume 9, Rev. 1, Page 245 of 418 DISCUSSION OF CHANGES ITS 3.4.12, LOW TEMPERATURE OVERPRESSURE PROTECTION (LTOP)

LESS RESTRICTIVE CHANGES None Davis-Besse Page 3 of 3 Attachment 1, Volume 9, Rev. 1, Page 245 of 418

Attachment 1, Volume 9, Rev. 1, Page 246 of 418 Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs)

Attachment 1, Volume 9, Rev. 1, Page 246 of 418

Attachment 1, Volume 9, Rev. 1, Page 247 of 418 CTS LTOPt 3.4.12 3.4 REACTOR COOLANT SYSTEM:(RCS) 3.4.12 Low Temperature Overpressure Protection (LTOP) t 3.4.2 LCO 3A: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' INET 1

.------ ...------ " NO-E°-....s -- -----.... --- -- --- .....----

1. [Two ma eup pumps] may be 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 maximu RCS pressure for thexisting ROS:temp rature allowed by the p essure.and temperat re limitcurves provi ed in the PTLR.

a. Pressu izer level < [220] inch sand an OPERABL power operated relief v lve (PORV) with a lift etp int of * [555] p ig or

b. The R S depressurized and n RCS vent of >_[. 5] square inch.

APPLICABILITY: MOD 4 When anvACS cold lecv 6mperature id'ý [28310F. I MODE 5,1 MODE 6 when the reactor vessel head is on.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. More than [on ] makeup A.1 Initi te action to verify only Immediately pump capable of [on ] makeup pump is, INSERT 2 injecting into e RCS.

R/S capable of injecting into th B; HPI activat d. B.1 Iitiate action to verify H I Immediately ieactivated. /

BVWOG STS 3.4.12-1 Rev. 3.0, 03131/04 Attachment 1, Volume 9, Rev. 1, Page 247 of 418

Attachment 1, Volume 9, Rev. 1, Page 248 of 418 3.4.12 CTS O* INSERT 1 S3.4.2 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.

o 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. 1, Page 248 of 418

Attachment 1, Volume 9, Rev. 1, Page 249 of 418 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. 1, Page 249 of 418

Attachment 1, Volume 9, Rev. 1, Page 250 of 418 CTS 0

LTOP tern 3.4.12 ACTIONS continuedL CONDITION REQi UIRED ACTION. COMPLETION TIME late affected CFT. 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months C..-A C. A CIFT i4lated CET not is, when lated when .CA1 I lateaffeded 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 RS pressure for existing t* mperature

.allowd in th '..PTLR.

D. Required A iornC.1*.not D.1 I crease RCS tempera ure 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months met withintt 4erequired )> 1751F. 1t Completion ime.

OR/

S2 )epressurize affected FT 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months I <[555] psig..

E. Pressurizer level IE estore pressurizer le. el to 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months ,

> [220] inc es. [220] inches.

F. Requiredition E.i not- F.A lose and maintain ci sed 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months

'met within he required hemakeup control va ve Completio Time. I and 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 GA not HA1 Reduce makeup tan level 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months met withi the re to < [70] inches.

Completi n Time.

.AND H.2 Deactivate low low keup. j12 hours tank.Ievel interlock t the borated water stora tank.

suction valves.

BWOG STS 3A4.12-2 Rev. 3.0, 03131104 Attachment 1, Volume 9, Rev. 1, Page 250 of 418

Attachment 1, Volume 9, Rev. 1, Page 251 of 418 CTS 0

LTOP t 3.4.12 ACTIONS(continued)

CONDITION REUREDA ACION' COMPLETION TIME ep,uiz Q* Tc.an Pressurizer leve I. 1. Dpr ssurizea 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

9.5] squa,re inch.

AND PORV inopera le.

OR LTOP Systo inoperable fo any reason other than Condition A hrough Condition H SURVEILLANCE REQUIREMENTS 4.4.2.b SR 3412. Verify IPOR* block ablve i o en? d`6 ou fr I RCS to DHR isolation valves open with control power removed. I BWOG STS 3.4.12-3 Rev. 3.0, 03/31104 Attachment 1, Volume 9, Rev. 1, Page 251 of 418

Attachment 1, Volume 9, Rev. 1, Page 252 of 418 CTS 0

S LTOP t

. 34.1,2 SURVEILLANCE REQUIREMEýTS (ontinued)

SURVEILLANCE FREQUENCY SR 3.4,12.6 erify required RCS vent 2 .75] square inch is 112 hours0.0013 days 0.0311 hours 1.851852e-4 weeks 4.2616e-5 months for pen. unlocked open vent valve(s)

AND 31 days for other vent path(s)

SR 3A41 Perform ANNEL FUN,TIC)NAL TEýf for PORV.I Within 12] hoaints[

4.4.2.a Verify DHR System relief valve lift setpoint 5 330 psig i] RCS tm tre accordance with the Inservice Testing (IST) Program. to 1 ays thereafte SR 3.4.12.8 Perform CHANNEL C;ABRATION for PORV/ 1181 months INSERT 3 INSRT4 BVVOG STS 3.4.12-4 Rev. 3.0,ý0313!'/04 Attachment 1, Volume 9, Rev. 1, Page 252 of 418

Attachment 1, Volume 9, Rev. 1, Page 253 of 418 3.4.12 CTS 0 INSERT 3 400 Figure 3.4-2a 350 in 300 4,

I-4, 250 0.

cc 4' 200 0

=

S 0

150 0

U 0

4, cc 100 50 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. 1, Page 253 of 418

Attachment 1, Volume 9, Rev. 1, Page 254 of 418 3.4.12 CTS 0 INSERT 4 400 Figure 3.4-2b

- -.-..-.-.- L--

350

- N I N N. -

300 ,.

N N N - -, ..-...- - -.... N I-250 C.

I- ,, -, , -. . .UNACCEPTABLE REGION

'I 200 0

=

0 150 0

I..

C 4-U 100 50 ACCEPTABLEB REGION ,,,,,,

0 40 80 120 160 200 24.0 Initial 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. 1, Page 254 of 418

Attachment 1, Volume 9, Rev. 1, Page 255 of 418 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 0 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 280°F 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. 1, Page 255 of 418

Attachment 1, Volume 9, Rev. 1, Page 256 of 418 Improved Standard Technical Specifications (ISTS) Bases Markup and Justification for Deviations (JFDs)

Attachment 1, Volume 9, Rev. 1, Page 256 of 418

Attachment 1, Volume 9, Rev. 1, Page 257 of 418 All changes are a unless otherwise noted ILow Temperatre Overpressure Pection LTOPI t (1)

B 3.4.12 B.3.4 REACTOR COOLANT SYSTEM (RCS)

B.3.4.12 Low Temperature Overpressure Protection (LTOP)S tr0 BASES BACKGROUND -------------REVIEWEgE'S NOTE ............ --------------------

For plants for ich the NRC has aplroved LTOP setpoi ts based on non-10 CFR 5), Appendix G, methodology, as allowed it NRC Generic Letter 88-11,.he following Bases m.jst be revised acoo/ringly.

Me LTOP t controls RCS pressure at low temperatures so the integrity of the reactor coolant pressure boundary (RCPB) is not compromised byviolating 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 LCQ 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 PIT limits.

This LCO provides RCS overpressure protection in the applicable MODES by ensuring an adequate pressure relief ca~pacity-nd ~aminmm through the ecay Heat Removal p oolanthe r n capaedlity vaTe (pr'ssure relief capacity tequires eieda /

(DHR) System relief valve.the power op d RV) lift setpointto b reuced and pressurizer *oolant level at or belw a maximum limit *r the RCS /

depressuriz/ed and with an RCS *'ent of suff icient size/to Ihandle the,/

limiting trapsient during LTOP. /-

The LTOP appr1 6ach to protecting th* vessel by limiting oolant addition capability allovs a maximum of [on ] makeup pump, a d requires deactivating 7Pl, and isolating the ore flood tanks (C-s).

B'AOG STS B 3.4.12-1 Rev. 3.0, 03/31104 Attachment 1, Volume 9, Rev. 1, Page 257 of 418

Attachment 1, Volume 9, Rev. 1, Page 258 of 418 All changes are (c-I:

unless otherwise noted J ILow Tempernatrre Overpressure P ection ýLTOI e B 3.4.12 Q

BASES BACKGROUNDb; (continued)

Should more th n [one] HPI pump inj on an HPI actuatin, the pressurizer leve and PORV, or anothe RCS vent cannot revent overpressurizin the RCS. Even with only one HPI pump OPERABLE, the vent can"not 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 ROS vent is the ov pressure protection d vice that acts as b ckup to the operator in ter inating an increasin pressure event.

With HPI dea tivated, the ability to rovide RCS coolant addition is restricted. To balance the possible eed for coolant ad tion, 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 flowwith t e OPERABLE makeup pu through the makeup control valve.

PORV Re uire ýnts 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 . Whuen the monito ed pressure meets or exce ds the setting, the P RV is signaled to o en. Maintaining the setpoint w thin 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 erse. 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 decrease 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 exc ed the capabilities the vent. Thus, the vent path must be capable of r ieving 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 0

Attachment 1, Volume 9, Rev. 1, Page 258 of 418

Attachment 1, Volume 9, Rev. 1, Page 259 of 418 B 3.4.12 O* INSERT I 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. 1, Page 259 of 418

Attachment 1, Volume 9, Rev. 1, Page 260 of 418 All changes are unless otherwise noted 9 ILow Temperatre Overpressure P6Bection' TOi:K] te B 3.,41!2 BASES BACKGROUND: (continued)* IFor the remaining portions of MODE 3, overpressure B*D (protection is provided by operating procedures.

For an RCS verst to meetthe flow ca acity, itýrequires rer oving a pressurizer saf~tyvalve, locking the RV in the open pbsition and disabling'its blkck valve in the opentosition,o.or similarl establishingI aý vent by.openipg an RCS vent valve/ The vent path(s) st'be abovethe level of reacto r coolant, so asnot t* drain the RCS fn: open.i APPLICABLE Safety analyses (Ref. 3) demonstratethat the reactor vessel can be portions of SAFETY adequately protected against overpressurization transients durngn t MODE ANALYSES E shutdown. In MODES m 2, and'3/and iXMODE 4with RG,*-temperature lexce~ding [283]°F7 the pressurizer safety valves wil. .pDrevent RCS.4

ýpressure from exceeding the Referenje ilimits. At nominally 3 w* overpressure prevention falls to Wa OPERABLEP an a. 280

("DHR System relief valve. *--a res5sicient ri e coo antRCS size levelvent.

in theEach pressuriz r ormeans toa depressur' ed RCS and to--4O*

of thM~se has alltited l Below 140°F cre, dible ss r r ... *"

ov rpressure relief capability. I -

/overpressurization sources J S are secured. ,

The actual temperature at which the. pressure in the. P/T limit curve falls below the pressurizer safety valve setpointiincreases as vessel material toughness decreases due to neutron embrittlement. Each time the PtT limit curves are revised, the LTOP e will be re-;evaluated to ensure that its functional requirements can still be met with the P an DHR System relief valve and lpressurizer iev* method or the depress fized and vented RCS co dition.

operating procedures.

Transients that are capable of overpressurizing the RCS have been High Pressure Injection (HPI) identified and evaluated. These transients-relate to either mass input or -CoreFlooding' heat input: actuating thie System, discharging the T , energizing Tanks (CFTs) d the pressurizer heaters, failing the makeup control valve open, losing decay heat removal,+starting a reactor coolant pump (RCP)Jwt a arge tepeatrimsmatch betwee th tMary and secondary 9boo.lantZ systems, d adding nitrogen to th'pressurize .

HPI actuation a d CFT discharge are he transients that r uit in exceeding PIT I its within < 10 minut ls, in which time no perator action is assumed to t ke place. In the rest, operator action afte that time precludes over ressurization. The a alyses 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:

BOINSERT 2

BVVUG STS B 3.4.12-3 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 1, Page 260 of 418

Attachment 1, Volume 9, Rev. 1, Page 261 of 418 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.

0 Insert Page B 3.4.12-3 Attachment 1, Volume 9, Rev. 1, Page 261 of 418

Attachment 1, Volume 9, Rev. 1, Page 262 of 418 I

All changes are unless otherwise noted 9 ILow Temper re Overpressure P ection TOP E B 3.4.12 Q

BASES APPLICABLE SAFETY ANALYSES (continued)

a. Deactivatini all but [one] makeu pump,

b. Deactivat g HPI, and

c. Immobili ing CFT discharge islation valves in the losed positions.

DHR System relieThe Reference 3 analyses demonstrate thE-e[ can maintain RCS pressure below limits when only one makeupfiump is actuated.

Consequently, tlA LCO allows only [one] makeup pump to be OPERABLE in the LTOP MODE$.

Since the POR\ cannot do this for or HPI pump and th* RCS vent cannot do this fbr even one pump, th* LCO also requires/the HPI actuation circu s deactivated and th CFTs isolated./

The isolated 4FTs must have their 4ischarge valves clo ed and the valve power breake s fixed in their open lositions. The anales showwthe f F ofarIlc~4,,n, c,,~n n, rrA~rPf'.Q f-r~ Jýt ifr^ -~nnal IZ l ý"."'j valve

'The DHRisSystem placed relief in" / 175F and low) than 17Fanb*owthntaofteLC that oft 28]FndbIw.

/service before IRCS , ,

Itemperature is reduced I *Fracture me ch dics analyses established th temperature fLO belo 28 Applicabilit* [28i3 F. Above this temperature, the pressurizer safety andoperating pro( valves~provide the reactor vessel pressure protection. The vessel s I materials were assumed to have a neutron irradiation accumulation equal to 21 effective full power years (EFPYs) of operation.

This LCO will d activate the HPI act-/tion when the RC temperature is

< [283]°F. Th consequences of a s all break LOCA in/LTOP MODE 4 conform to 10 CFR 50.46 and 10 C-7R 50, Appendix K Refs. 4 and 5);,

requirements by having a maximu 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 nnce /

The fracture chanics analyses slbw that the vessel i protected when the PORV is t to open at _ [555] Isig. The setpoint i derived by modeling the erformance of the L'OP System, assu ing the limiting allowed LTO transient of uncontr tlied HPI actuation f one pump.

