ML082270664

Improved Technical Specification Conversion License Amendment Request, Volume 4, Revision 1, ITS Chapter 2.0, Safety Limits

ML082270664
Person / Time
Site:	Davis Besse
Issue date:	08/07/2008
From:	FirstEnergy Nuclear Operating Co
To:	Office of Nuclear Reactor Regulation
References
L-08-240, TAC MD6398
Download: ML082270664 (35)

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DAVIS-BESSE NUCLEAR POWER STATION UNIT 1 IMPROVED TECHNICAL SPECIFICATION CONVERSION LICENSE AMENDMENT REQUEST VOLUME 4 (Rev. 1)CHAPTER 2.0 -SAFETY LIMITS Attachment 1, Volume 4, Rev. 1, Page i of i Summary of Changes ITS Chapter 2.0 Change Description JI Affected Pages The changes described in the Davis-Besse Page 8 response to question 200711281431 have been made. This change clarifies DOC M01 to reference CTS Figure 2.1-1.The changes described in the Davis-Besse Pages 19, 21, and 30 response to question 200712101402 have been made. This change corrects the usage of the words setpoints and Allowable Values.The changes described in the Davis-Besse Pages 6, 8, and 14 response to question 200712101412 have been made, except that the term "Safety Limit Met" has also been changed back to "Acceptable Operation" in ITS Figure 2.1.1-1. This change adds back into the ITS Figure the representation of the RPS Trips.The changes described in the Davis-Besse Page 19 response to question 200712101433 have been made. This adds back into the Bases an additional description of SL 2.1.1.2. In addition, the change incorporates the NRC reviewer's comment concerning a typo in the draft markup posted in the database.The changes described in the Davis-Besse Pages 22 and 27 response to question 200712101440 have been made. This is an editorial correction to a Reference to 10 CFR50.36.The changes described in the Davis-Besse Page 28 response to question 200712101444 (1 st one) have been made. The Reference has been modified to be less explicit.Added titles for UFSAR Appendix 3D references in Pages 23 and 28 the Bases. This Is the approach Davis-Besse chose to address the editorial comments made by RAI's 200711301132, 200711301150, 200712101441, 200712101442, and 200712101344.

Typographical error corrected in DOC M01 (word Page 8"the" deleted from second paragraph, fifth sentence).

I>.-e.Page 1 of 1 Attachment 1, Volume 4, Rev. 1, Page i of i Attachment 1, Volume 4, Rev. 1, Page 1 of 33 ATTACHMENT 1 VOLUME 4 DAVIS-BESSE IMPROVED TECHNICAL SPECIFICATIONS CONVERSION ITS CHAPTER 2.0 SAFETY LIMITS (SLs)Revision 1 0 0 Attachment 1, Volume 4, Rev. 1, Page 1 of 33 Attachment 1, Volume 4, Rev. 1, Page 2 of 33* LIST OF ATTACHMENTS

1. ITS Chapter 2.0 0 Attachment 1, Volume 4, Rev. 1, Page 2 of 33 Attachment 1, Volume 4, Rev. 1, Page 3 of 33 VATTACHMENT 1 ITS 2.0, SAFETY LIMITS (SLs)0 0 Attachment 1, Volume 4, Rev. 1, Page 3 of 33 Attachment 1, Volume 4, Rev. 1, Page 4 of 33 WCurrent Technical Specification (CTS) Markup and Discussion of Changes (DOCs)0 0 Attachment 1, Volume 4, Rev. 1, Page 4 of 33 Attachment 1, Volume 4, Rev. 1, Page 5 of 33 ITS Chapter 2.0 ITS 2.0 SAFETY LMAITSIAIND4MITING SAFEý SYSTEM SEDL LS]2.1.1 2.1.1.2 2.1 SAFETY LIMITS REACTOR CORE 2.1. The .combination ofthe reactorcoolant core outlet pressure and outlet temperature shall not exceed the safety limit shown in.Figure 2.1-1.APPLICABILITY:

MODES I and 2.ACTION: Whenever the point definedby thecombination ofreactor coolant core outlet pressure and outlet temperature has exceeded the safety limit, be in HOT STA.NDBY within one hour.Srestore RCS pressure and temperature and REACTOR CORE 2.2.2 2.1.1 MOl 121.2 The combination of reactor THERMAL POWER and AXIAL POWER LMBA IILANCE 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.

