ML072200461

Volume 4, Revision 0, Davis-Besse, Unit 1 - Improved Technical Specifications Conversion, ITS Chapter 2.0 Safety Limits.

ML072200461
Person / Time
Site:	Davis Besse
Issue date:	08/03/2007
From:	FirstEnergy Nuclear Operating Co
To:	Office of Nuclear Reactor Regulation
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Download: ML072200461 (33)

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Attachment 1, Volume 4, Rev. 0, Page 1 of 33 ATTACHMENT 1 VOLUME 4 DAVIS-BESSE IMPROVED TECHNICAL SPECIFICATIONS CONVERSION ITS CHAPTER 2.0 SAFETY LIMITS (SLs)

Revision 0 Attachment 1, Volume 4, Rev. 0, Page 1 of 33

Attachment 1, Volume 4, Rev. 0, Page 2 of 33 LIST OF ATTACHMENTS

1. ITS Chapter 2.0 Attachment 1, Volume 4, Rev. 0, Page 2 of 33

, Volume 4, Rev. 0, Page 3 of 33 ATTACHMENT 1 ITS 2.0, SAFETY LIMITS (SLs) , Volume 4, Rev. 0, Page 3 of 33

Attachment 1, Volume 4, Rev. 0, Page 4 of 33 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs)

Attachment 1, Volume 4, Rev. 0, Page 4 of 33

Attachment 1, Volume 4, Rev. 0, Page 5 of 33 ITS Chapter 2.0 ITS 2.0 SAFETY LIMITS IAND IITING SAFE0 SYSTEM S E 'GSý 2.1 SAFETY LIMITS 2.1.1 REACTOR CORE 2.1.1 The combination of the reactor coolant core outlet pressure and outlet temperatmre shall not 2.1.1.2 exceed the safety limit shown in Figure 2.1-1.

APPLICABILITY: MODES 1 and 2.

ACTION:

2.2.2 Whenever the point defined by the combination of reactor coolant core outlet pressure and outlet temperature has exceeded the safety limit be in HOT STANDBY within one hour.

2.1.1 restore RCS pressure and temperature andt-REACTOR CORE 3

2.1.2 The combination of reactor THER.MLAL POWER and AXIAL POWER IMBALANCE shall not exceed the protective limit shown in the CORE OPERATING LIMITS REPORT for the 2.1.1.1 various combinations of three and four reactor coolant pump operation.

APPLICABILH'Y: MO1D ACTIO:

2.2.1 Whenever the point defined by the combination of Reactor Coolant System flow, AXLkL POWER IMBALANCE and THERMAL POWER has exceeded the appropriate protective limit, be in HOT STANDBY within one hour.

REACTOR COOLANT SYSTEM PRESSURE 2.1.2 2.1.3 The Reactor Coolant System pressur. shall not exceed 2750 psig.

APPLICABILITY: MODES 1, 2, 3, 4 and 5.

ACTION:

2.2.3 MODES I and 2- Whenever the Reactor Coolant System pressure has exceeded 2750 psig, be in HOT STANDBY with the Reactor Coolant System pressure within its liirt within one hour.

2.2.4 MODES 3.4 and 5 - Whenever the Reactor Coolant System pressure has exceeded 2750 psig, reduce the Reactor Coolant System pressure to within its limit within 5 minutes.

DAVIS-BESSIE, UNIT 1' 2-I Amendmen' No. f?, 272 Page 1 of 3 Attachment 1, Volume 4, Rev. 0, Page 5 of 33

Attachment 1, Volume 4, Rev. 0, Page 6 of 33 ITS Chapter 2.0 Figure 2.1.1-1 FiL'ur 2.- I-1 Reactor Core Sat*:. Limia 2500 2400 /

2300 2200 C.

2100 2000 19001-1800 1700 580 DAVIS-BIESSE, UNIT I 2-2 ;cne.d:ntnt,. No. 11, 33, 415, Page 2 of 3 Attachment 1, Volume 4, Rev. 0, Page 6 of 33

, Volume 4, Rev. 0, Page 7 of 33 ITS Chapter 2.0 Page 3 of 3 , Volume 4, Rev. 0, Page 7 of 33

Attachment 1, Volume 4, Rev. 0, Page 8 of 33 DISCUSSION OF CHANGES ITS 2.0, SAFETY LIMITS (SLs)

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 Figure 2.1-1 provides a SL curve and includes a detail identifying the RC High Pressure Trip, RC Low Pressure Trip, RC Pressure - Temperature Trip, and the RC High Temperature Trip Allowable Values. ITS Figure 2.1.1-1 does not include the details identifying the Reactor Protection System Allowable Values.