BWOG STS B 3.4.12-4 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 1, Page 262 of 418

Attachment 1, Volume 9, Rev. 1, Page 263 of 418 KAll changes are unless otherwise noted Low Temper re Overpressure P ection LTOIj t B 3.4.12 0

BASES APPLICABLE SAFETY ANALYSES (continued)

These: analyse consider pressure overshoot and underooat beyond ithe PORV openin and closing, resultinq from signal processing and valvý stroke times. /he PORV.setpoint ax or below the.deriv//d limit ensure.

the Referenc 1 limits will be met. / /

As required by License L

The PORVconflict PIT limits set~oint willthe with beLTOP re-evaluat d for ana sis compliance limits.l The P/Twheý limitsthe periodically modified as the reactor vessel material toughness decreases arerevised due to embrittlement induced by neutron irradiation. Revised PIT 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,4or LCO 3.4.3 91lscuss these ,aminations.I modifications, as necessary, to ensure continued means of protection for LTOP events The PORV is co sidered an active co ponent. Therefore, its failure will be provided to the NRC.

represents the orst case LTOP singl active failure.

Pressurizer Lev I Performance Analyses of opp rator response time s ow that the pressur er level must be maintained [2201 inches to provi e the 10 minute acti n time for correcting trans ents.

The pressurize level limit will also be re-evaluated for co pliance each time P/T limit c rves are revised bas d on the results of t e vessel material surveilance.

RCS Vent Perf rmance With the RCS epressurized, analys s show a vent of [0.. 51 square inches is capa le of mitigating the tr nsient resulting fro full openingof 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 ps g back pressure, w ich is less than the maximum RCS pressure on the T limit curve in LC 3.4.3.

The RCS yen size will also be re-e aluated for complia ce each time P/T limit curves a e revised based on th results of the vess I material surveillance.

The vent is p ssive and is not subj t to active failure The LTOP S stem satisfies Criteria 2 of 10 CFR 50.36 c)(2)(ii).

BVWOG STS B 3.4.12-5 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 1, Page 263 of 418

Attachment 1, Volume 9, Rev. 1, Page 264 of 418 All changes are unless otherwise noted 1Low Temper.re Overpressure PRection (LTOIqSt B 3.4.12 BASES LCO The LCO requir sanLTOP-System 0 ERBLE wtha Ii tedcoolant input capabil nda pressure relief pability. To limit co lantlinput, the LCO requires a aximum of [one] mal eup pump OPERA LE, the HPI deactivated, an the CFT discharge is lation valves close and immobilized. F r pressure reliefit reduires eitheruthe pre surizer coolant at~or below a ximum level and the RV OPERABLE.. th a lift setting INET 3ýý at the LTOP in t or the RCS depress rized and a vent es a blished.

The LCO is mo ified by two Notes. ote 1 allows [two m keup pumps] to be made capa le of injecting for 5 1 our during pump s p operations.

One hour provi es sufficient timetos felytcompletethea ualtransfer and to complet the administrative c ntrols and surveilla e requirements 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 OFT p essure is more than or equal t the maximum RCS ressure for the exis ng RCS temperature, allowed in LCO 3.4. . This Note permit the CFT discharge valv surveillance perfor d only under these pressure and temperature c 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 s [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 [0.751 square in' es.

APPLICABILITY S This LCO is applicable in MODE4hen y ROS cold leg tenperature is

~~r/i[Wr]F, in MgtIDE 51 and in MODE 6 when the reactor vessel head is d ::-*on.l The Applicability Itempera.*e .o[23° is established by fracture n 5 mechanics analyses. The pressurizer safety valves provide overpressure protection to meet LCO 3.4.3 P/T limits above [283l*9. VVth the vessel head off, overpressurization is not possible. in MODES 1,2, and 3 ]

LCO 3.4.3 provides the operational PIT 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 3, and MO. above [283]°FI.

BWOG STS B 34.12-6 Rev. 3.0, 03/31104 Attachment 1, Volume 9, Rev. 1, Page 264 of 418

Attachment 1, Volume 9, Rev. 1, Page 265 of 418 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-12) 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. 1, Page 265 of 418

Attachment 1, Volume 9, Rev. 1, Page 266 of 418 All changes are unless otherwise noted J Low Temprer re Overpressure Pection LTOPM] te B 3.4.12 BASES ACTIONS Al and BA 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 HPl. Emphasis is on mmediate deactivation be Iause inadvertent inje tion with [one] or m re HPI pump INSERT 4 OPERABLE is he event of greatest s gnificance, since it. uses the.

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 uickly proceeding wit the Required Action C.1 D.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 en the CFT pressure is at r more than the maxi um RCS pressure or the existing temperature a 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 t LTOP limits if both tanks ar fully injected. Depre surizing the CFTs b low the LTOP limit of [555] ig also prevents exc eding the LTOP limrs in the same event.

The Completi n Times are based o operating experien e that these activities can e accomplished in th se time periods an on engineering evaluations i dicating that a limiting LTOP event is not Ii ely in the allowed time E.1 F.1 and F.2 With the pro surizer level more tha [220] inches, the ti e for operator action in a pr ssure increasing eve t is reduced. The stulated event most affecte in the LTOP MODE 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, 03131/04 Attachment 1, Volume 9, Rev. 1, Page 266 of 418

Attachment 1, Volume 9, Rev. 1, Page 267 of 418 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 -<73inches 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. 1, Page 267 of 418

Attachment 1, Volume 9, Rev. 1, Page 268 of 418 All changes are 01 unless otherwise noted i ILow Ternper atre Overpressure P ection LTOI'P] t B 3.4.12 BASES ACTIONS (continued)

If restoration Required wit inF.II hour Actior and F2in either mustb cai e cannot bewithin performed acco 1 hours1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months lished,to close the makeup con rol valve and its isola ion valve. These equired Actions limitthe makeu capability which is trequired with a hi h pressurizer level, and permi cooldowh and depr surization to conti ue. Heatup must be stoppe because heat additi n decreases the re ctor coolant density aný5 inc easesthe pressurize level.

The Completio Times again are bas d on operating ex rience that these activities n be acromplished in these time perio s and on engineering ev luations indicating th It a limiting LTOP tr nsient is not likely in the all ed times.

G.1, H.1 and .2 With the R, inoperable, overpre ure relieving capa ility is lost, and restoration of t e PORV within 1.ho r is required. If that cannot be accomplished, the-ability ofthe Mak up System to add ter must be limited within t e next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

If restoration nnot be completedI 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 perfor ed to limit RCS wa er addition capability. M keup is notdeactivat d to maintain the R S coolant level.

Required Acti n H.land Required ction H.2 require r ucing the makeup tank vel to 70 inches and deactivating the Io low makeup tank level interlock to the borated water torage tank. This akes the available ma up water volume ins fflcient to exceed t e LTOP limit by a makeup cont Iavalve full opening.

These Compl tion Times also cons der these activities n be accomplishe in these time periods. A limiting LTQP e ent is not likely in those times.

Some PORV testing or maintenan e can only be perfo med at plant shutdown. S ch activity is permitt d if Required Actio HA and Required Action H.2 a taken to compensat for PORV unavail bility.

BWOG STS B 3.4.12-8 Rev. 3.0, 03/31/04 0

Attachment 1, Volume 9, Rev. 1, Page 268 of 418

Attachment 1, Volume 9, Rev. 1, Page 269 of 418 All changes are Z5 unless otherwise noted J SLow Temper;re Overpressureo Pection LTOiP]e 93.4.12 BASES ACTIONS (continued).

iA With the press izer level above (220 inches and the PO V inoperable or the LTOP S tern inoperable for a y reason other than cited in Condition A thr ugh H, Required Acti n 1.1 requires thed R, S depressurized nd vented within 122 urs from the timeo ither Condition started.

One or more v nts may be used. A ent.size ofŽ [0.75] quare inches is specified. Thi Vent size assumes 1 0 psig baCkpressur . Because makeup may required, the vent si eaccommodates i advertent full makeup syste operation. Such a v nt keeps the press re from full flow of [one] make p pump with a wide0o en makeup~control alve within the LCO limit.

The PORV ha a larger area and y be used for venti g 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 inad ertently actuated.

Therefore, ve ification of the deacti ation of two HPI pu ps, HPI injection, and the CFT must accompanythe depressurizing and .enting. Since these syste 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 experienc that this activity can be acco plished in this time p nod and on engine ring evaluations indicating th t a limiting LTOP tran ient is not likely imt is time.

SURVEILLANCE SR 3.4.12.1, S 3.4.12.2, and SR 3 4.12.3 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. TheseSurv illances ensure the minimum olant input capability will not create an R S overpressure condition to c 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 etty analysis.

BWMG STS B 3.4,12-9 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 1, Page 269 of 418

Attachment 1, Volume 9, Rev. 1, Page 270 of 418 All changes are 2 unless otherwise noted 1Low Temper.re Overpressure.ýection i 0 T:o te B 34.41.2:

BASES SURVEILLANCE REQUIREMENTS (continued)

SR 3.4.12:4 Verification of t pressurizer level at ý[220] inches by ob ervihg control roomor other i ications ensures a c shion of sufficient s ze is, available to reduce the r e, of pressure increas , from potential tran iients.

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 hen the ends of thes 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. f r 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. j SI 34.12S Verification that he PORV block valv is open ensures a f ow path to the PORV. This is equired at 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months in rvals.

INSERTS5 The interval h s been shown by ope ating practice suffic ent to regularly assess condit ons for potential degr* dation and verifyoa eration is within the safety an lysis.

SIR 3.4.12.6 The RCS vent f at least [0.75] squar. inches must be ve ified open for relief protection only if the vent is bei g used to satisfy th requirements of this LCO. F r a vent valve not loc ed open, the Frequ ncy is every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , Val s that are sealed or ecured in the open osition are considered 'lo ked' in this context. or other vent path( (e.g., a vent valve that is io ked, sealed, or secur d in position, a re ved pressurizer safety valve, r open manway), the quired Frequency i every 31 days.

Again, the Fr quency intervals con der operating practi e to determine adequacy to egularly assess condi ions for potential de radation and verify operati n within the safety a lysis.

BWOG STS B 3.4.12-10 Rev. 3.0, 03131/04 Attachment 1, Volume 9, Rev. 1, Page 270 of 418

Attachment 1, Volume 9, Rev. 1, Page 271 of 418 B 3.4.12 Q 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. 1, Page 271 of 418

Attachment 1, Volume 9, Rev. 1, Page 272 of 418 All changes are unless otherwise noted ILow Temperare Overpressure P ection TOP(JTte B 3.4.12 BASES SURVEILLANCE REQUIREMENTS (continued)

The passive vpnt path arrangementwust only be open tJ6be OPERABLE.

SR 3.4:12 A CHANNEL F NCTIONAL TEST is r quired within [12] h urs after decreasing RC temperature to < [28- ]°F and every 31 d s thereafter to STSTF changes443 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 INSERT 6 clarifies what i 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 hvapplicable extensions. RV actuation is not eeded, as it could di pressurize the RCS.

The [12] hour Frequency considers he unlikelihood of a ow temperature overpressure event during the time. The 31 day Freque cy is based on industry acce ted practice and is a eptable by experie ce with equipment re liability.

SR 3A4.12.8 The performan e of a CHANNEL CA IBRATION is requir d every

[181 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 considers a typical r fueling cycle and i dustry accepted practice.

REFERENCES 1. 10 CFR 50, Appendix G.

2. Generic Letter 88-11.

3.. FSAR, Section ,5 0

4. 10CFR 0.46. 0

5. 100 50, Appendix K.

B\AOG STS B 3.4.12-11 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 1, Page 272 of 418

Attachment 1, Volume 9, Rev. 1, Page 273 of 418 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 for proper 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.

0 Insert Page 3.4.12-11 Attachment 1, Volume 9, Rev. 1, Page 273 of 418

Attachment 1, Volume 9, Rev. 1, Page 274 of 418 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 0 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. 1, Page 274 of 418

Attachment 1, Volume 9, Rev. 1, Page 275 of 418 Specific No Significant Hazards Considerations (NSHCs)

Attachment 1, Volume 9, Rev. 1, Page 275 of 418

Attachment 1, Volume 9, Rev. 1, Page 276 of 418 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. 1, Page 276 of 418

, Volume 9, Rev. 1, Page 277 of 418 ATTACHMENT 13 ITS 3.4.13, RCS OPERATIONAL LEAKAGE , Volume 9, Rev. 1, Page 277 of 418

, Volume 9, Rev. 1, Page 278 of 418 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) , Volume 9, Rev. 1, Page 278 of 418

Attachment 1, Volume 9, Rev. 1, Page 279 of 418 ITS GTO ITS 3.4.13 REACTOR COOLANT SYSTEM OPERATIONAL LEAKAGE LIMITING CONDITION FOR OPERATION to:I 3.4.13 3.4.6.2 Reactor Coolant System operational leakage shall, be limited

a. No PRESSUREBOUNDARY 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 theReactor Coolant System,

e. 10 GPjkI CONTROLLED LE KAGE.and L01 f 5 GPM leakage from any Reactor Coolant System: Pressure Isolation Valve, as specified in Table 3.4-1 3.4.14 See ITS ']

9 APPLICABILITY: MODES 1, 2,3 amd 4 ACTION:

ACTION B a. With any PRESSURE BOUNDARY L.EAKAGE, or With primary to secondiry leakage not within limit, be in at least I-OT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months andin 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 Systern operational leakage greater than 'anly one of the above limits, excluding PRESSUREBOUNDARY 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 least HOT STANDBY withinlthe 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 except as perni ed by paragraph c below.

C. hi the event that integrity of any pressure isolation valve specified in Table 3.4 cannot be demonstrated, POWER OPEILATION may continue, provided tlhat at, least two valves in each high pressure line having a non-functional valve are in and remain in. the mode corresponding to the isolated condition.V) See ITS 3.4.14

d. The provisions of Section 3.0.4 are not applicable for entry into MODES 3 and 4 fobr the purpose of testing the isolation valves in Table 3.4-2.