APPLICABILIT:MD 0 M02 2.2.1 ACTION: Whenevcr the point defined by the comnbination of Reactor Coolant System flow, AXIAL POWER '-MBALANCE, and THERMvlAL POWER has exceeded the appropriate protective limit, be in HOT STANDBY within one hour: REACTOR COOLANT SYSTEMPRESSURE 2.1.2 2.1.3 The Reactor Coolant System pressure shall not exceed 2750 psig.

MODES 1, 2, 3, 4 and 5.ACTION: 2.2.3 2.2.4 MODES I and 2 -MODES 3, 4 and 5 -DAVIS-BESSIE, UN'IT Whenever the Reactor Coolant System pressure has~excueded 2750 psig, be in HOT STANDBY with the Reactor Coolant System pressure within its limit within one hour, Whenever the Reactor Coolant System pressure has exceeded.2750. psig, reduce the Reactor Coolant System pressure to within its limit within 5 minutes..I 0 2-1 Amendment 272 Page 1 of 3 Attachment 1, Volume 4, Rev. 1, Page 5 of 33 Attachment 1, Volume 4, Rev. 1, Page 6 of 33 ITS Chapter 2.0 OITS0 Figure 2.1.1-1 gu%!:2 I RL I 1i 'ir COrc Sar' tlv Limit.9)2500 2400 2300'2200 2100o 2000 1900 R 1800 1700 580 600 :610 620 630 Reactor Outlet TeMper;ture, F!) A V IS -r-SSE, UgN IT Aine, J::,wnt..Nv.

I I, 33,,ý45, 123. 12S, ;492`4 0 Page 2 of 3 Attachment 1, Volume 4, Rev. 1, Page 6 of 33 Attachment 1, Volume 4, Rev. 1, Page 7 of 33 ITS Chapter 2.0 0 0 Page 3of 3 Attachment 1, Volume 4, Rev. 1, Page 7 of 33 Attachment 1, Volume 4, Rev. 1, 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.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 (i.e., CTS Figure 2.1-1) 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 (keff < 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 MODE 3 does not specifically establish the conditions consistent with restoring the limits to within the requirements of CTS Figure 2.1-1 (ITS Figure 2.1.1-1), 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.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.Davis-Besse Page 1 of 2 Attachment 1, Volume 4, Rev. 1, Page 8 of 33 Attachment 1, Volume 4, Rev. 1, Page 9 of 33 DISCUSSION OF CHANGES ITS 2.0, SAFETY LIMITS (SLs)RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES None LESS RESTRICTIVE CHANGES None 0 Davis-Besse Page 2 of 2 Attachment 1, Volume 4, Rev. 1, Page 9 of 33 Attachment 1, Volume 4, Rev. 1, Page 10 of 33 p Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs)0 Attachment 1, Volume 4, Rev. 1, Page 10 of 33-V Attachment 1, Volume 4, Rev. 1, Page 11 of 33 SLs 2.0.CTS 2.0 SAFETY LIMITS (S.Ls)2.1 SLs 22.1. 1 ReactorCoreSLs.

2.1.2 2.1.1 2.1.1.1 In MODES 1 and 2dthe maximum ioc I fuel pin centerinet mperature:sha'll be [5080- (.x0MD/U)Ff]., 2.1.1.2 In M ES 1 and 2, the departur from nucleate boiling rtio shall be mai ained greater than the limi of [1.3.-for the: BAW-2 correlation and 1.1 for the BWC correlation].

0 0 2.1.1~In.MODES 1 ;and 2, Reactor Coolant System (RCS),.core outlet temperature and pressure shall bemaintained abovei and to 'the left of the SL shown in Figure 2..1-t1.2.1.2 Reactor Coolant System Pressure.