This changes the CTS by deleting the Allowable Values from the SL curve.

This change is acceptable because the Allowable Values for the Reactor Protection System are included in ITS 3.3.1 and have not changed. The Acceptable Region is the same and is above and to the left of the SL shown in ITS Figure 2.1.1-1. This change is designated as administrative because it does not result in technical changes to the CTS.

MORE RESTRICTIVE CHANGES M01 The CTS 2.1.1 Action states that whenever the point defined by the combination of reactor coolant core outlet pressure and outlet temperature has exceeded the safety limit to be in HOT STANDBY (MODE 3) within one hour. Under the same conditions in the ITS, ITS 2.2.2 requires the restoration of RCS pressure and temperature to within limits and to be in MODE 3 within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. This changes the CTS by adding a requirement to restore the RCS pressure and temperature to within limits in addition to the requirement to be in MODE 3.

The purpose of the CTS 2.1.1 Action is to place the plant in a condition where the limits are not required to be met. This change adds an explicit requirement to restore the RCS pressure and temperature to within limits in addition to the requirement to be in MODE 3. MODE 3 is defined by a reactivity condition (kerr < 0.99) and an average reactor coolant temperature of > 280°F. Placing the plant at this reactivity state will help change the conditions of the core and reduce reactor coolant outlet temperature and place the plant within the limits. However, since the definition of the MODE 3 does not specifically establish the conditions consistent with the curve the added phrase is necessary. This change has been designated as more restrictive because a specific requirement has been added to restore RCS pressure and temperature to within limits when the SL is exceeded.

Davis-Besse Page 1 of 2 Attachment 1, Volume 4, Rev. 0, Page 8 of 33

Attachment 1, Volume 4, Rev. 0, Page 9 of 33 DISCUSSION OF CHANGES ITS 2.0, SAFETY LIMITS (SLs)

M02 CTS 2.1.2 is applicable in MODE 1. ITS 2.1.1.1 is applicable in MODES 1 and 2.

This changes the CTS by requiring the SL to be met in MODE 2.

The purpose of CTS 2.1.2 is to ensure the reactor core SL is met during plant operation in MODE 1. This limit ensures the maximum local fuel pin centerline temperature and the departure from nucleate boiling ratio limits are not exceeded. This change will require the SL to be met in MODE 2. In MODES 1 and 2, the reactor may be critical and there is a potential for violating these limits.

This change has been designated as more restrictive because it requires the SL to be met in MODE 2.

RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES None LESS RESTRICTIVE CHANGES None Davis-Besse Page 2 of 2 Attachment 1, Volume 4, Rev. 0, Page 9 of 33

Attachment 1, Volume 4, Rev. 0, Page 10 of 33 Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs)

Attachment 1, Volume 4, Rev. 0, Page 10 of 33

Attachment 1, Volume 4, Rev. 0, Page 11 of 33 SLs 2.0

\ CTS 2.0 SAFETY LIMITS (SLs) 2.1 SLs 2.1.1 Reactor Core SLs 2.1.2 2.1.1.1 In MODES 1 and 2,1the maximum lo I fuel pin centerline tt mperature 2.1.1.2 shallbe [5080- (6.5 x 10 MWD/

In M ES 1 and 2, the departur from U)°F]. /

nucleate boiling r tio shall be 0

mairained greater than the limir of [1.3 for the BAW-2 correlation and 1.1,8 for the BWC correlation].

0 2.1.1 211 In MODES 1 and 2, Reactor Coolant System (RCS) core outlet temperature and pressure shall be maintained above and to the left of the SL shown in Figure 2.1.1-1.

2.1.2 Reactor Coolant System Pressure SL 2.1.3 In MODES 1, 2, 3, 4, and 5, the RCS pressure shall be maintained _<[2751~psig. 0 2.2 SAFETY LIMIT VIOLATIONS With any SL violation, the following actions shall be completed:

2.1.2 2.2.1 In MODE 1 or 2, if SL 2.1.1.1 o is violated, be in MODE 3 within 0 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

2.1.1 Action 2.2.2 In MODE 1 or 2, if SL 2.1.1. is violated, restore RCS pressure and temperature 0

within limits and be in MODE 3 within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