[""Motor operated valves shall be placed in the closed position and power supplies deenergized. -- S4 ITS DAVIS-BESSE, UNIT 1 3/4 4-15 Ot.fdt -4L20841-Amnendment No.-t t-22,,2-*, 276 Page 1 of 2 Attachment 1, Volume 9, Rev. 1, Page 279 of 418

Attachment 1, Volume 9, Rev. 1, Page 280 of 418 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 theconta/nment 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 .

-- L0e

b. Monitoring the:con/ainment 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 .

c. Measurement of the NTROLLED LEAKAGE from thefeactor coolant pump seals Lol to the makeup syst when the Reactor Coolant System pyessure is 2185 +/- 20 psig at least once pr31 das.

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. ()(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 theprevious 9 months, and

c. Prior to returning the valye to service following maintenancerepair 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-d-teo-45O2g}-

Amendment No. 4 20, 276 Page 2 of 2 Attachment 1, Volume 9, Rev. 1, Page 280 of 418

Attachment 1, Volume 9, Rev. 1, Page 281 of 418 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. 1, Page 281 of 418

Attachment 1, Volume 9, Rev. 1, Page 282 of 418 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 (Category 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. 1, Page 282 of 418

Attachment 1, Volume 9, Rev. 1, Page 283 of 418 Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs) 0 Attachment 1, Volume 9, Rev. 1, Page 283 of 418

Attachment 1, Volume 9, Rev. 1, Page 284 of 418 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 RCSoperational 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.

BWOG STS 3.4.13-1 Rev. 3.1. 12/01/05 Attachment 1, Volume 9, Rev. 1, Page 284 of 418

Attachment 1, Volume 9, Rev. 1, Page 285 of 418 CTS RCS Operational LEAKAGE 3.4.13 SURVEILLANCEREQUIREMENTS _

SURVEILLANCE FREQUENCY 4.4.6.2.1.d SR 31413.1 ----- --'-ýNOTES- -------------

(including 1t Not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months footnotes after establishmentof steady stateoperati6n.

(1) and (2))

2. Not applicableto primaryto secondary LEAKAGE.

Verify RCS operational LEAKAGE is withinlimits 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 daythrough any one SG.

BVVG STS 3.4.13-2 Rev. 3.1, 12/01/05 Attachment 1, Volume 9, Rev. 1, Page 285 of 418

Attachment 1, Volume 9, Rev. 1, Page 286 of 418 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. 1, Page 286 of 418

Attachment 1, Volume 9, Rev. 1, Page 287 of 418 Improved Standard Technical Specifications (ISTS) Bases Markup and Justification for Deviations (JFDs)

Attachment 1, Volume 9, Rev. 1, Page 287 of 418

Attachment 1, Volume 9, Rev. 1, Page 288 of 418 RCS Operational LEAKAGE B 3.4.13 B 3.4 REACTOR-COOLANT SYSTEM (RCS)

B3.4.13 RCS Operational LEAKAGE BASES BACKGROUND Components that contain or transport the coolant to or fromthe 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 Systems.

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.

BWVOG STS B 3.4.13-1 Rev. 3.1, 12/01/05 Attachment 1, Volume 9, Rev. 1, Page 288 of 418

Attachment 1, Volume 9, Rev. 1, Page 289 of 418 RCS Operational LEAKAGE B 3A.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) is[ 1 gallon per minutel or increases.to D gallon per minute]as a result of 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 Li~~IJ containment resulting from a4steam line break CSLB) accident. To a lesser extent, other accidents or transients involve secondary steam 0

release tothe atmosphere, such as a steam generator tube rupture (SGTR). The leakage contaminates the secondary fluid.

Th**-SAR (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[MI gprn] primary to secondary LEAKAGE safety analysis assumption is relatively inconsequential.

00 00 The SLB more limitinqor site radiat n releases. The safety analysis 0 0

or e LB accident assumes the entire D gprrn primary to secondary LEAKAGE is through the affected generator as an initial condition. The dose consequences, resulting from the SLB accident are well within the limits defined in 10 CFR 100.

RCS operational LEAKAGE satisfiesCriterion 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.

BVWOG STS B 3.4.13-2 Rev. 3.1, 12/01/05 Attachment 1, Volume 9, Rev. 1, Page 289 of 418

Attachment 1, Volume 9, Rev. 1, Page 290 of 418 RCS Operational LEAKAGE B 3.1413 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 does not include pressure boundary LEAKAGE or controlled reactor Seucoolant pump (RCP) seale (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. 1, Page 290 of 418

Attachment 1, Volume 9, Rev. 1, Page 291 of 418 RCS Operational LEAKAGE B 3A.13 BASE$

APPLICABILITY (continued)

LCO 3.4.14, "RCS Pressure Isolation Valve (PM) Leakage,' measures leakage through each individual PIV and can impact this LCO. Ofthe 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 ad steady state operating conditions (stable temperature, power level, pressurizer nd makeup tank levels, m-m- aeupand letdown, [and RCP seall BWOG STS B 3.4.13-4 Rev. 3.1, 12/01/05 Attachment 1, Volume 9, Rev. 1, Page 291 of 418

Attachment 1, Volume 9, Rev. 1, Page 292 of 418 RCS Operational LEAKAGE B 3.4.13 WAE$:

SURVEILLANCE REQUIREMENTS (continued)

Q) 0 0

0 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. If this 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. 1, Page 292 of 418

Attachment 1, Volume 9, Rev. 1, Page 293 of 418 RCS Operational LEAKAGE B 3,.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 em raturpowerIleve'l pr ssurizer and makeu* tank levels, akeup land let wn, and RCP sea/injection and return fov.

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 inthe 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. 10CFR 50,,Appendix A, GDC 30.

2. Regulatory*Guide 1.45, May 1973.

FY %SAUR, 5Mq 00 4, NEI 97-06; "Steam Generator Program Guidelines."

5. EPRI, "Pressurized Vater Reactor Primary-to-Secondary Leak Guidelines.,"

BWOG STS B 3.4.13-6 Rev. 3.1, 12/01/05 Attachment 1, Volume 9, Rev. 1, Page 293 of 418

Attachment 1, Volume 9, Rev. 1, Page 294 of 418 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.

0 Davis-Besse Page 1 of 1 Attachment 1, Volume 9, Rev. 1, Page 294 of 418

Attachment 1, Volume 9, Rev. 1, Page 295 of 418 Specific No Significant Hazards Considerations (NSHCs)

Attachment 1, Volume 9, Rev. 1, Page 295 of 418

Attachment 1, Volume 9, Rev. 1, Page 296 of 418 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. 1, Page 296 of 418

Attachment 1, Volume 9, Rev. 1, Page 297 of 418 ATTACHMENT 14 ITS 3.4.14, RCS PRESSURE ISOLATION VALVE (PIV) LEAKAGE Attachment 1, Volume 9, Rev. 1, Page 297 of 418

, Volume 9, Rev. 1, Page 298 of 418 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) , Volume 9, Rev. 1, Page 298 of 418

Attachment 1, Volume 9, Rev. 1, Page 299 of 418 ITS 3.4.14 ITS RMACTOR CONDT.ANT SFOR'OERAM OPERATIONAL. LEAKAGE LIMITINGO CON DITION FOR OPERATION

.3.4-6,2 Reactor Coolant 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 ITS 3.413 13

d. 10 GPM IDENTIFIED LEAKAGE from the Reactor Coolant Systemn;

e. 10 GPM CONTIROLI.,)I., LEAKAGE, and

e. IGPNICONI'RT.,L],'I)IEAKGF,.jidj partLCO 13414 . GPM eakage from any Reactor Coolant.System.Piessure Isolation Valve. as SIR 3.4.14.2L0 APPLICABILITY: MODES 1, 2, 3 and -

ACTION: Add proposed ACTION Note 2

a. With aim PRE"SSURE B3OUNDARY LE.-.AKAGE', or with) primary toscnaySee ITS leakage not within limmi be niat least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in 34.413 COLD SHUTDOWN withiin the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

b. Wli-ih any Reactor Coolant System operational leakage greater than any one of the ACTION A - above limnits, excluding PRESSURE BOUNDARY LEAKAGE or prinmary to L02 secondary leakage, ,reduce the leakage raie to within limits within4-ours, rE inat ACTION B A I Beast 'S IAND13Y within th~e next 6hourm and inCOLD SHUTDOWN within

[the following30 hours except as pennitted by paragraph c below,

. in ihe event that integrity of any pressure isolation valve specified in Table 3.4-2 cannot be demonstrated, POWER OPERATION may continue, provided that at Add ACTION A proposed least two valves in each high pressure line having a non-functional valve are-in and Required remain in, the mode coiresponding to the isolaied.condition.!') Actions A. 1 Note SIR 3.4.14.2 Note

d. The provisions of Section 3.0.4 are not applicable for entry into. MOD1ES 3 and.4 for the purpose of testing the isolation valves in Table 3.4-2.

7and A.2 M01 ACTION A or operated valves shall be placed in the closed position and power supplies deenergized.

DAVIS-BESSE. UNIT I 3/4,4-15 Amendment No. f3:5,7, "t-87,)2-5( 276 Page 1 of 9 Attachment 1, Volume 9, Rev. 1, Page 299 of 418

Attachment 1, Volume 9, Rev. 1, Page 300 of 418 ITS 3.4.14 ITS REACTOR COOLANT SYSTEM SURVEILLANCE REQU IREMENTS 4A.462*2 Reactor Coolant System operational leakages shall be demonstrated tobe within each ofte 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.psigat least once per31 days.

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)

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)

SR 3.4.14.2 4.4.6.2.2 Each .Reactor Coolant System Pressure Isolation Valvelspecified iyt'able 3.4-2 shall be individually demonstrated OPERABLE by verifying leakage testing.(or the equivalent) to be within itsflimit prior to entering MODE 2:

a. After each refueling outage,

b. Whenever the plant hasbeen in COLD SHUTDOWN, for 7 days, or more, and if.

leakage testing has not been performed in the previous 9 months, and

c. 1Prior to returning thev ve to service following maintenance, pair or. replacementw Lh03 work on the valve, L0I 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 int grity of a pressure isolation valve listed'in Ta c 3.4-2 cannot be demonstrated, determine ard record the integrity of the high pressure flow ath on a daily basis.

Integri.y shall be determinld by performing either a leakage test of the re aining pressure isolation. valve, or a comb'ed leakage test of the remaining pressure isol tion valve in a series L0G with the closed motor-op ated containment isolation valve. In addition, record the position of the closed motor-operate containment isolation valve located in the hig, pressure piping on a daily basis.

See ITS

/II)Not applicable to primary to secondary leakage.

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.

3.4.13 DAVIS-BESSE, UNIT I 3/4 4-16 rder- d3O -

Amendment No. , 276 Page 2 of 9 Attachment 1, Volume 9, Rev. 1, Page 300 of 418

Attachment 1, Volume 9, Rev. 1, Page 301 of 418 ITS 3.4.14 ITS TABLE 3.4-2

,R .ACoR'COOLANTr SYSTEM P'RESSURE ISOLATI'ON VALVES

2. Decay H at Removal CF-310 5.0 gpm

3. Decaw eat Removal IH1T7 <5.0 gpm

4. Da veatt Removal 0H.-77 S.O gpm SR 3.4.14.2 Notes:

SR 3.4.14.2 k) ,l. Leakage rates less than or equal to 1.0 gpm are considered acceptable.

2. Leakage rates greater than 1.0 gpm but tess than or equal to 5.0 gpm are considered acceptable if the latest measured rate has not exceeded the rate determined by the previous 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.

3. Leakage rates greater than 1.0 gpm out less than or equal to 5.0 g~m a~r considered unacceptable if the latest measured rate ex=cded 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 9pm by 50% or greater.

4. Leakage rates greater than 5.0 gpm are considered unacceptable.

(b) IValves CF. 30 ad CF-31 will be tested-with the I ctor Coolant s~yste press re 34200 psig. ,Valves M1-76 and 011 7 will be tested with aorl I r Flooding Tank pressure which IPs 575 psig. M ni-a=, iffer lal test pressure across each valve hall not be less than 150 cps d.

()To satisfy RA requiremets, leakage may be M asured indirectly (as from. performance of pressure indicators if accomplished iaco ce with approved procedures and sup rted by computations showing Ith t he methodd is capable of demonst ing valve compliance with the I akale criteria.

DAVIS-BES*E, UNIT 1 3/4 4-16a Order dtd. 4/20/81 Page 3 of 9 Attachment 1, Volume 9, Rev. 1, Page 301 of 418

Attachment 1, Volume , Rev. 1, Page 302 of 418 I

ITS 601G I0.4.14 TABLE3.3-3 (Continued)

SAFETY FEATiURFS ACTUiATION SYSTEM INSTRUMENTATION

,MINIMUM TOTAL NO. UNITS UNITS APPLICABLE FUNCTIONAL UNIT OF UNITS. TO TRIP OPERABLE MODES, .ACTION

3. MANUAL ACTUATION a..SFAS (except Containment See ITS 3.3.6 Spray and Emergency Sump I Recirculation) 2 2 2 1,2;3,4 12

b. Containment Spray: 2: 2 2 1,2,3,4: 12

4. SEQUENCE LOGICtCHANNELS

a. Sequencer .4 2/BUS 2/BUS 1,2i3,4 154#

b. Essential Bus Feeder See ITS 3.3.8 Breaker Trip and ITS 3.8.1 J Degraded Voltage Relay (DVR) .***,* 2/BUS 2/BUS 1,23,4: 15#

c. Diesel Generator Start, Load Shied on Essential Bus Loss of Voltage Relay (LVR) 4. 2/BUS '2/B.US 1,2,3"4 15#

5. INTERLOCK CHANNELS A04 LCO 3.4.14 I a. Decay Heat Isolation Valve I I 1,2.3 13 A05 part 2 See ITS

b. Pressurizer Healers 2 2 2 3*** 14 3.3.5 Page 4 of 9 Attachment 1, Volume 9, Rev. 1, Page 302 of 418

Attachment 1, Volume 9, Rev. 1, Page 303 of 418 ITS 3.4.14 0

TAELE 3.3-3 t(Con&re TABLE NOAflQj

Tip fricon may bebypssod inthsmoDE withRcs pra~nwebow 180OP4 0 be automat Bypass sal MMo&d he UCS r*sMr xcees 1800 p*s,. See ITS 1 3.3.5 IWOfunction may be bypsed inR*w this MODE with RCS P*= "

.b60*i. yp be automaically removed when RCS presureexceed 660 p4g DELET SDELETE Allfuntinaltat$may be bypAssd form to one0 PUPor Ckwcultn~trwawum.'