SL 2.1.3 in MODES 1,2, 3, 4, and 5, the RCS pressure shafi.be maintainecd 0 2.2 SAFETY LIMIT VIOLATIONS With any SL violation,.the following actions shall be completed:

.1.2.2.2.1 In MODE 1 or2, if SLC2.1...orS A .1l.2 is vi0lated,.be.

in MODE 3within 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 2A.1".1 is.violated, restore RCS pressure and temperature 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 I or :2, if S.L 2.1.2 is no e restore compliance within limits and be in 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 />.0 0 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 restore- RCS pressure to_275CI psig within 5 minutes.BWOG STS 2.0-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 4, Rev. 1, Page 11 of 33 Attachment 1, Volume 4, Rev. 1, Page 12 of 33 2.0 0 INSERT 1 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.

0 Insert-Page 2.0-1 Attachment 1, Volume 4, Rev. 1, Page 12 of 33 Attachment 1, Volume 4, Rev. 1, Page 13 of 33 SLs 2.0 O CTS Figure 2.1.1-1 0 Figure 2.1.1-1 (page 1 of 1)Reactor Coolant System -Departure from Nucleate Boiling Safety Limits 0 BWOG STS 2.0-2 Rev. 3.0, 03/31/04 Attachment 1, Volume 4, Rev. 1, Page 13 of 33 Attachment 1, Volume 4, Rev. 1, Page 14 of 33 2.0 0 INSERT 2 2500 2400 2300 2200 0 LM a.0)Co 2100 2000--1900--1800-1700, 580 590 600 610 620 630 Reactor Outlet Temperature (OF)640 650 0 Insert Page 2.0-2 Attachment 1, Volume 4, Rev. 1, Page 14 of 33, Attachment 1, Volume 4, Rev. 1, 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 AQOs.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 2.1.1.2.0 0 Davis-Besse Page 1 of I Attachment 1, Volume 4, Rev. 1, Page 15 of 33 Attachment 1, Volume 4, Rev. 1, Page 16 of 33 pImproved Standard Technical Specifications (ISTS) Bases Markup and Justification for Deviations (JFDs)0, 0 Attachment 1, Volume 4, Rev. 1, Page 16 of 33 Attachment 1, Volume 4, Rev. 1, Page 17 of 33 Reactor Core SLs B z2.11 B 2.0 SAFETY LIMITS (SLs)B:2..1 -Reactor Core SLs BASES UFSAR, Appendix :3D.1.6 (Ref.1)1 BACKGROUND GD 10 requires that reac cor s r specified.

ainy combination of normal cceptable fuel design, limits are not exceeded during steadtate 0 o peration, including the ý eaooandanticipated operational reflects of occurrences (AO~s). This isaccomplished by having a departure from nucleate boiling (DNB) design basis, which corresponds toa t95% -probability, at a 95% confidence level (95/95 DNB criterion) that DNIB~wi~l not occur and by requiring that the fuel centerline .temperature stays below the melting'temperature.

Sthese SLs The restrictionsof j prevent overheating of the fuelsand cladding ,and possible cladding perforation that would result in the release:of fission products to the reactor coolant. Overheating of thefuel is prevented by maintaining the; steady state peak linear heat rate (LHR),below tho" level at which fuel centerline melting occurs. Overheating of the fuel cladding is prevented by restricting fuel operation to0within the nucleate boiling regime, whereithe heat transfer coefficient, is large and.the cladding surface temperature is slightly above the coolant saturation O ~temperature.