2.1.3 Action 2.2.3 In MODE 1 or 2, if SL 2.1.2 is restore compliance within limits and be in 0

MODES 1 and 2 MODE 3 within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

restore RCS pressure to 0

2.1.3 Action MODES 3, 4 and 5 2.2.4 In MODES 3, 4, and 5, if SL 2.1.2 is

_1[275CI psig within 5 minutes. 0 BWOG STS 2.0-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 4, Rev. 0, Page 11 of 33

Attachment 1, Volume 4, Rev. 0, Page 12 of 33 2.0 0 INSERT I the combination of reactor THERMAL POWER and AXIAL POWER IMBALANCE shall not exceed the protective limit shown in the COLR for the various combinations of three and four reactor coolant pump operation.

Insert Page 2.0-1 Attachment 1, Volume 4, Rev. 0, Page 12 of 33

Attachment 1, Volume 4, Rev. 0, Page 13 of 33 SLs 2.0 CTS Figure 2.1.1-1 Figure 2.1.1-1 (page 1 of 1)

Reactor Coolant System Departure from Nucleate Boiling Safety Limits BWOG STS 2.0-2 Rev. 3.0, 03/31/04 Attachment 1, Volume 4, Rev. 0, Page 13 of 33

Attachment 1, Volume 4, Rev. 0, Page 14 of 33 2.0 2500 0 INSERT 2

_ _ _ _ _ I _ _ _ _ _ _ _

2400 2300 2200 SAFETY LIMIT MET (622.1, 2179.8) 0.

I_

SAFETY LIMIT 2100 CL E

Cl)

- _ _ /_

2000 (612.0, 1979.8)

SAFETY LIMIT VIOLATED 1900 (602.7,1829.8) 1800 1700 4-580 590 600 610 620 630 640 650 Reactor Outlet Temperature (OF)

Insert Page 2.0-2 Attachment 1, Volume 4, Rev. 0, Page 14 of 33

Attachment 1, Volume 4, Rev. 0, Page 15 of 33 JUSTIFICATION FOR DEVIATIONS ITS 2.0, SAFETY LIMITS (SLs)

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

2. The proper plant specific information/value has been provided.

3. Editorial change made for consistency.

4. ISTS 2.1.1.1 and ISTS 2.1.1.2 provide Reactor Core Safety Limits. ISTS 2.1.1.1 provides a maximum local fuel pin centerline temperature limit and ISTS 2.1.1.2 provides a nucleate boiling ratio limit. The Davis-Besse current licensing basis meets the above safety limits by constraining power operation within the axial power imbalance protective limits given in the COLR during normal operation and AOOs.

Therefore, ISTS 2.1.1.1 and ISTS 2.1.1.2 have been combined into a single Safety Limit, ITS 2.1.1.1, which states "In MODES 1 and 2, the combination of reactor THERMAL POWER and AXIAL POWER IMBALANCE shall not exceed the protective limit shown in the COLR for the various combinations of three and four reactor coolant pump operation." Furthermore, the current Safety Limit is a process variable that can actually be monitored by plant personnel, whereas the two Safety Limits provided in ISTS 2.1.1.1 and ISTS 2.1.1.2 cannot be monitored by any Davis-Besse process variable. Due to this change ISTS 2.1.1.3 has been changed to ITS 2.1.1.2.

Davis-Besse Page 1 of 1 Attachment 1, Volume 4, Rev. 0, Page 15 of 33

Attachment 1, Volume 4, Rev. 0, Page 16 of 33 Improved Standard Technical Specifications (ISTS) Bases Markup and Justification for Deviations (JFDs)

Attachment 1, Volume 4, Rev. 0, Page 16 of 33

Attachment 1, Volume 4, Rev. 0, Page 17 of 33 Reactor Core SLs B 2.1.1 B 2.0 SAFETY LIMITS (SLs)

B 2.1.1 Reactor Core SLs BASES _ .[UFSAR,Appendix 3D.1.6 (Ref.1)

BACKGROUND IG lORefy )requires that ra cr.L rspecified any combination of norIat acceptable fuel design limits are not exceeded duringe e operation. including the o era i0 open 0 a ran t, and lanticipated operational fects of occurrences (AOOs). This is accomplished by having a departure from nucleate boiling (DNB) design basis, which corresponds to a 95%

probability at a 95% confidence level (95/95 DNB criterion) that DNB will not occur and by requiring that the fuel centerline temperature stays below the melting temperature.