016Mnute when sarftin echc Red. C4)kew 3 S.~ j.

0*00atSee *'V M dha MY ITS]

ACTION10- With the nunlbcebf OPBRADL fucional units ame ks th an h TotalNumbwr SeefT Unlts6 STAUUPsazidrPOWEiR OPMA1 pW,0fd~pwkle4 wiffifn M0e iour(exceet asoteit Iw1 tlinq-% A_&t See3.5

..- . C-.A ~ A. -.

codiimWh'w u %a nit sploacd kuaon mpemble "atu9ole for. I perfomanmn efta C IA*4 LFIJNcno '1 MM .Sa doolwadw~o of kqxesaffiy and assocated eatry into th& AcflNmdaxeaamnybo delayed fiwtip to '8 hoiwxprvdod at lugtWo* cnwoodarftfctdwal unis aoptreOPA LE ACT'om i - Wit anyoozponet in fte Outpu Loe& Wnpemble, ftip fte associatedecomponetst See ITS withr ow, e bcmw or

be In 8tl as HTSTANDBY widthh next 6 ho=r md InCOLM 3.3.7 ]

SHU DOWN within he foowin 30 ho.

DAVIS-BESS=, UNIT I 3/43-12 Amendment No. W8~742,102, 135, 159,18621 ,21=1,2~259 Corected by lette date 6W2~8 9

Page 5 of 9 Attachment 1, Volume 9, Rev. 1, Page 303 of 418

Attachment 1, Volume 9, Rev. 1, Page 304 of 418 ITS 3.4.14 ITS TABLE 3.3-3 (Continued)

ACTION STATEMENTS ACTION 12 - With the number of OPERABLE Units one less than the TotalNumberm of Units, restore the inoperable functional unitltoOPERABLE 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 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 .

See ITS 3.3.6 I ACTION C ACTION 13 - a. With less than the Minimum Units OPERABLE and indicated reactor coolant pressure > 328 slg. bothDecay Heat Isolation Valves

. J.(DH]Id DHI2) shall1 be verified closed.

b. With Less than the Minimum Units OPERABLE and indicated reactor coolant pressure < 328 psig operation may cont'inue; however, the functional unit shall be OPERABLE.prior to increasing indicated reactor coolant pressure above 3.28 psig.

I.i ACTION 14 -. With lessthan the MinimumUnits 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, or the inoperable functional unit shall be placed in the tripped state.

SeeITS 3.3.5 I ACTION 15 - a. With the number of OPERABLE units one less thanwtheMinimum 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 I."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. Z8,52,1D2,1,5,211, 2 Page 6 of 9 Attachment 1, Volume 9, Rev. 1, Page 304 of 418

Attachment 1, Volume , Rev. 1, Page 305 of 418 I I0.4.14 ITS TABLE 3.3-4 SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION ALLOWABLE VALUES FUNCTIONAL UNIT ALLOWABLE VALUES0#

.z FNSTRUMENT ST1RINGS,

a. DELETED DELETED

b. Containment Pressure - High < 19.38 psia

c. Containment Pressure - High-High

d. RCS Pressure - Low

< 41.65 psia See ITS 3.3.5

]

> 1576.2 psig e, RCS Pressure - Low-Low Ž-,441,42 psjg

f. BWST Level >.101.6 and

- 11 5.4 in, 11,0 SEQUENCE LOGIC CHANNELS

a. Essential Bus Feeder Breaker Trip, >_3712"volts (dropout) and

< 3771 volts (pickup)

Degraded Voltage Relay (D.VR) Witlt a time delay 61 See ITS t

5 6A arid

7.9 see 3.3.8J

b. Diesel Generator Start, Load Shed on Essential Bus >-,2071 volts (dropouLH and Loss of Voltage Relay (LVR) _ 2492 volts (pickup) with a time delay of

> 0.42 and

0.58 sec INTERLOCK C1IANNELS.

SR 3.4.14.3, a. Decay Heat Isolation Valve eIT < 328 psig

SR 3.4.14.4 rid AaPiesurizer.eater CIN .e TES LA03 I* Referenced to th~e RC-ýrssure insrmnainip*'I(

1" All,,wable Values fo0r CHA NNEL. FUNC:'F7O.NAI;'! nS See ITS 3.3.5 ]

Page 7 of 9 Attachment 1, Volume 9, Rev. 1, Page 305 of 418

ITS

0 Attachment 1, Volume 9, Rev. 1, Page 306 of 418 10.4.14 TABLE; 4,3-2 (C.on__tinued)

SAFETY FEATURES ACTUATION SYSTEM INSTRUMENPTA3ION SURVEILLANCE REQUIREMENTS CHANNEL. MODES n,:WHICH CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST REQUIIRE,

4. SEQUENCE LOGIC CHANNELS a- Sequencer S NA M 1, 2i 3j 4

b. Essential Bus Feeder Breaker Trip, S A(3) M(3) I.2, 3, 4 See ITS 3.3.8 Degraded Voltage Relay. (DVR) and ITS 3,8.1 J

c. Diesel Generator Start, Load Shed on: S A(3) M(3) 1,2,3, 4 Essential Bus, Loss of Voltagc Relay (LVR)

SR 3.4.14.3, SR 3.4.14.1 SR 3.4.14.5 SR 3.4.14.4

5. INTERLOCK CHANNELS

a. Decay Heat Isolation Valve S R I I 'I I t,*

11-Prpetrv,er t-Ta~t.*r Pressurizer Heater S TABLE NOTATION p.

R 4* 3 # See ITS 3.3.5 I

(1) Manual actuation switches shall be tested at least once per REFUELING INTERVAL. All other circuitry associated with manual safeguards.acttition shall receive a CHANNELFUNCTIONAL TEST at least once per 31. days.

See ITS 3.3.6 I

(2) The CHANNEL FUNCTIONAL TEST shall include exercising the transmitter by applying either vacuum orpressureto the appropriate See ITS side of the transmitter. 3,3.5 (3) The as-let instrument setting shall be returned to a setting within the tolerance band of the trip setpoint established to protect the safety limit.

[tsee 1.

ITs3 See Specification 4.5.2.d.)

/ .#H When either Decay Heat Isolation Valve is open. 1M33IS ..

See ITS]

" 4SR3.4.14.3, SSR3.4.14.4 Page 8 of 9 Attachment 1, Volume 9, Rev. 1, Page 306 of 418

Attachment 1, Volume 9, Rev. 1, Page 307 of 418 ITS 3.4.14 ITS Revised by UM Letter Dated

$!RMELILARCIF RFnUTRFUMFNL firnntlneiMi June 6, 1995

b. At least once each REFUELING INTERVAL, or prior to operation after ECCS piping has been drained by verifying that the EMCS piping is full of water by venting the EZS pW ,casings and discharge piping high points.

C. By a visual inspection which verifies that no loose debris. (rags.

trash1 clothing, etc.) is present in the containment widch could be transported to the containment emergency strmp and cause restriction of the pump suction during LOCA conditions. This See ITS 315 .2 visual inspection shall be performed:

1. For all accessible areas of the containment prior to establishing CONTAINHERT INTEGRITY, and

2. For all areas of containment affected by.an entry, at least once daily while work is ongoing and again during thefinal exit after completion of work (containment closeout) when CONTAINMENT INTEGRITY is established.

LCO 3.4.14 part 2

d. At least once each REFUELING INTERVAL by:

I1. Verifying that the interl, I

a)

SR 3.4.14.4 SR 3.4.14.3 Note, no required if -the valve Is'closed -and 480 V AC power

,is disconnected from its-motor operators.

SR 3.4.14.4 Note b) Prevent the opening of IDH-11 add Mmiaa<

SR 3.4.14.3 simulated or actual reactor coolant system pressure which Is greater: than the Allowable Value (,i= psig) I is appIed Add proposed SIR 34.1_4.a5 - nM I

2. a) A visual Inspection of the containment "emergency. sUP which verifies'that the subsystem section "nlets are not restricted by debris and that the sum c ents.

(trash racks ereens. 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 1 for the MVT outlet isolation valves and the 3.5.2 ]

containment emergency suM recirculation valves energized, the B Outlet Valve HV-DH7A (HV-D4l7B) automatically close in $75 seconds after the operator manually .pushes the control -switch to open the Containment Emergency Sut* Valve HV-DHgA (HIV*.O9B) which should be verified to open 'in V7S seconds.

-3. Deleted DAVIS-BESSE, UNIT I 3/4 5-4 Amendment No. 8.254)8 A0 77 l5 Page 9 of 9 Attachment 1, Volume 9, Rev. 1, Page 307 of 418

Attachment 1, Volume 9, Rev. 1, Page 308 of 418 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. 1, Page 308 of 418

Attachment 1, Volume 9, Rev. 1, Page 309 of 418 DISCUSSION OF CHANGES ITS 3.4.14, RCSK PRESSURE ISOLATION VALVE (PIV) LEAKAGE two decay heat isolation valves (DH-1 1 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.

A05 CTS Table 3.3-3, Functional Unit 5.a (Decay Heat Isolation Valve) includes a Note # that applies to Action 13 and states the provisions of Specification 3.0.4 are not applicable. ITS 3.4.14 does not include this Note. This changes the CTS by deleting the specific exception to Specification 3.0.4.

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

CTS 3.0.4 has been revised as discussed in the Discussion of Changes for ITS Section 3.0. ITS 3.0.4, in part, states that when an LCO is not met, entry into a MODE or other specified condition in the Applicability shall only be made when the associated ACTIONS to be entered permit continued operation in the MODE or other specified condition in the Applicability for an unlimited period of time.

ITS 3.4.14 ACTION C requires the plant to isolate the affected line and allows operation to continue for an unlimited period of time. Therefore, because the ITS still allows the plant to change MODE or other specified condition in the Applicability, this change is considered to be consistent with the current allowance provided by the CTS Note. This change is designated as administrative because it does not result in a technical change to the CTS.

MORE RESTRICTIVE CHANGES M01 CTS 3.4.6.2 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 Davis-Besse Page 2 of 7 Attachment 1, Volume 9, Rev. 1, Page 309 of 418

Attachment 1, Volume 9, Rev. 1, Page 310 of 418 DISCUSSION OF CHANGES ITS 3.4.14, RCS PRESSURE ISOLATION VALVE (PIV) LEAKAGE 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.

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 1 - 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 Davis-Besse Page 3 of 7 Attachment 1, Volume 9, Rev. 1, Page 310 of 418

Attachment 1, Volume 9, Rev. 1, Page 311 of 418 DISCUSSION OF CHANGES ITS 3.4.14, RCS PRESSURE ISOLATION VALVE (PIV) LEAKAGE 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.

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 Davis-Besse Page 4 of 7 Attachment 1, Volume 9, Rev. 1, Page 311 of 418

Attachment 1, Volume 9, Rev. 1, Page 312 of 418 DISCUSSION OF CHANGES ITS 3.4.14, RCS PRESSURE ISOLATION VALVE (PIV) LEAKAGE 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 2 - 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 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 Davis-Besse Page 5 of 7 Attachment 1, Volume 9, Rev. 1, Page 312 of 418

Attachment 1, Volume 9, Rev. 1, Page 313 of 418 DISCUSSION OF CHANGES ITS 3.4.14, RCS PRESSURE ISOLATION VALVE (PIV) LEAKAGE 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 (Category 5- 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 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).

Davis-Besse Page 6 of 7 Attachment 1, Volume 9, Rev. 1, Page 313 of 418

Attachment 1, Volume 9, Rev. 1, Page 314 of 418 DISCUSSION OF CHANGES ITS 3.4.14, RCS PRESSURE ISOLATION VALVE (PIV) LEAKAGE 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 if the DHR valve interlock is inoperable. This change is acceptable because the 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. 1, Page 314 of 418

Attachment 1, Volume 9, Rev. 1, Page 315 of 418 Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs) 0 Attachment 1, Volume 9, Rev. 1, Page 315 of 418

Attachment 1, Volume 9, Rev. 1, Page 316 of 418 CTS ROS PIV Leakage 34.14 3.4 REACTOR COOLANT SYSTEM (RCS) 34.14, RCS Pressure -Isolation Valve (PlV)"Leakage 3.4.6.2.f LCO 3A.14 Leakage from each RCS PIV shali be within limits.

0 APPLICABILITY: MODES 1, 2, and:3,_

MODE 4, except valves in the e a.t-r-emoval DHIJ flow path when 00 in, or during the transition to or from, the DHR mode of operatio

[and the DHR System interlock function 0

ACTIONS

- .... .... ... .... .... ... NOT ES ..--- .. .. ....

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 A2 within limit, must have been verified to meet SR.3.4.1.4. and be onlthe RCS pressure boundary[,r the high 0 pressure portion of the systenrf 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 BWVG STS .3.4.14-1 Rev. 3.0, 03131104 Attachment 1, Volume 9, Rev. 1, Page 316 of 418

Attachment 1, Volume 9, Rev. 1, Page 317 of'418 3.4.14 CTS INSERT 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. 1, Page 317 of 418

Attachment 1, Volume 9, Rev. 1, Page 318 of 418 CTS RCS PIV Leakage 3.4.14 ACTIONS (ontnued CONDITION REQUIRED ACTION COMPLETION TIME

.A.2 ]lsolate 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 I second closed manual, deactivated automatic, or check valve.

[or]

Restore 0D I

limits.