Fuel centerline melting occurs whenthe 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 claddingiwater (zirconiumr water) 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

• activity to the reactor coolant.The proper functioning of the Reactor Protection System (RPS)e " prevents violation of the reactor core SLs.O BWOG STS B 2.11-1 Rev. 3.0,03/31104 Attachment 1, Volume 4, Rev. 1, Page 17 of 33 Attachment 1, Volume 4, Rev. 1, Page 18 of 33 0 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, ODNBR, 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. 1, Page 18 of 33 Attachment 1, Volume 4, Rev. 1, Page 19 of 33 Reactor Core SLs B 2:1.1 BASES-APPLICABLE SAFETY ANALYSES The fuel cladding must not .sustain damage: as.a result of normal operationand AQes. The reactor core SLs are established to preclude 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 hotfuel rod in the core does not experience DNBEand b. The hot fuel pellet in the core must not experience fuel centerline.

melting.The RPS setpoints in combination with all the LCOs, Wd ned.to prevent any anticipated combination of transient conditions for Reactor Coolant System (RCS) temperature, pressure, and THERMAL POWER level that would result in a, depaiture from nucleate boiling ratio (DN BR) of lessthan the DNBR limit and preclude the existence of flow instabilities.

Automaticenforcement of these reactorcore SLs is provided by the following:. , _ .0 Temperature a:., RC_ High le1 trip&b. RC_ Low Pressure 0 c.0 rj-Temperature

.d.

trip[High FluxlN umbetrof ---- -[Highe. FReactor Coolant PumpL 4vrltrip __so and (Flux -AFlux -Folow f..i I tri4 1ýr.These reactor core SLs ýrepresentj a design requirement for establishing the RPS trip setpoints identified previously.

0 SAFETY LIMITS:SL 2.1.1.1[ and SL 2.1 ..3enure that the minimum DNBR is: not less than the :safety analyses limit and that fuel centerline temperature Stays below the melting point, or the averageenthalpy in the hot leg is less than or equal to the enthalpy of saturated liquid, or the exit quality is within the limits defined by the DNBR.correlation.

In addition, SL 2.1.1.1-CK].shows the. pressure/temperature operating region that keeps the reactor from.reaching an SL when operating up to design powerý andtdefines (1th era-fe m-tna r r itt le razture-co-ncernsl.

INS 8 ERT 3 0 BWOG STS B 2.1.1-2 I Rev. 3.0, 03/31/04 Attachment 1, Volume 4, Rev. 1, Page 19 of 33 Attachment 1, Volume 4, Rev. 1, 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.0 0 Insert Page B 2.1.1-2 Attachment 1, Volume 4, Rev. 1, Page 20 of 33 Attachment 1, Volume 4, Rev. 1, Page 21 of 33 Reactor Core SLs B2.1.1 BASES SAFETY LIMITS (continued)

INSERT4 -- e Snsreserve monitoring the process variable AXIAL POWER IMBALANCE W ensurethat the core operates within the fuel WI INSERT 5 ---design criteria.

AXIAL POWER IMBALANCE protective limits N' l Rlowaladuen deiv h COn r red erebydu measurement sysptem infdependent AXIAL POWER IMBALANCE

/instrument errors not included poetv ii ie nteC L oalwfrm aue etsse I An estabEishing the poeowabteiv limtvs aIre seL Rate a n cf tA SValues are included in the Tripl = observabilif~y and instrumentation SSetpoints as described in thell Background section of le wi 1 ... .6*e. .--..... \, LCO 3.3.1 Bases. ratio nwtlnfese lhmitai-ns d

\POWER IMBALANCE protective lirnits~presre ythicrepn~ding

ýiý*' ia as~speci0fied' in,-the COLR .AIL-PWR ..7.IMBALAN.C.Eprb'tective limits are-separate adisntfrmheAXIAL

.POWER IMBALANCE operating limits definedby LCO 3.2.3, "AXIAL POWER IMBALANCE Operating Limits." The AXIAL POWER IMBALANCE operatinglimits in LCO 3.2.3, also specified in the COLR, preserve initial "co.nditions:of the safety analyses.but are not reactor core.SLs: 0 0 0 APPLICABILITY