L Hs's The restrictions oý prevent overheating of the fuel and cladding and possible cladding perforation that would result in the release of fission products to the reactor coolant. Overheating of the fuel is prevented by maintaining the steady state peak linear heat rate (LHR) below the level at which fuel centerline melting occurs. Overheating of the fuel cladding is prevented by restricting fuel operation to within the nucleate boiling regime, where the heat transfer coefficient is large and the cladding surface temperature is slightly above the coolant saturation temperature.

Fuel centerline melting occurs when the local LHR, or power peaking, in a region of the fuel is high enough to cause the fuel centerline temperature to reach the melting point of the fuel. Expansion of the pellet upon centerline melting may cause the pellet to stress the cladding to the point of failure, allowing an uncontrolled release of activity to the reactor coolant.

Operation above the boundary of the nucleate boiling regime could result in excessive cladding temperature because of the onset of DNB and the resultant sharp reduction in heat transfer coefficient. Inside the steam film, high cladding temperatures are reached, and a cladding water (zirconium Iwater) reaction may take place. This chemical reaction results in oxidation of the fuel cladding to a structurally weaker form. This weaker form may lose its integrity, resulting in an uncontrolled release of IINSERT 1 activity to the reactor coolant. INSERT 2 preventsProtection The proper functioning of the Reactor System violation of (RPS)ca the reactor core SLs.

BWOG STS B 2.1.1-1 Rev. 3.0, 03/31(04 Attachment 1, Volume 4, Rev. 0, Page 17 of 33

Attachment 1, Volume 4, Rev. 0, Page 18 of 33 B 2.1.1 O INSERT I The 95 percent confidence level that DNB will not occur is preserved by ensuring that the DNBR remains greater than the DNBR design limit based on the applicable critical heat flux (CHF) correlation for the core design. In the development of the applicable DNBR design limit (Ref. 2),

uncertainties in the core state variables, power peaking factors, manufacturing-related parameters, and the CHF correlation are statistically combined to determine a statistical DNBR design limit. This statistical design limit protects the respective CHF design limit. Additional retained thermal margin may also be applied to the statistical DNBR design limit to yield a higher thermal design limit for use in establishing DNB-based core safety and operating limits.

In all cases, application of statistical DNB design methods preserves a 95 percent probability at a 95 percent confidence level that DNB will not occur.

O INSERT 2 DNB is not a directly measurable parameter during operation and therefore THERMAL POWER and reactor coolant temperature and pressure have been related to DNB using CHF correlations. The local DNB heat flux ratio, DNBR, defined as the ratio of the heat flux that would cause DNB at a particular core location to the local heat flux, is indicative of the margin to DNB.

The BWC and BHTP CHF correlations have been developed to predict DNB for axially uniform and non-uniform heat flux distributions. The BWC correlation (Ref. 2) applies to Mark-B fuel with zircaloy or M5 spacer grids. The BHTP correlation (Ref. 2) applies to the Mark-B-HTP fuel.

The minimum value of the DNBR during steady state operation, normal operational transients, and anticipated transients is limited to 1.18 (BWC) and 1.132 (BHTP). The value corresponds to a 95 percent probability at a 95 percent confidence level that DNB will not occur and is chosen as an appropriate margin to DNB for all operating conditions.

Insert Page B 2.1.1-1 Attachment 1, Volume 4, Rev. 0, Page 18 of 33

Attachment 1, Volume 4, Rev. 0, Page 19 of 33 Reactor Core SLs B 2.1.1 BASES APPLICABLE The fuel cladding must not sustain damage as a result of normal SAFETY operation and AQOs. The reactor core SLs are established to preclude ANALYSES violation of the following fuel design criteria:

a. There must be at least 95% probability at a 95% confidence level (95/95 DNB criterion) that the hot fuel rod in the core does not experience DNB and

b. The hot fuel pellet in the core must not experience fuel centerline melting. Allowable Values in LCO 3.3.1. "Reactor

• Protection System (RPS) Instrumentation," are/{

The RPS stt is ,in combinationwithalltheLCOs, designed to prevent any anticipated combination of transient conditions for Reactor Coolant System (RCS) temperature, pressure, and THERMAL POWER level that would result in a departure from nucleate boiling ratio (DNBR) of less than the DNBR limit and preclude the existence of flow instabilities.

Automatic enforcement of these reactor core SLs is provided by the following:

a. RCg Highl

b. RCg Low Pressure trip"' P E,,

c. Lr trip 0 erature Te High FluxNumbe o d. RC---Pressure trip [

~E.l Reactor Coolant PumpoP rtrip and Flux. -FL erFlu J*ow eF-tip47j (These reactor corersLs *

(Allowable Values]

representf a design requirement for establishing the RPSPFPI identified previously.