Action b 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 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 C. (( Decay Heat Removal IC 1 Isolate the affected *t 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months Table 3.3-3 Action 13 (DHR) System _-pene ation by use of[e 3G uocl I ureý interlock L!*2 closed man or function: inoperable, valve.*----

deactivated automatic [i 0

BVWOG STS 3.4.14-2 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 1, Page 318 of 418

Attachment 1, Volume 9, Rev. 1, Page 319 of 418 3.4.14 CTS

( INSERT 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. 1, Page 319 of 418

Attachment 1, Volume 9, Rev. 1, Page 320 of 418 CTS RCS PIV Leakage 3.4.14' SURVEILLANGEREQUIREMENTS SURVEILLANCE FREQUENCY Action c footnote (a),

4.4.6.2.2.d SR 3 .4 .. 14 .f-' ----------.-- NOTE*---------

required to be performed:in MODIE INSERT 3 0 0

2. Not requirej to be performed on Ihe RCS PlVs located in tpe DHR flow path wh)*n in the DHR mode of o eration. W/

0 3,

3, RCS PIVs a/tuated uring t e peiwormance of this SurveilHance are. not require to be tested more than knce if a repetitivete ting loop 0

cannot be Tvoided. I r %Tý6eakage from each 4.4.6.2.2.a, RCS PIV is equivalent to In accordnce 4.4.6.2.2.b, * ~~~pmj0TnominaI inch fvlesz ~oa withthe nservice Table 3.4-2 mat an RCS pressure Testing rrogramr psi ,andr5[2, psia.i or [81months. G©Q2

--- of 2155 psig and]

AND INSERT 4 Prior to entering MODE 2 0

whenever the unit has been in MODE 5for 7 days or more, if leakage testing has not been performed in the previous.9 months 0

AND

[ Within hours following valve actuatio due to automat c or manual ction or flow thr ugh the valve]

BWOG STS 3.4.14-3 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 1,.Page 320 of 418

Attachment 1, Volume 9, Rev. 1, Page 321 of 418 3ý.4.14 CTS 0 INSERT 3 Table 4.3-2 Functional 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 Unit 5.a System interlock channel common to Safety Features Actuation System (SFAS) instrumentation.

(* INSERT 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. 1, Page 321 of 418

Attachment 1, Volume 9, Rev. 1, Page 322 of 418 CTS RCS PIV Leakag'e.

3.4.14 SURVEILLANCE REQUIREMENTS (continued)'d SURVEILLANCE FREQUENCY Table 4.3-2 SR 3.4. -NOTE Functional [ Not required to be met when the DHR System Unit 5.a, qtaclurel interlockos disabled in accordance with 4.5.2.d.1 .b) LCO 3.4.12. function Verify DHR System autocsure interlock revents [11+monthslJ 0

the valves from being opened with a simulated or actual RCS pressure signal z 0@

Table 4.3-2 SR 3.4.4 ----...-------..... NOTE ---------

Functional ((Not required to be met when the DHR System Unit 5.a, laqtcu interlocks disabled in accordance with 4.5.2.d.1 .a) LCO 3.4.12.

G0 Verify DHR System autocsuinterloccauses the I ] monthsr valves to close automatically with a simulated or 3 actual RCS pressure signal .

BWOG STS 3.4.14-4 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 1, Page 322 of 418

Attachment 1, Volume 9, Rev. 1, Page 323 of 418 3.4.14 CTS (D 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. 1, Page 323 of 418

Attachment 1, Volume 9, Rev. 1, Page 324 of 418 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 MODE 4 Applicability exception has been modified to also exclude the DHR System interlock function, consistent with the Davis-Besse Current Technical Specifications (CTS).

The 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.

Davis-Besse Page 1 of 2 Attachment 1, Volume 9, Rev. 1, Page 324 of 418

Attachment 1, Volume 9, Rev. 1, Page 325 of 418 JUSTIFICATION FOR DEVIATIONS ITS 3.4.14, RCS PRESSURE ISOLATION VALVE (PIV) LEAKAGE

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 2 of 2 Attachment 1, Volume 9, Rev. 1, Page 325 of 418

Attachment 1, Volume 9, Rev. 1, Page 326 of 418 Improved Standard Technical Specifications (ISTS) Bases Markup and Justification for Deviations (JFDs)

Attachment 1, Volume 9, Rev. 1, Page 326 of 418

Attachment 1, Volume 9, Rev. 1, Page 327 of 418 RCS PIV Leakage B[3.4.14 B,~,4 REACTOR COOLNT SYSTEM (ROS)

B3.4.14 RCSPressure Isolation Valve (PIV) Leakage BASES BACKGROUND 10 CFR 50.2. 10 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 any twolnormally closed valves in pressure boundary series within the [ýlressure boundary that separate the high pressure reactor valves, which are RC.S from an attached low pressure system. During their lives, these valves can produce varying amounts of reactor coolant leakage through oolant o 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 PlVs in a line must be included as part of the identified LEAKAGE, governed by LCO 3.4.13, "RCS Operational LEAKAGE." This istrue 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 aline is not RCS operational LEAKAGE ifthe 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 Safet* 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 follown picalyn Iconnete ssems:

SDecay Heat Removal (DHR) System.-]

BWOGSTS B 3.4.14-1 Rev. 3.1, 12/01/05 Attachment 1, Volume 9, Rev. 1, Page 327 of 418

Attachment 1, Volume 9, Rev. 1, Page 328 of 418 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. 1, Page 328 of 418

Attachment 1, Volume 9, Rev. 1, Page 329 of 418 RCS PIV Leakage, B 34.14 BASES BACKGROUND (continued)

Ia. Decay Heat Reval (DHR) System]

1b. Emergency Core C g System-(ECCS), and 1c. Makeup and ification System. CF-30, CF-31, DH-76, andD -7

ý The PIVs are Ilisted in [FSAR sion] Reference . o Violation of this LCO could result in continued degradation of a PIV, which could lead to overpressuiization of a low pressure system and the loss of the integrity of a fission product bar. NE A APPLICABLE Reference 4 identified potential intersystem LOCAs asa *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. Becauset ow 1pressure portion o the DH R System is[ designed or sig, n to handle normal RCS pressures, overpressurization failure of the DH R low pressure line would result in a LOCA outside containment and subsequent risk'of core melt.

Reference 5 evaluat various PIV-configurations, leaka e testing of the valves, and operati al changes to determine~the effe on the probability of intersystem LO As. This study concluded that perodic leakage testing of the PIVs can ubstantially reduce the probability an intersystem 0

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. significarntly~suggests that something is operationally wrong and corrective action must be taken.

The LCO PIV leakage limit isJ0.5 gpm-per nominal inch of/valve size with a maximum limi o0 gpm. e previous criterion of 1 g for all valve sizes imposed n unjustified penalty on the larger valve without INET2providing infor ation on potential valve degradation an resulted in higher person el radiation exposures. A study conclud d a leakage rate limit based on valve size was superior to a single allo ble value.

BWVOG STS B 3.4.14-2 Rev. 3.1, 12/01/05 Attachment 1, Volume 9, Rev. 1, Page 329 of 418

Attachment 1, Volume 9, Rev. 1, Page 330 of 418 B 3.4.14 0 INSERT 1A 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.

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. 1, Page 330 of 418

Attachment 1, Volume 9, Rev. 1, Page 331 of 418 RCS PIV Leakage B 3414 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 leakageochannel opening, In such cases, the observed rate maybe adjusted to the:rnaximum pressuredifferential:

by-assuming leakage; is directly proportional to the pressure differential to INSERT 3 th one half power. 0 APPLICABILITY In MODES 1, 2, 3, and 4, this LCO applies. because the PIV leakage the potential is greatest when the RCS is pressurized. In MODE 4,fva yes in. 0 and the DHR System interlock function is notl*

the DHR flow path are not'required to meet the requirements of this. LCO when in, or during "

the transition to or.from, the DH R mode.ofoperatiol. 0 required ome heurmentsIn of this LCOJ MODES 5 and 6, leakage limits are not provided because the lower reactor coolant.pressureý results in a reduced potential for leakageand for a LOCA outside the containment.

ACTIONS The ACTIONS are modified by two Notes. Note 1 is added'toprovide 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 PIViS 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 If the leakage from one or morenoRCS PIvs wihi t limit, theis J MTe flow path must . . be isolated

. e.. by two valves. . Required.Actions

. . A.t 0

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 pressureboundary[Pr the high pressure portion of the syster.i 0

Required Action Al requires that the isolation with one valve ýmust 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. 1, Page 331 of 418

Attachment 1, Volume 9, Rev. 1, Page 332 of 418 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. 1, Page 332 of 418

Attachment 1, Volume 9, Rev. 1, Page 333 of 418 RCSý PIV Leakage B 3.4.14, BASES ACTIONS (continued)

J Required Action A.2 specifies that the doubler isolation barrier of two Q valves be restored by closing some other-valve qualified for isolation or restoring one leaking PIV... [*The ,72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months time after exceeding the limit 0 considers the time required to complete-the Action andthe low probability of a second valve failingduring this time period.

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 PIVto OP ABLE status. This timeframe co siders the time 0

required to comple e this Action and the low probabili y of a second valve failing during this eriod. ]

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

Two options are pro ided for Required Action A.2. Th second option (72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months restoratio is appropriate if isolation ofa se ond valve would place the unit in an unanalyzed 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 Q

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.1 -fad2 5yson The inoperability of the DHR autoc)sure interloc enders 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 as the DHR suctio-ýene ion is by at lea t one closed ma ua l 5 il trd eactivate au oma ic va v wthin 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .T This action accomplishes the purpose of the auto sur function. '

interlock BVWOG STS B 3.4.14-4 Rev. 3.1, 12/01/05 Attachment 1, Volume 9, Rev. 1, Page 333 of 418

Attachment 1, Volume 9, Rev. 1, Page 334 of 418 B 3.4.14 O* 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. 1, Page 334 of 418

Attachment 1, Volume 9, Rev. 1, Page 335 of 418 RCS PIV Leakage B 3.4.14 BASES SUR*VEILLANCE REC .UIREMENTS:

SR 3.4*14.E C 2 0

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 leakaje is INER 6

{belovlthe specified limit and to identify each leaking valve. W~he. leakage, li-ýmit.*f 0.5 gpm p 'rinch of nominal valve imtruot p aximumJ 0 t ....

[INSERT 7 Japplies:to eachvvIve.FIag tesin requires a *stable. pressure condition.f 0

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 [1 ] months, a typical refueling cCye if 24 the plant does not go into MODE 5 for at least 7 days. The (( nttj]

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 apply during an outage and the potential for an unplanned transient if the Surveillance were performed with the plant at power.

[In addition, testing Pust be performed once after the falve has been opened by flow or Ixercised to ensure tight reseatinr. PtVs disturbed in the performance ot this Surveillance should also be ýsted unless documentation sh4ws that an infinite testing loop ca not practically be avoided. Testing fnust be performed within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after the valve has 0

been reseated. /ithin24 hours is a reasonable an/ practical time limit for performing thi test after opening or reseating a valve. ]

( performed I The leakage limit is to be"*at the RCS pressure associated with 0

MODES .1 and 2. This permits leakage testing at high differential pressures with stable conditions not possible in the MODES with lower pressures. t L 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 BVVOG STS B 3.4.14-5 Rev. 3.1, 12/01/05 Attachment 1, Volume 9, Rev. 1, Page 335 of 418

Attachment 1, Volume 9, Rev. 1, Page 336 of 418 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.

OINSERT 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. 1, Page 336 of 418

Attachment 1, Volume 9, Rev. 1, Page 337 of 418 RCS PIV Leakage B 314.14 BASES SURVEILLANCE REQUIREMENTS (continued) in MODE 5 for 7 days or more, if leakage testing has not been performed in the previous 9 monthsI In addition, thisSurveillan e is not required to be-performedon thel DHR System When the DHR Sy1tem is aligned to the RCS in the de, y heat-removal mode of operati n. PIVs contained ir the DHR flow path ýust be leakage: rate.tested afteý DHR is. secured and 0

stable unit conditiohs and the necessary differential /pressures are established.

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

The "24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ..." Fre uencyof performance for Surveill nce Requirement 3,4.14. is not, required for B&W Owner' Group plants licensed prior to 198 . These plants were licensed pri r to the NRC establishing fornal. echnical Specification controls fo pressure isolation valves. Subsequen ly, these earlier plants had their Ii enses 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 time, the content of those Orderswas considered acc ptable. Since 0 1980, the.NRC Sta has determined anadditional 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 orde to ensure the PIVs ability to support th integrity of the reactor coolant pressure 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 ..."

Frequencyto refle current NRC Staff position on th need to include this test requirement thin Technical Specifications.

MSR 3.4.142and SR 3.,4.14.r 430 psig, the pressure at which this section of DHR piping was tested 00 00 Verifying thatthe DHR auto osure interlocks are OPERABLE ensures at the RCS pressure that RCS pressure will not pressurize the DH R system beyond 1 instrumentation tap lits design presstre of [6001 psi The interlock setpoint that prevents the allows DH-11 and DH-12 to be opened by the operator prior to 328 valves fr m being opened is set so.the actual RCS pressure must be psig to open the valves. This setpoint ensures th gDHR desig Iressure will noy be exceeded and the R relief valve'will not lifit. The 00 suction pressure is lost to the reactor coolant pumps

ý24t- month Frequency. is based on the need to perform this Surveillance under the conditions that apply during a plant outage and the potential for 0

24 an unplanned transient if the Surveillance was performed with the reactor at power.0--3month Th Frequency is also acceptable based on 0 consideration of the design reliability (and confirming operating experience) of the equipment.

BVWOG STS B 3.4.14-6 Rev. 3.1, 12/01/05 Attachment 1, Volume 9, Rev. 1, Page 337 of 418

Attachment 1, Volume 9, Rev. 1, Page 338 of 418 RCS: PIV Leakage' B 3.4.14 BASES SURVEILLANCE REQUIREMENTS (continued) System intelock These SRs are modified by Notes allowing the DHR autodlosurej function

'to be disabled when using the DHR System suction relief valvefor cold 0

.overpressure protection in accotdancewith LCO 3.4.12.M] ---

REFERENCES 1. 10CFR5O.2.

2. 10CFR55a(c).

3. 10 CFR 50, Appendix A, Section V, GDC 55,

.4, NUREG-751014, Appendix V, October 1975.

16. [Document mning list of PIVs.] 0

7. ASME Code for Operation and Maintenance of Nuclear Power Plant%-. , 1995 Edition with 1996 Addenda. 0

8. 10 CFR 50.55a(g).

0 Letter from D.G. Eisenhut, NRC, to all LWR Licenses, LWR Primary Coolant System Pressure Isolation Valves, February 23, 1980.