0 SAFETY LIMITS SL 2.1.1 .1 and SL 2.1.1 .-L s-ure that the minimum DNBR is not less than the safety analyses limit and that fuel centerline temperature stays below the melting point, or the average enthalpy in the hot leg is less than or equal to the enthalpy of saturated liquid, or the exit quality is 17 within the limits defined by the DNBR correlation. In addit* n, SL 2.1.1.3 shows he pr ssure/temperaturý operating region that keps the reactor from reachi an SL when opetating up to design pow , and it definest the safe opirating region frorT brittle fracture concern INSERT 3 BWOG STS B 2.1.1-2 Rev. 3.0, 03/31/04 Attachment 1, Volume 4, Rev. 0, Page 19 of 33

Attachment 1, Volume 4, Rev. 0, Page 20 of 33 B 2.1.1 O INSERT 3 The curve of Figure 2.1.1-1 is the most restrictive of all possible reactor coolant pump-maximum THERMAL POWER combinations. This curve is based on the design hot channel factors with potential fuel densification and fuel rod bowing effects.

Insert Page B 2.1.1-2 Attachment 1, Volume 4, Rev. 0, Page 20 of 33

Attachment 1, Volume 4, Rev. 0, Page 21 of 33 Reactor Core SLs B 2.1.1 BASES SAFETY LIMITS (continued)

INSERT 4 e s reserve monitoring the process variable AXIAL POWER IMBALANCE F ensure~that the core operates within the fuel ID INSERT . desi criteria.

qn

• Imeasurement AXIAL POW ER IMBALANCE protective limits M system independent AXIAL POWER

/protectivelimit given in he rts prie inthe COLR. by adjusting t e Allowable are derivedIMBALANCE

• observability the COLR to allow for and instrumentation measurement system/

Vaue 0

errorsf us 0 OperatioeWse limit re b ce with the AXIAL POWER IMBALANCE protective limits preserved by their corresponding A llowable Values RPj S st ]ý sin LC O3.3.1 ," at*r~ r r ~st-en'tr e 0 fi*k~as specified in the COLR.'The AXIAL POWER IMBALANCE protective limits are separate an istinct from the AXIAL POWER IMBALANCE operating limits defined by LCO 3.2.3, "AXIAL POWER IMBALANCE Operating Limits." The AXIAL POWER IMBALANCE operating limits in LCO 3.2.3, also specified in the COLR, preserve initial conditions of the safety analyses but are not reactor core SLs.

APPLICABILITY SL 2.1.1.1 I and SL 2.1.1 . apply in MODES 1 and 2

-on because these are the only MODES in which the reactor is critical.

Automatic protection functions are required to be OPERABLE during MODES 1 and 2 to ensure operation within the reactor core SLs. The

, automatic protection actionszserve to prevent RCS heatup to reactor core SL conditions or to initiate a reactor trip function, which 0

forces the unit into MODE 3.

specified in LCO 3.3.1. (able

• for the [reafor ri functions are Values nstmmentaton 0

In MODES 3, 4, 5, and 6, Applicability is not required, since the reactor is not generating significant THERMAL POWER.

SAFETY LIMIT The following SL violation [rsiLare applicable to the reactor core VIOLATIONS SLs.

2.2.1 and 2.2.2 If SL 2.1.1.1 or SL 2.1.1 glivolated, the requirement to go to MODE 3 places the plant in a MODE in which these SLs are not applicable. _*,N ET 6 BWOG STS B 2.1.1-3 Rev. 3.0, 03/31/04 Attachment 1, Volume 4, Rev. 0, Page 21 of 33

Attachment 1, Volume 4, Rev. 0, Page 22 of 33 B 2.1.1 0 INSERT 4 The fuel centerline melt and DNBR fuel design limits are not directly monitored by installed plant instrumentation. Instead, 0 INSERT 5 With AXIAL POWER IMBALANCE within the protective limits, fuel centerline temperature and DNBR are also within limits. Therefore, the Safety Limit is specified to be the 5 INSERT 6 This ensures compliance with 10 CFR 50.36 (c)(1)(i)(A), which requires a shutdown when safety limits are violated. In addition, if SL 2.1.1.2 is violated, the requirement is to restore the RCS pressure and temperature to within limits. Exceeding SL 2.1.1.2 may cause immediate fuel failure; therefore it is necessary to restore RCS pressure and temperature to within limits.

Insert Page B 2.1.1-3 Attachment 1, Volume 4, Rev. 0, Page 22 of 33

Attachment 1, Volume 4, Rev. 0, Page 23 of 33 Reactor Core SLs B 2.1.1 BASES SAFETY LIMIT VIOLATIONS (continued)

The allowed Completion Time of 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> recognizes the importance of bringing the plant to a MODE of operation where these SLs are not applicable and reduces the probability of fuel damage.

REFERENCES 1. 1i ", UFSAR. Appendix 3D.1.6I 0 DT7--C

2. 4FSc,-Setjofi,[j.-

BWOG STS B 2.1.1-4 Rev. 3.0, 03/31/04 Attachment 1, Volume 4, Rev. 0, Page 23 of 33

Attachment 1, Volume 4, Rev. 0, Page 24 of 33 B 2.1.1 O INSERT7 BAW-10179P-A, "Safety Criteria and Methodology for Acceptable Cycle Reload Analyses" (revision specified in Specification 5.6.3)

Insert Page B 2.1.1-4 Attachment 1, Volume 4, Rev. 0, Page 24 of 33

Attachment 1, Volume 4, Rev. 0, Page 25 of 33 RCS Pressure SL B 2.1.2 B 2.0 SAFETY LIMITS (SLs)

B 2.1.2 Reactor Coolant System (RCS) Pressure SL BASES .UFSAR.Appendix 3D.1.11I 4

BACKGROUND According toil 0 CFR 50, App*'ndix A, GDC 14, "React r Coolant LPressure Bundary." and GDQ 15, "Reactor Coolant rystem Design"/

(Ref. 1), the reactor coolant pressure boundary (RCPB) design conditions are not to be exceeded during normal o eration nor during anticipated UFSAR.

operational occurrences (AOOs). (Re 1), 3pDe1d do not result in specifies that reactivity accidents including rod ejection damage to the RCPB greater than limited local yielding.

The design pressure of the RCS is 2500 psig. During normal operation and AOOs, the RCS pressure is kept from exceeding the design pressure he design code by more than 10% in order to remain in accordance with f t

• e desi.gancodes Hence, the safety limit is 2750 psig. To ensure system integrity, all RCS components are hydrostatically tested at 125%

of design pressure nor to initial operation, according to the requirements . Inse ice operational ydrotesting at 100% of design design code Jpressure is also rec uired whenever t, e reactor vessel head h s beenn

• . removed or if othe* pressure bo undary joint a he rations have *ccurred.\

r-Following inceptio* of unit operatiop, RCS components shal be pressure Itested, in accordaince with the reaUirements of ASME Code', Section XI/

(Ref. ý3)

APPLICABLE code The RCS pressurizer safety valves, operating in conjunction with the SAFETY \ Reactor Protection System trip settings, ensure that the RCS pressure ANALYSES SL will not be exceeded.

The RCS pressurizerr safety valves are sized to prevent system pressure from exceeding the design pressure by more than 10%, in accordance mostfor that isCode of the ASME Power Nuclear for Plant Components a

is rtcl with 2). The IIItransient (Ref.Section influential establishing the required IISR condition relief capacity, and hence the valve size requirements and lift settings, is a rod withdrawal ror* *  ! During the transient, lno ctrol actions (D are assumed excef)t that the safety v vvves on the seconc~f plant are l assumed to o~pn when the stearr0ressure reaches secondary plantl safety valv/settings, and norrKal feedwater supp is maintained.

The overpressure protection analyses (Ref-R) and the safety analyses are performed using conservative assumptions relative to pressure control devices.

BWVOG STS B 2.1.2-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 4, Rev. 0, Page 25 of 33

Attachment 1, Volume 4, Rev. 0, Page 26 of 33 B 2.1.2 O INSERT8 A system leakage test at normal operating pressure is required near the end of each refueling outage. Pressure tests are performed per ASME Code,Section XI (Ref. 4) following repair or replacement activities.

0 INSERT 9 the analysis assumes full reactor coolant flow but no heat transfer out of the primary system to maximize system conditions.

Insert Page B 2.1.2-1 Attachment 1, Volume 4, Rev. 0, Page 26 of 33

Attachment 1, Volume 4, Rev. 0, Page 27 of 33 RCS Pressure SL B 2.1.2 BASES APPLICABLE SAFETY ANALYSES (continued)

More specifically, no credit is taken for operation of the following:

p -flow a. Pressurizer F5Eoperated relief valveg (PORVM)W.-.