6. Letter from J.F. Stoltz, NRC, to R.P. Crouse, Order for Modification of License Conceming 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. 1, Page 338 of 418

Attachment 1, Volume 9, Rev. 1, Page 339 of 418 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. 1, Page 339 of 418

Attachment 1, Volume 9, Rev. 1, Page 340 of 418 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. 1, Page 340 of 418

Attachment 1, Volume 9, Rev. 1, Page 341 of 418 Specific No Significant Hazards Considerations (NSHCs)

Attachment 1, Volume 9, Rev. 1, Page 341 of 418

Attachment 1, Volume 9, Rev. 1, Page 342 of 418 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. 1, Page 342 of 418

Attachment 1, Volume 9, Rev. 1, Page 343 of 418 0 ATTACHMENT 15 ITS 3.4.15, RCS LEAKAGE DETECTION INSTRUMENTATION Attachment 1, Volume 9, Rev. 1, Page 343 of 418

, Volume 9, Rev. 1, Page 344 of 418 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) ., Volume 9, Rev. 1, Page 344 of 418

Attachment 1, Volume 9, Rev. 1, Page 345 of 418 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 stimoleyel anlow onitoring system, and LA01

b. One0containment atmosphere radioactivity monitor (gaseous or particulate).

APPLICABILITY:: MODES 1, 2, 3 and 4.

ACIN: Add proposed Required Action A. 1 Note L0 ACTION A a. *With the required containment sump le an" oW_[monitoring system inoperable, operation may continue up to 30 days provided Surveillance Requirement 4.4.6.2.1 .d is perforned 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 re'quired 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 ýB.1.2 Note * *L01 ACTION C c. With the above required ACTION and associated completion time not met, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWNwithin 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:

SIR 3.4.15A1, 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 Page 1 of 5 Attachlment 1, Volume 9, Rev. 1, Page 345 of 418

Attachment 1, Volume 9, Rev. 1, Page 346 of 418 ITS 3.4.15 ITS REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued)

SR 3.4.15.4 b. Containment sump.leyl an[fl wmonitoring system-performance of CHANNEL CALI BRAT least once each REFUELING INTERVAL.

SR 3.4.15.1, c. Containment atmosphere gaseous monitoring system-per formance of SR 3.4.15.2, CHANNEL CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST at the frequencies specifiedlin"Tabl.e 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. 1, Page 346 of 418

Attachment 1, Volume 9, Rev. 1, Page 347 of 418 0

ITS 3.4.15 ITS TABLE 3.3-6 e

RADIATION MONITORING INSTRUMENTATION C,,

MINIMUM CHANNELS APPLICABLE ALARM/TRIP MEASUREMENT STI INSTRUMENT OPERABLE MODES SETPOINT RANGE ACTION i 1. AREA MONITORS See ITS

a. Fuel Storage Pool Area 3.3.14 ]

Emergency Ventilation System Actuation 1** < 2 x background 0..- 107mr/hr .22

2. PROCESS MONITORS

a. Containment LCO 3.4.15.b i. Gaseous Activity RCS Leakage Detection 1" 1, 2, 3, & 4 Not Applicable 21 B LCO 3.4.15.b ii. Particulate Activity RCS Leakage 1"

Detection 1, 2, 3, & 4 Not Applicable 21 B I

As recuired by Specification 3.4.6.1.

__**Wit__fel__n____storage___________di__ tSee ITS

-N I

- With fuel in the storage pool or building, 3.3 .14 J z9 Page 3 of 5 Attachment 1, Volume 9, Rev. 1, Page 347 of 418

Attachment 1, Volume 9, Rev. 1, Page 348 of 418 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 the Minimum 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 ACTION requirement- of Specification 3.9.42. 3.3.14 DAVIS-BESSE, UNIT I 3/4 3-33 Amendment No. 135 Page 4 of 5 Attachment 1, Volume 9, Rev. 1, Page 348 of 418

Attachment 1, Volume 9, Rev. 1, Page 349 of 418 IOS3.4.15 ITS AOl TABLE,4.3-3

~n. RADIATION MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS SR 3.4.15.1 SR 3.4.15.3 SR 3.4.15.2 C CHANNEL MODES IN WHICH z CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE INSTRUMENT CHECK CALIBRATION TEST REQUIRED I, AREA MONITORS See ITS a: Fuel Storage Pool Area 3.3.14 .

Emergency Ventilation System.

Actuation S E M **

2. PROCESS MONITORS

a. Containment

i. Gaseous Activity RCS Leakage Detection* S-1 E-3 M -2 1,2j3&4 ii, Particulate Activity RCS Leakage Detection' S -1 E -3 M -2 1, 2,3 &4 LCO 3.4.15

ff required by Specification 3.4.6.1 to be OPERABLE.

-**Withfuel in the storage pool or building F { 3.3.14 See ITS z0 Page 5 of 5 Attachment 1, Volume 9, Rev. 1, Page 349 of 418

Attachment 1, Volume 9, Rev. 1, Page 350 of 418 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. 1, Page 350 of 418

Attachment 1, Volume 9, Rev. 1, Page 351 of 418 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 (Category4- 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. 1, Page 351 of 418

Attachment 1, Volume 9, Rev. 1, Page 352 of 418 Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs)

Attachment 1, Volume 9, Rev. 1, Page 352 of 418

Attachment 1, Volume 9, Rev. 1, Page 353 of 418 RCS: Leakage Detection Instrumentation

. CTS 3,4.15 34;: REACTOR COOLANT SYSTEM (RCS) 3.4.15 oRCS Leakage Detection Instrumentation 3.4.6.1, LCO 3.415 The follwivng 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. Onewcontainment 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 . 7---- .- NT---- ...---..

sump:monitor Not required until.12.hours 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. 1, Page 353 of 418

Attachment 1, Volume 9, Rev. 1, Page 354 of 418

. CTS RCSILeakage, Detection Instrurnentation 3.4.15 A CTIO NS : (continued) . . . .. ............. ... .... .. ... .. .

CONDITION REQUIRED ACTION COMPLETION TIME I.

B.1U2--, ....... ;-NOTE".-ý- .......

Not, required until 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months sI aftr .establishment of steadystate operation.

PerformrSR 3.4113.1-. 1Once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months AND B;2 Restore required .30 days 6onitain ment atmosphere radioactivity monitor tI 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 B C.2 Be in MODE 5. :36 hours Q)

Action d D. Both required monitors D0.1 Enter 1LCQ03.O3. 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 i -- days 4.4.6.1.c containment atmosphere radioactivity'monitor.

BVVOG STS 3.4.1i572 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 1, Page 354 of 418

Attachment 1, Volume 9, Rev. 1, Page 355 of 418 RCS.Leakage Detection Instrumentation

.CTS 3.415 SUR*VEILLANCE REQUIREMENTS:(continued) .. . .. .. .

SURVEILLANCE .FREQUENCY 4.4.6.1.b SR 3.415g]k Perform CHANNEL CALIBRAT ION of required containment sumprmonitor.

M11. mtronths 0 4.4.6.1.a, SR Perform CHANNEL CALIBRAT ION of required i 'f*months 4.4.6.1.c, Table 4.3-3 containment atmosphere radioa ctivity monitor.

Instruments 2.a.i and 2.a.ii BWOG STS 3.4.15-3ý Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 1, Page 355 of 418

Attachment 1, Volume 9, Rev. 1, Page 356 of 418 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.

0 Davis-Besse Page 1 of 1 Attachment 1, Volume 9, Rev. 1, Page 356 of 418

Attachment 1, Volume 9, Rev. 1, Page 357 of 418 Improved Standard Technical Specifications (ISTS) Bases Markup and Justification for Deviations (JFDs)

Attachment 1, Volume 9, Rev. 1, Page 357 of 418

Attachment 1, Volume 9, Rev. 1, Page 358 of 418 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 Ato 10 CFR 50 (Ref. 1) requires means for Although not detectin and,tothe extent practical, identifying the location of the source committed to of RCS LEAKAGE. Regulatory Guide 1-45 (Ref. 2) describes acceptable methods forselecting leakage detection systems. i, Q

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 FaiiVow..j.containment sump used to collect unidentified LEAKAGE is instrumented to alar fo increases of 0.5 to 1.0 gpm in the normal flow rates. This etectg ------ 0 sensitivity is acceptable for detecting increases in unidentified LEAKAGE.

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- pCi/cc radioactivity for particulate monitoring and of 10s pCilcc 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 hu -ity of the containmer atmosphere would indicate release of 'wateruydpor tothe containme 4'. Dew point temp ature measurements/,n thus be used to rnitor humidity leve of the containrent*mosphere as an ndi tor of potential CR LEAKAGE. A Q

1F increaa in dew point is well thin the sensitivity nge of available instrume s. .e "

0 BW2OG STS B 3.4.15-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 1, Page 358 of 418

Attachment 1, Volume 9, Rev. 1, Page 359 of 418 RCSLeakage Detection Instrumentation B 3.4A15 BASES BACKGROUND: (continued)

Since the :humidity vet is influenced by $everal factors, a qupntitative evaluation of an i dicated leakage rate this means may b questionable a should be coiplare

observedincrea s in liquid flow into or fro the containment su p [and condensate ow from air coolers]. H idity level monitorin is~considered most seful as an indirect al m or indication to ale .the operator to a p tential problem.

Q Humidity nitors are not requ'edforthis LCO.

Air temperature and pressure monitoring methods may also be used to infer unidentified LEAKAGE to thecontainment. 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 leakageto the containment.

Temperature and pressure monitorS are not irequired 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 verify with indications ANALYSES

.from other Systerns is necessary. Te :systemviresponse/timesan Isensitivities are deq ribed in the FSAR (Ref. 3). Multiple instrument 0

locations are utilized, if needed, to ensure the transport delay time of the

.leakage from its source to anýinstrument locationyields 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.

(Refer to the Bases of LCO 3.4.13, "RCS Operational LEAKAGE,"

Ifor further information regarding RCS LEAKAGE. I 0 BWOG STS B 3.4.1572 Rev. 3.0, 03131/04 Attachment 1, Volume 9, Rev. 1, Page 359 of 418

Attachment 1, Volume 9, Rev. 1, Page 360 of 418 RCS.Leakage Detection Instrumentation 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, mon.itoring principles to be OPERABLE.to provide a high degree of oonfidence that extremely small,leaksiare detected in time to allow actions to 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,. "-

0*

provides an acceptable minimum. lel(both thef 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 isS 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, noother form of sampling can provide the equivalent information..

However, the containment atmosphere ct ity 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 3.413.1 is:not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months afterestablishing stea state operation stable ED temperature, power levelq pressurizerland makeup tank levelImkj Iand letdown, and [RCP qeainT-ction and return flowsl). 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. containment' Restoration of the required sump monitor to OPERABLE status is required to regain the function in a.Comnpletion 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.

BVV3G STS B 3.4.15-3 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 1, Page 360 of 418

Attachment 1, Volume 9, Rev. 1, Page 361 of 418 RCS Leakage Detection Instrumentation B 3.4.15 BASESl ACTIONS (continued) 1.3.t1 BA1,2,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 inforration. 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 informationthat 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 stead state operation.

(stable temperature, power level, pressurizer Fnd ma~ up tank level((

1make nd letdowni and R 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 processall necessary data after stable plant conditions are established. The 30,day Completion Time recognizes at least one other form, of leak detection :is

,available.

CA and C.2 if a' Required Action of Condition.A or B cannot be metwithin 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, 03/31104 0

Attachment 1, Volume 9, Rev. 1, Page 361 of 418

Attachment 1, Volume 9, Rev. 1, Page 362 of 418 RCS Leakage Detection Instrumentation B314.15 BA-SES SURVEILLANCE SR 3..15.

REQUIREMENTS SR3.4.115.1 requires the performance of a CHANNEL CHECK of the required contafinnent atmosphere radioactivity monitor. The check gives reasonable"confidence that each 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 tES fcathebrequired containment aemosphere radioactivity monitor. Aits

,successful test:'of/tle required. on C~~f~hannoel relaýYmay be erormed:b th (Veiiaon 0fth*hag of sfctate a sgecnat 1f tf Sthe rely 3 T.413arifie3.at4 T EST - f a..rel y, 'This%

acceptable CHANNE.FUNCTIONAL s pccepw le because all of the fither required 0

contacts 0f' erelay are verfi:ed byýother Technical Sj~ecifications and non-Te~chr,(al SpecificatinStests at least once per yefUeling interval with considers ~f~t reibiiy Intmn andoertng exprinc ha shown it proper~~~~ac

-:Appicabi* exten'siorns. dgaain IThe test ensures that the, monitor can perform itsoofeetn

,function: in the desired nmanner. The,test verifies the alarm setpoint and

.relativeaccuracy . f the instrument string. 'The Frequency of

ýco .Siders inistru-menlt reliabilityi, and oeprating experience has shown it s 0

ro~per~for detecting: degradation.

SR :31AJ5.3 andSR: 3.4.15.4 These SRs requiire the.performance of a*CHANNEL CALIBRATIONfor (or24 months, each ofthe required RCS leakage detection inStrumentation channel. ( as applicable, The calibration verifies-the accuracy of the. instrument string, including the instrument:slocated inside containment. The Frequency:ofT!A nmonth

~typicalret u fg cyc-. and considers channell eliabilityYEE-[9d Q

.operating experience has proven this Frequency is acceptable.

REFERENCES' 1. 10 CFR 50, Appendix A, Section IV,GDC 30.

2. Regulatory.Guide 1.45.

13. FSARe'ction [ 1.1 0 BWVG STS B 3.4.15-5 Rev. 3.0. 03/31104 Attachment 1, Volume 9, Rev. 1, Page 362 of 418

Attachment 1, Volume 9, Rev. 1, Page 363 of 418 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. 1, Page 363 of 418

Attachment 1, Volume 9, Rev. 1, Page 364 of 418 Specific No Significant Hazards Considerations (NSHCs)

Attachment 1, Volume 9, Rev. 1, Page 364 of 418

Attachment 1, Volume 9, Rev. 1, Page 365 of 418 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.4.15, RCS LEAKAGE DETECTION INSTRUMENTATION There are no specific NSHC discussions for this Specification.