0 0

b. ISte am-tinfet urb pef*asvTv e s,. 0 0 secondaryhea' c. Contro sy n Uacko ý ndt owe and 0 d etrinsfer a []

d. Pressurizer spray valve.

SAFETY LIMIT The maximum transient pressure allowed in the RCS pressure vessel under the ASME Code,Section III, is 110% of design pressure. The maximum transient pressure allowed in th eRCS piping, valves, and L.E =

• ANS fittings

~~The underUSAS, rost r* iting/ofthesa-to 1eo designg pressure.fZ2I-o

.0(Ref.si), is -24%9'of Section B311j6wan 0 EH'*ssur , theTtre the SL on maximum allowable RCS pressure is 2750 psig.

Overpressurization of the RCS can result in a breach of the RCPB. If such a breach occurs in conjunction with a fuel cladding failure, fission products could enter the containment atmosphere, raising concerns relative to limits on radioactive releases specified in 10 CFR 100, "Reactor Site Criteria" (Ref. 7).

APPLICABILITY SL 2.1.2 applies in MODES 1, 2, 3, 4, and 5 because this SL could be approached or exceeded in these MODES during overpressurization events. The SL is not applicable in MODE 6 because the reactor vessel head closure bolts are not fully tightened, making it unlikely that the RCS can be pressurized.

SAFETY LIMIT The following SL violation r are applicable to the RCS pressure SL.

VIOLATIONS 2.2.3 with 10 CFR 50.36 (c)(1)(i)(A), which requires a shutdown Placing when safety limits are violated.

the unit in MODE 3 ensures compliance If the RCS pressure SL is violated when the reactor is in MODE 1 or 2, I

the requirement is to restore compliance and be in MODE 3 within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

Exceeding the RCS pressure SL may cause immediate RCS failure and create a potential for radioactive releases in excess of 10 CFR 100, "Reactor Site Criteria," limits (RefI`.Z. 0 BWOG STS B 2.1.2-2 Rev. 3.0, 03/31/04 Attachment 1, Volume 4, Rev. 0, Page 27 of 33

Attachment 1, Volume 4, Rev. 0, Page 28 of 33 RCS Pressure SL B 2.1.2 BASES SAFETY LIMIT VIOLATIONS (continued)

The allowed Completion Time of 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> is based on the importance of reducing power level to a MODE of operation where the potential for challenges to safety systems is minimized.

2.2.4 va Ifthe RCS pressure SL is in MODE 3, 4, or 5, RCS pressure must be restored to within the SL value within 5 minutes.

0 at V the RCS pressure SL in MODE 3, 4, or 5 is potentially more 0 severe than exceeding this SL in MODE 1 or 2, since the reactor vessel temperature may be lower and the vessel material, consequently, less ductile. As such, pressure must be reduced to less than the SL within 5 minutes. This action does not require reducing MODES, since this would require reducing temperature, which would compound the problem by adding thermal gradient stresses to the existing pressure stress.

REFERENCES 1. A10 CFRB5operndPessArGDCe14eseDCod5eectnI,a, A 1988l

• -.*UFSAR, Appendix 3D.1.1 1 and Appendix 3D.1.24

2. ASME Boiler and Pressure Vessel Code, Sectio-n III, -ArticleNB-7000.

INSERT 10 4 -:. ASME Boiler and Pressure Vessel Code, Section XIVje 2-5aM.

10 L: 7 BAW-10043, May 1972. FExamination Category BP and IWA 4540 Code Case N or 16-3

.10 0

SectionR10

16. ASME..&-.1, Standslard-C or-Press iT1967.ý a

7. IOCFRIOO.

BWOG STS B 2.1.2-3 Rev. 3.0, 03/31/04 Attachment 1, Volume 4, Rev. 0, Page 28 of 33

Attachment 1, Volume 4, Rev. 0, Page 29 of 33 B 2.1.2 O INSERT 10

3. ANSI USAS B31.7 Draft, Nuclear Power Piping, February 1968 with Errata dated June 1968.

Insert Page B 2.1.2-3 Attachment 1, Volume 4, Rev. 0, Page 29 of 33

Attachment 1, Volume 4, Rev. 0, Page 30 of 33 JUSTIFICATION FOR DEVIATIONS ITS 2.0 BASES, SAFETY LIMITS (SLs)

1. Information related to the NRC-approved Statistical Core Design (SCD) methodology has been added. The SCD methodology is approved for use in the reload analyses performed to determine compliance with departure from nucleate boiling (DNB) acceptance criteria.

2. Typographical/grammatical correction has been made.

3. Specific details relating to the two critical heat flux (CHF) correlations approved for use for the Davis-Besse fuel designs have been included. This information is consistent with the current licensing basis.

4. Reference to the Main Steam Safety Valves (MSSVs) functioning to prevent violation of the reactor core Safety Limits (SLs) has been deleted. The Davis-Besse Updated Final Safety Analysis Report (UFSAR) does not explicitly credit the MSSVs in the safety analyses in order to ensure that the SLs are not exceeded.

5. The Davis-Besse UFSAR does not provide the Reactor Protection System (RPS) setpoints. The UFSAR refers to the Davis-Besse Technical Specifications for these trip setpoints. The Davis-Besse Technical Specifications, and the proposed ITS 3.3.1, provide the Allowable Values for the RPS trip functions. The reference to the UFSAR for these values has been deleted, and each reference to the trip setpoints has been replaced with a reference to the Allowable Values, as provided in LCO 3.3.1, "Reactor Protection System (RPS) Instrumentation."

6. The reference to RCS High Pressure trip has been deleted, and replaced with the RC High Temperature trip. The RCS High Pressure trip does not provide protection of the reactor core SLs that are based on ensuring that the DNB ratio (DNBR) limit is not exceeded. Instead, the RC High Temperature trip has been added since this trip ensures that the DNBR limit is not exceeded.

7. 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.

8. Information has been incorporated to establish that the SLs of Figure 2.1.1-1 represent the most limiting condition of Reactor Coolant System (RCS) pressure and core outlet temperature for reactor coolant pump (RCP) maximum THERMAL POWER combinations. Analyses have been performed for four RCP operation and three RCP operation that demonstrate that the four RCPs operating curve is bounding. Incorporation of this statement clarifies the acceptability of operating with less than four RCPs.

9. Specific reference to the ASME code has been deleted and replaced with a more generic reference to "design codes" to more accurately reflect the design codes applicable to the design and construction of the RCS.

10. The Davis-Besse design code for piping, valves, and fittings is ANSI B31.7, which provides for a maximum transient pressure of 110% of design pressure.

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

Davis-Besse Page 1 of 2 Attachment 1, Volume 4, Rev. 0, Page 30 of 33

Attachment 1, Volume 4, Rev. 0, Page 31 of 33 JUSTIFICATION FOR DEVIATIONS ITS 2.0 BASES, SAFETY LIMITS (SLs)

12. 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.

13. Davis-Besse was designed and under construction prior to the promulgation of 10 CFR 50, Appendix A. The design of Davis-Besse meets the intent of 10 CFR 50, Appendix A published in the Federal Register on February 20, 1971, and as amended in Federal Register on July 7, 1971. Bases references to the 10 CFR 50, Appendix A criteria have been replaced with references to the appropriate section of the UFSAR.

14. Changes are made to reflect the Specifications.

15. ISTS 2.2.2 requires the plant to "restore RCS pressure and temperature to within limits." The ISTS 2.2.1 and 2.2.2 Safety Limit Violations Bases discussion does not include a discussion to "restore RCS pressure and temperature to within limits." This change to the Bases is made to be consistent with the requirements in the ISTS. In addition, ISTS 2.2.1, 2.2.2, and 2.2.3 Safety Limit Violations Bases discussions do not include a discussion of why the plant must be shut down. Thus, the reason has been provided.

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

17. ISTS Safety Limit 2.1.1.3 does not define the safe operation region from brittle fracture concerns. The RCS Pressure and Temperature Limits in ITS 3.4.3 establish the operating limits that provide a margin to brittle failure.

Davis-Besse Page 2 of 2 Attachment 1, Volume 4, Rev. 0, Page 31 of 33

Attachment 1, Volume 4, Rev. 0, Page 32 of 33 Specific No Significant Hazards Considerations (NSHCs)

Attachment 1, Volume 4, Rev. 0, Page 32 of 33

Attachment 1, Volume 4, Rev. 0, Page 33 of 33 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 2.0, SAFETY LIMITS (SLs)

There are no specific NSHC discussions for this Specification.

Davis-Besse Page 1 of 1 Attachment 1, Volume 4, Rev. 0, Page 33 of 33