0 Davis-Besse Page 1 of 1 Attachment 1, Volume 9, Rev. 1, Page 365 of 418

, Volume 9, Rev. 1, Page 366 of 418 ATTACHMENT 16 ITS 3.4.16, RCS SPECIFIC ACTIVITY , Volume 9, Rev. 1, Page 366 of 418

Attachment 1, Volume 9, Rev. 1, Page 367 of 418 Current Technical Specification (CTS) Markup

and Discussion of Changes (DOCs) 0 Attachment 1, Volume 9, Rev. 1, Page 367 of 418

Attachment 1, Volume 9, Rev. 1, Page 368 of 418 ITS 3.4.16 ITS REACTOR COOLA.VT SYSTEM SPECIFIC ACTIVITY LIMITING CONDITION 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. 4 1.0 pCi/gram DOSE EQUIVAfLENT I1131, and SR 3.4.16.1 b. < 100/f IeCi/gram APPLICABILITY: MODES 1, 1, L-l ACTION:

MODES 1, 2 and 3*; Add proposed ACTION A Note ACTION A a. with the specific activity of the primary coolant >. 1.0 pCi/gram DOSE EQUIVALENT 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 " exceeding the limit line shown ACTION B -_- on gure 3-, e in at least HO0 STANDBY vith Tv< 530OF A

within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

ACTION B b. With the specific activity of. the. primary coolant > 200/E I pCi/gram, be in at least HOT STANDBY vithTavS < 530 0TF.ithin 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

MODES 1,, 2. ff3, ý.

ACTION A a. With the specific activity of the primary coolant > 1.0 L03 pCi/gratr DOSE. EQUIVALEN'T 1-131 jor Ci/graz, perform, the L0 sampling and analysis requirements of item 4 a) of Table 4.4-.4 S LC..vtbi-3-s pecific activity of th I o -i t A02

{o limits, jlFor reportiz:::ýrequiremeatsree o___ --- ,

SSect ion 6,9.1.-Sý.* Annual pperatinga Rep.ort. I * - A*

SR 3.4.16.1, 4.4.8 The specific 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 *Vith Tavg > 5300 F.

DAVIS-BESSE, UNIT I 3/4 4-20 Amendment No. 91.14 Page 1 of 3 Attachment 1, Volume 9, Rev. 1, Page 368 of 418

Attachment 1, Volume 9, Rev. 1, Page 369 of 418 0 1 B.4.16 ITS TABLE 4.4-4 PRIMARY COOLANT SPECIFIC ACTIVITY SAMPLE 0

ANU ANALYSIS PRUOHRAM tn TYPE OF MEASUREMENT SAMPLE AND MODES IN WHICH SAMPLE AND ANALYSIS ANALYSIS FREQUENCY 7days AND ANALYSIS REQUIRED SR 3.4.16.1 1. Gross Activity Determination At least once each 7 urs SR 3.4.16.2

2. (LA01-per 14 oays I SR 3.4.16.3

3. 1 per 6 months*

L01 Required I 4. a) Once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , whenever 1 A, S#

Action A.1 ,1 -the-specific activity exceeds 1.0 vCi/aram DOSE EQUIVALENT L03*

1-131 For-6TO i*--i/qram ,and SR 3.4.16.2 b) One sample between 2 and 6 1. 2, 3LOl hours following a THERMAL PuwER change exceeoing 15 per-cent of the RATED THERMAL POWER within a one hour period.

I"Until the specific activity " " lnt system is restored within its limits..I A02 SR 3.4.16.3 *Sample to be taken after a minimuRof2 EFPD and.20 days. ofRPOWER OPERATION have elapsed since the Note reactor was last subcritical for 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or longer.

Page 2 of 3 Attachment 1, Volume 9, Rev. 1, Page 369 of 418

Attachment 1, Volume 9, Rev. 1, Page 370 of 418 ITS 3.4.16 ITS Figure 3.4.16-1 U 0 40 go, so 70 so 90 100 PERMICT Ol. RATED THERMAL POWER FIGURE 3.4-1 DOSE EQUIVALENT 1-131 Primy Coolant Specific Activity Uimt Vmrsu Pco.. of RATED THERMAL POWER with the Pilmary Coolant Specific Act > 1.0,&CI/pam DOSU EQUIVALENT 1-131 I DAVIS-SEUs, uxYT i 3/4 4-23 .1nendnient :No. 135 Page 3 of 3 Attachment 1, Volume 9, Rev. 1, Page 370 of 418

Attachment 1, Volume 9, Rev. 1, Page 371 of 418 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 required 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 across-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. 1, Page 371 of 418

Attachment 1, Volume 9, Rev. 1, Page 372 of 418 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 (Category 2- 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 T.ag > 530 0 F.

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'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. 1, Page 372 of 418

Attachment 1, Volume 9, Rev. 1, Page 373 of 418 DISCUSSION OF CHANGES ITS 3.4.16, RCS SPECIFIC ACTIVITY require the unit to be shutdown to MODE 3 with RCS Tavg < 530'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 (Category 4 - 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 f.Ci/gm. The ITS does not contain this Action. This changes the CTS by eliminating a conditionally performed Surveillance when gross activity exceeds 100/f p.Ci/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. 1, Page 373 of 418

Attachment 1, Volume 9, Rev. 1, Page 374 of 418 DISCUSSION OF CHANGES ITS 3.4.16, RCS SPECIFIC ACTIVITY activity exceeds 100/E ýtCi/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 < 5300F 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/E kaCi/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 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 E 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 E when the isotopic concentrations in the core are stable. This change is acceptable because the new Surveillance Frequencyhas 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. 1, Page 374 of 418

Attachment 1, Volume 9, Rev. 1, Page 375 of 418 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. 1, Page 375 of 418

Attachment 1, Volume 9, Rev. 1, Page 376 of 418 Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs)

Attachment 1, Volume 9, Rev. 1, Page 376 of 418

Attachment 1, Volume 9, Rev. 1, Page 377 of 418 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.

5 a302 APPLICABILITY: MODES 1 and 2, MODE 3 with RCS average temperature (To).vg F ACTIONS CON DITION REQUIRED ACTION COMPLETION TIME Action a A. DOSE EQUIVALENT NOTE----------------. . --------------------.

(MODES 1, 2, and 3* 1-131 > 1.0 ItCilgm. 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 8.1 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 T;vg <WOF. 0 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 OR (MODES 1, 2, and 3*) 2 Gross specific activity of the reacto coolant not within limit.

BWOG STS 3.4.16-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 1, Page 377 of 418

Attachment 1, Volume 9, Rev. 1, Page 378 of 418 CTS RCS Specific Activity 3.4.16 AýCTIONScniud__

CONDITION, T- REEQUIRED ACTION COMPLETION TIME 2

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 1 100/1E pCi/gm.

LCO 3.4.8.a, SR 3.4.16.2 NO TE ------------------........

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 pCilgm.

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 > 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 3A4.16.3 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 Ž 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months .

Determine E. 184 days BWOG STS 3.416-2 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 1, Page 378 of 418

Attachment 1, Volume 9, Rev. 1, Page 379 of 418 CTS RCS Specific Activity 3.4.16 Figure 3.4-1 3

Figure 3.4.16-1 (page 1 of 1)

Reactor Coolant DOSE EQUIVALENT 1-131 Specific Activity Limit Versus Percent of RATED THERMAL POWER With Reactor Coolant Specific Activity >1.0 t.Ci/gm DOSE EQUIVALENT 1-131 BWOG STS 3.416-3 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 1, Page 379 of 418

Attachment 1, Volume 9, Rev. 1, Page 380 of 418 3.4.16 CTS INSERT 1 3

(275, 26) 275 Figure 3.4-1 250 kX 225 N I PRTO 200 E

" 175 E

S150 I-zW.J125

-- I 80

!60 -

LLI (100 ACCEPTABLE 00 OPERATION 75 (60,100) 50 25 0

20 30 40 50 60 70 80 90 100 RTP (%)

Insert Page 3.4.16-3 Attachment 1, Volume 9, Rev. 1, Page 380 of 418

Attachment 1, Volume 9, Rev. 1, Page 381 of 418 JUSTIFICATION FOR DEVIATIONS ITS 3.4.16, RCS SPECIFIC ACTIVITY

1. The0MODE 3 Applicability for this Specification has been changed from 500OF to 530 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 < 5000 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. 1, Page 381 of 418

Attachment 1, Volume 9, Rev. 1, Page 382 of 418 Improved Standard Technical Specifications (ISTS) Bases Markup and Justification for Deviations (JFDs)

Attachment 1, Volume 9, Rev. 1, Page 382 of 418

Attachment 1, Volume 9, Rev. 1, Page 383 of 418 RCS Specific Activity B1314.116 B3.4 REACTOR COOLANT SYSTEM (RCS)

B 134.16 RCS Specific Activity BASES BACKGROUND The Code of Federal Regulations, 10 CFR 100 (Ref. 1)i 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 atthe site boundary will not exceed a small ANALYSES fraction of the 10 CFR 100 dose guideline limits following an SGTR a accident. The SGTR safety analysis (Ref. 2) assumes Aespecific activity f the re t at the LCO limitsd and an existing reactor coolant steam generator (SG) tube leakage rate of 1 gpm. The analysis also valueaequivalent assumes a reactor trip and a turbine trip lat t mtime as th .TR following a The analysis for the SGTR accident establishes the acceptance limits for 3

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.

Lanalysis The assumed RCS specific activity in the SGTR bounds the LCO limit for RCS specific activity.

BWOG STS B 3.4.16-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 1, Page 383 of 418

Attachment 1, Volume 9, Rev. 1, Page 384 of 418 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 atmnosvalves or the 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 the 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 specificactivity 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 may result 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. *such 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. 1, Page 384 of 418

Attachment 1, Volume 9, Rev. 1, Page 385 of 418 RCS Specific Activity B:3.416 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 UK) andgross specific activity are necessary to contain the potential consequences of an SGTRto within the acceptable site boundary dose values.

For operation in MODE 3 with RCs average temperature <F, and in 3D 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 thelatmos valves andi main steam safety valves.

ACTIONS A.1 and A.2 With the DOSE EQUIVALENT 1-131 greater than the LCO limit, samples.

G2 at intervals of 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months must be taken to :st~r~a~th-e limits of lerify An isotopic analysis of a Figure 3.4.16-1 are not.exceeded.c The Completion Time of 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months is reactor coolant sample must 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.

.31 If a Required Action and associated Completion Time of Condition A are F n 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 or if the gross specific average temperature <5°"Fwithin 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . The Completion Time of activity is notrwithin limit, 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months is required to get to MODE 3 belows- F without challenging 0 6 reactor emergency systems.f BWOG STS B 3.4.16-3 Rev. 3.0, 03/31104 Attachment 1, Volume 9, Rev. 1, Page 385 of 418

Attachment 1, Volume 9, Rev. 1, Page 386 of 418 RCS Specific Activity B3.4.16 BSES ACTIONS (continued)

With the gross cific 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 i 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 s of the main steam safety valves, and prevents venting the SG to the env nment 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 E . 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 remains within limit during normal operation and following fast power verification that the reactor coolant changes when fuel failure is more apt to occur. The 14 day Frequency is DOSE EQUIVALENT 1-131 specific adequateactivity to trend changes inevery is monitored 7 days.

the iodine The Frequency, activity betwen level considering gross Sactivity is within limit. This / specific 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 I hour period, performing an isotopic analysis of is established because the iodine levels peak during this time following

ýareactor coolant sample. Ifuel 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. 1, Page 386 of 418

Attachment 1, Volume 9, Rev. 1, Page 387 of 418 RCS Specific Activity B3.4.16 BASES SURVEILLANCE REQUIREMENTS (continued)

SR 314,16.3 SR 3.4.163 requires radiochemical analysis for E.determination every 184 days t with the plant operating in MODE 1 equilibrium 2 conditions. The P determination directly relates to the LCO and is required to verify plant operation within the specific gross activity LCO limit. The analysis for L is a measurement of the average energies per disintegration for isotopes with half lives longer than 15 minutes, excluding iodines. The Frequency of 184 days recognizes E does not change rapidly. ta-tes is not required}

This SR has been modified by a Note that 6 samplingto be 2 L@tpeformed31 days after a minimum of 2 EFPD and 20 days of MODE 1 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 P is representative and not skewed by a crud burst or other similar abnormal event.

REFERENCES 1. 10 CFR100.11. 54.

. . .... .. Section.[ 6.3. 3 4 BWOG STS B 3.4.16-5 Rev. 3.0, 03/31/04.

Attachment 1, Volume 9, Rev. 1, Page 387 of 418

Attachment 1, Volume 9, Rev. 1, Page 388 of 418 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. 1, Page 388 of 418

Attachment 1, Volume 9, Rev. 1, Page 389 of 418 Specific No Significant Hazards Considerations (NSHCs)

Attachment 1, Volume 9, Rev. 1, Page 389 of 418

Attachment 1, Volume 9, Rev. 1, Page 390 of 418 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. 1, Page 390 of 418

Attachment 1, Volume 9, Rev. 1, Page 391 of 418 ATTACHMENT 17 ITS 3.4.17, STEAM GENERATOR (SG) TUBE INTEGRITY Attachment 1, Volume 9, Rev. 1, Page 391 of 418

, Volume 9, Rev. 1, Page 392 of 418 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) , Volume 9, Rev. 1, Page 392 of 418

Attachment 1, Volume 9, Rev. 1, Page 393 of 418 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 foreach SG 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 next refueling outage or SG tube. inspection,fr be in HOT STANDBY within the ACTION B n:ext 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

2. Plug or repair the affected tube(s) in accordance with the Steam Generator

. ACTION A Program prior to entering HOT SHUTDOWN following the next refueling outage or SG tube inspection.

ACTION B b. With SG 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 SG tube integrityin accordance with the Steam Generator Program.

SR 3.4.17.2 4.4.5.2 Verify that each inspected SG 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 31/44-6 Amendment No.-,-24-8-2-77 62;-

(next page is 3/4 4-13) .-- l-4t 4 43*-l-,- 4 ,-9 -27a-a0;-

Z26,-2-5-2 276 Page 1.of 1 Attachment 1, Volume 9, Rev. 1, Page 393 of 418

Attachment 1, Volume 9, Rev. 1, Page 394 of 418 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. 1, Page 394 of 418

Attachment 1, Volume 9, Rev. 1, Page 395 of 418 Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs)

Attachment 1, Volume 9, Rev. 1, Page 395 of 418

Attachment 1, Volume 9, Rev. 1, Page 396 of 418 CTS SG Tube Integrity 3,4,17 3.4 *REACTOR COOLANT SYSTEM (ROS) 3A4.17 SteamGenerator (SG) TUbe Integrity 3.4.5 LCO 3.4.17 ,SG tube integrity Shall, be maintained.

AND

,Al!, SG tubes satisfying the tube repair criteria shall be pluggedoor

repaire~din ac.crdance: with the Steam Generator Program. 0 APPLICABILITY: MODES, 1; 2, 3, and 4.

ACTIONS

.. r...o.t.n.. ---------- -- NOTE----

Separate. Condition~entry is allowed for -each SG~tube, 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 accordance With the refueling outage or SG tube inspection.

0

.Steam Generator Program. AND A,2 P,Iug[wr repairlthe affected tube(s) in accordance with Prior to entering MODE 4 following the 0

the.Steam Generator next refueling outage Program. or SG tube inspection.

Actions a. 1 B, Required Action and B13t 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 I 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.

BVWOG STS 314.17-1 Rev. 3.1, 12/01/05 Attachment 1, Volume 9, Rev. 1, Page 396 of 418

Attachment 1, Volume 9, Rev. 1, Page 397 of 418 CTS

SG Tube Integrity 3.4.17 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY 4.4.5.1 SR 3.41711 VerifySG tube integrity in accordance with the. In accordance Steam Generator Program. with the Steam Generator Program 4.4.5.2 SR 314:17.2 Verify that each inspected SG tube that'satisfies the Prior to entering tube repair criteria, is plugged Por repaiiredin accordance withthie Steam Generator Program.

MODE 4 following a SG tube 0

inspection BWOG;STS 3.4.17-2 Rev. 3.1, 12/01105 Attachment 1, Volume 9, Rev. 1, Page 397 of 418

Attachment 1, Volume 9, Rev. 1, Page 398 of 418 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.

S Davis-Besse' Page 1 of 1 Attachment 1, Volume 9, Rev. 1, Page 398 of 418

Attachment 1, Volume 9, Rev. 1, Page 399 of 418 Improved Standard Technical Specifications (ISTS) Bases Markup and Justification for Deviations (JFDs)

Attachment 1, Volume 9, Rev. 1, Page 399 of 418

Attachment 1, Volume 9, Rev. 1, Page 400 of 418 SG Tube Integrity.

B

.. 1,7 B 3.4 REACTOR COOLANT SYSTEM (RCS)

B13.4.17 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 removalfunction'is addressed by LCO 3.4.4, "RCS Loops - MODES 1 and 2," LCO3.4.5, "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, 0

and o erational LEAKAGE. The SG performance criteria are described in Specification 5. -. Meeting the SG performance criteria provides 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, 12101/05 Attachment 1, Volume 9, Rev. 1, Page 400 of 418

Attachment 1, Volume 9, Rev. 1, Page 401 of 418 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 tosecondary 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 viaksafety valves hnd th Vmajority dischargeo the mainondense. mainsteam 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 notto rupture.) In these analyses, the steam discharge to the SGs of [M ga lion atmosphere per minutel is based on the Pr is a~ss*Oed total primary totosecondary increase to 11 gallon from~all LEAKAGE e (0 Sminute] ap;a result-of accident induce"d conditions For a 5,dients that ýdo/

1not involv~puýl da magqe, the -primarycoolant activity levy, e f FDO-SE equivalent to 1%'* EQUIVALENT 1-131 is assumed to be equal to the LCO/.4.16, "RCS

/ failed fuel in the Specific Agfivity," limits. For accidents that assume fil damage, the accident analyses primary olant activity is a function of the amountg activity released from the amaged fue. The dose consequences of these events are

.within the limits of GDC 19 (Ref. 2), 10 CFR 100 (Ref. 3) orthe 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 plugged*or repaired*]in accordance with the Steam Generator Program.

During an SG inspection, any inspected tube that satisfies the Steam Generator Program repair criteria islrepaired or)removed from service by plugging. If a tube was determined to satisfy the repair criteria but was not plugged[E0r 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 wallpnd any repairs made to itj between the tube-to-tubesheet weld at the tube inlet and the tube-to-0 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. 8 The SG performance criteria are defined in Specification 5.5.

lGenera rrogram," and describe acceptable SG tube performance-The Steam Generator Program also provides the evaluation process for determining conformance with the SG performance criteria.

0 BWOG STS B 3.4.17-2 Rev. 3.1, 12/01/05 Attachment 1, Volume 9, Rev. 1, Page 401 of 418

Attachment 1, Volume 9, Rev. 1, Page 402 of 418 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 rneet 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 effecton 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 factorsand applicable design basis loads based on ASME Code, Section It!, 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 accident analysis assumptions. The accident analysis assumes that accident induced leakage does not exceedm1 gpm prS exigept Tor specific types o/f/degradation at spec~yc location_

etheRCIre has approved gr,eater accident induce; leakag 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.

BVWOG STS B 3.4.17-3 Rev. 3.1, 12101/05 Attachment 1, Volume 9, Rev. 1, Page 402 of 418

Attachment 1, Volume 9, Rev. 1, Page 403 of 418 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, "'RCSOperational 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 [ýr repaired]in accordance with the Steam Generator Program as required by SR 3.4,17,2. 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 or repaired]has tube integrity, an evaluation must be completed that 0 demonstrates that the. SG performance criteria will continue to be met until the next refueling outage or SG tube inspection. The tube integrity BWOG STS B 3.4.17-4 Rev. 3.1, 12/01/05 Attachment 1, Volume 9, Rev. 1, Page 403 of 418

Attachment 1, Volume 9, Rev. 1, Page 404 of 418 SG Tube Integrity B 3.4.17 BASES ACTIONS tcoitinued) 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.

If the 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 re paired*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 If the 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.

BWOGSTS B 3.417-5 Rev. 3.1, 12101/05 Attachment 1, Volume 9, Rev. 1, Page 404 of 418

Attachment 1, Volume 9, Rev. 1, Page 405 of 418

SG Tube Integrity B 3.4.17 BASES SURVEILLANCE REQUIR EMENTS (continued)

The Steam:Generator Program determines the-scope of the inspection and the methods used to determine whether the tubes contain flaws satisfying thetube.:repair criteria. Inspection-scope (i.e., which tubes or areas oftubing 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 usesinformation 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 5.5.015intainsdI i 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 isLrepaired olrremoved from service by plugging. The tube repair criteria delineated in Specification 5.5 re -

intended to ensure that tubes accepted for continued service satisfy the SG performance criteria with allowance for error in the flaw size measurement andfor 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.

[JSteam:generator tube repairs are only performed using approved repair methods as described in the Steam Generator Program.E 11 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 pluggedPor repaired]prior to subjecting the SG tubes to significant primary to secondary pressure differential.

hI BWOG STS B 3.4.17-6 Rev. 3.1, 12101/05 Attachment 1, Volume 9, Rev. 1, Page 405 of 418

Attachment 1, Volume 9, Rev. 1, Page 406 of 418 SG Tube Integrity B 3A4A17 BASES REFERENCES . .NEt 976'o steam GeIneratorProgram Guidelines;"

2. 10 CFR 50 Appendix A, GDC 19.

3. *10.0FR .:100*

4. ASME Boiler and Pressure VesselCode,Section III, Subsection NB.

5. Draft-RegulatoryGuide'.1121, "Basis for Plugging Degraded Steam Generator Tubes," August: 1976.

6. EPRI ",Pressurized,WNater Reactor Steam Generator Examination Guidelines."

BWOG STS B 3.4.17-7 Rev. 3.1, 12101/05 Attachment 1, Volume 9, Rev. 1, Page 406 of 418

Attachment 1, Volume 9, Rev. 1, Page 407 of 418 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. Not used.

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. 1, Page 407 of 418

Attachment 1, Volume 9, Rev. 1, Page 408 of 418 Specific No Significant Hazards Considerations (NSHCs)

Attachment 1, Volume 9, Rev. 1, Page 408 of 418

Attachment 1, Volume 9, Rev. 1, Page 409 of 418 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. 1, Page 409 of 418

, Volume 9, Rev. 1, Page 410 of 418 ATTACHMENT 18 Relocated Current Technical Specifications , Volume 9, Rev. 1, Page 410 of 418

Attachment 1, Volume 9, Rev. 1, Page 411 of 418 CTS 314.4.10.1, ASME CODE CLASS 1, 2, AND 3 COMPONENTS Attachment 1, Volume 9, Rev. 1, Page 411 of 418

, Volume 9, Rev. 1, Page 412 of 418 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) , Volume 9, Rev. 1, Page 412 of 418

Attachment 1, Volume 9, Rev. 1, Page 413 of 418 CTS 3/4.4.10.1 REA R COOLAUNM SYSTEM

.4.10 S tuCtuRALI R ASS =EO CLASS 2jand3 0 PONENTS L;MITING COND ON FOR 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 MOD ACTION:!"

a. With the structural inte rity of any ASME Code Class 1 component(s) not conforming to the

ýabove requirements, re fore the structural integrity of the affected co ponent(s) to within its limit or isolate the affe ted component(s) prior to increasing the Reac r Coolant System temperature more tha 50. TF above the minimum temperature requi by NDT considerations.

b. With th e structural in grity of any ASME Code Class 2 component( ) not conforming to the ROl above requirements, estore the structural integrity of the affected c mponent(s) towithin its limit or isolate the af ected component(s) prior to increasing the R tor Coolant System temperature above 2 T.

c. With the structural i tegrity of any ASME Code Class 3 compone (s) not conforming to the

above requirementss restore the structural integrity of the compon t(s) to within its limit or

.isolate the affected aomponent(s) from service.

d. The provisions of pecification 3.0.4 are not applicable.

JVEILANC R~E U REANTS 4.4.10.1 In addition to t requirements of Specification 4.0.5:

a, inservice-inspection of each reactor coolant pump flywheelshall be performed at least once every 10 years. The inservice inspection shall be either an ultrasonic examination of the circle of one-half the outer radius, or a See ITS volume from the inner bore of the flywheel to the flywheel. The recommendations 5.5 surface examination of exposed surfaces of the disassembled 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 1 3/4 4-30 Amendment No. 232 0 Page 1 of 2 Attachment 1, Volume 9, Rev. 1, Page 413 of 418

Attachment 1, Volume 9, Rev. 1, Page 414 of 418 CTS 3/4.4.10.1 REACTOR C0OLANT'SYSTEM

.SURVEILLANCE REQUIREMENTS (Continued)

b. Each internals vent valve hail be demonstrated OP1ERABLE at least one per

.24 months* during shutd vn by:.

1. Verifying through vi al inspection that the valve body and valve di, c exhibit no abnormal 'degradatio,

2. Verifying the valve i not stuck in an open position, and.

3. Verifying through nrmual actuation that the, valve is fully open wh n a force of

< 400 lbs. Js applie 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 cquirement may be delayed until March 25, 2 6.

DAVIS-BESSE, UNIT .1 3/4 4-31 Amendment No. 23,95, 165, 268 Page 2 of 2 Attachment 1, Volume 9, Rev. 1, Page 414 of 418

Attachment 1, Volume 9, Rev. 1, Page 415 of 418 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. 1, Page 415 of 418

Attachment 1, Volume 9, Rev. 1, Page 416 of 418 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. 1, Page 416 of 418

Attachment 1, Volume 9, Rev. 1, Page 417 of 418 Specific No Significant Hazards Considerations (NSHCs)

Attachment 1, Volume 9, Rev. 1, Page 417 of 418

Attachment 1, Volume 9, Rev. 1, Page 418 of 418 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. 1, Page 418 of 418

v t e Letter Sequence Request
TAC:MD6398, Increase Flexibility in Mode Restraints, Addition of LCO 3.0.8, Inoperability of Snubbers, Deletion of E BAR Definition and Revision to RCS Specific Activity Tech Spec, Changes to Reflect Revision of 10 CFR 50.55A, Limit Inservice Testing Program SR 3.0.2 Application to Frequencies of 2 Years or Less, Eliminate Second Completion Times Limiting Time from Discovery of Failure to Meet an LCO, Correct Example 1.4-1, Containment Structural Integrity, TSTF-28, Correct LCO 3.0.6 Bases, Add Restriction for Lowest LOOP Tave During Mode 2 Physics Testing (Approved, Closed) TAC:MD6319, DC Electrical Rewrite - Update to TSTF-360, Eliminate Use of Term Core Alterations in Actions and Notes, Provide Actions for One Steam Supply to Turbine Driven Afw/Efw Pump Inoperable, DC Electrical Rewrite, Clarification of Required Features in 3.8.1 Actions (Approved, Closed)
Initiation Request, Request, Request, Request, Request, Request, Request Acceptance, Acceptance, Acceptance, Acceptance Supplement, Supplement Administration Meeting, Meeting, Meeting Questions 18 June 2008 20 June 2008 Answers 23 July 2008 3 September 2008 Results Approval Other: L-08-161, Copy of Applicable Portions of the NRC and Davis-Besse Improved Technical Specifications Conversion Website - Sections 3.0, 3.1, 3.2, 3.4, and 3.6 and Chapter 4.0, L-08-241, Attachment 1, Applicable Portions of the U.S. Nuclear Regulatory Commission (NRC) and Davis-Besse Nuclear Power Station (DBNPS) Improved. Technical Specification (ITS) Conversion Website, Section 3.5 RAIs, L-08-338, Copy of Improved Technical Specifications (ITS), L-08-378, Implementation of Improved Technical Specification Amendment No. 279, ML073531253, ML081350505, ML082000451, ML082270661, ML082270662, ML082270663, ML082270664, ML082270665, ML082270666, ML082270667, ML082270669, ML082270670, ML082600577, ML082840099, ML082900630, ML082900882, ML082910125
MONTHYEAR2008-02-04 [Table View]

ML081480468

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