'SL 2. 1.1 2,and SL 2.1.1. oy apply in MODES 1 and 2 because these are the only MODES in which the reactor is critical.Automaticý protectionfunctions are required to: be OPERABLE during MODES 1 and 2 to ensure-operation within the reator core SLs. The'Sautomatic protection actionsoserve to prevent RCS heatup to reactor core SL conditions orto initiate a reactor trip function, which forces the unit into MODE 3. for the or tripifunctions are specified in LCO 3.3.1. Awable Values Rstrumentation In MODES 3, 4, 5, and 6, Applicability is not required, since the reactoris not generating significant THERMAL POWER.0 0 SAFETY LIMIT VIOLATIONS The following SL violation re sare applicableto the reactor core S LS.2.2.1 and.2.2.2 If SL 2.1.1.1 SL,:1.1.2 or SL 2.1.1.3s1violated, the requirement to go to MODE 3 places the plant in a MODE in which these SLsare not applicable.

0 0 BVYOG STS B 2.1.1-3 Rev. 3.0, 03131/04 Attachment 1, Volume 4, Rev. 1, Page 21 of 33 Attachment 1, Volume 4, Rev. 1, Page 22 of 33 B 2.1.1INSERT 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 OINSERT 6 This ensures compliance with 10 CFR 50.36 (d)(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.0 Insert Page B 2.1.1-3 Attachment 1, Volume 4, Rev. 1, Page 22 of 33 Attachment 1, Volume 4, Rev. 1, 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 thesIe Sssare not applicableand reduces the probabilityof fuel damage.REFERENCES

1. 1-CF UFSAR, Appendix 3D.1.6, EI(Criterion 10 -Reactor Design I 2. INSERT 7 0 BWOG STS B 2.1.1-4 Rev. 3.0, 03/31/04 Attachment 1, Volume 4, Rev. 1, Page 23 of 33 Attachment 1, Volume 4, Rev. 1, Page 24 of 33 B 2.1.1 (O INSERT 7 BAW-101 79P-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. 1, Page 24 of 33 Attachment 1, Volume 4, Rev. 1, Page 25 of 33 RCS Pressure SL B 2.1..2 B,2.0 SAFETY LIMITS (SLs)B2.1.2 Reactor Coolant System (RCS) Pressure SL BASES fUfSAR, Appendix 3D.1.1 1 BACKGROUND According to 10 CFR 50; App ndix A, .GDC 14, "React r, Coolant IPressure Boundarý "and GDC 15. "ReactorCoolant rystemr Design/(Ref. 1), the reactor coolant pressure boundary (RCPB). design conditions are not to be exceeded duringrnormal operation nor during anticipated UFSA operational occurrences (AOOs). ýGDL2Ref." 1), IAppedx specifies that reactivity.

accidents including rod ejection do not result in damage to'the RCPBgreater 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 thedesign pressure'.by, more than 10% in orderto remain in accordance Wth, 0 Rthe design codes 3). Hence, the safety limit is 2750 psig. To ensure 0 ,system integrity, all RCS components are hydrostatically tested at 125%51 of design pressure ior to initial operation, according toithe[A M requirements.

Inserice operational .ydrotesting at 100% of dfsign design code pressure.

is also re uired~whenever t/hei,reactpr,.vessel h ead h/ s Ibeen...... rermovedri~rf othe pres~sur~e~boundýry:jo ht~aiterat tons-taiveD -Following, inceptio .of unit operatiop,.

RCS components sha (becressu re O It~elsted, in accord /nce the req~iremenits.

0f.ASME"Cod z, Seto APPLICABLE code The RCS pressurizer safety-valpes,roperating in conjunction with the SAFETY "Reactor Protection System trip settings,.ensurethatthe RCS pressure ANALYSES SL willlnot be exceede.d.

The RCS pressurizerisafety valves are sized to prevent systempressure 0 from exceeding the design pressure by more than 10%,- inmaccordance a with Section III.of the ASME Code for Nuclear Power Plant Components critical f. 2). The transient that is most influential for establishing the required r~elief capac~ity, and. hence the vailve :siz~e requirements and lift Settings, is..:a rod withdrawal liDrn th ransient.

no c/rtrol are assumed exc hth aeyve ntescn.y plan I r assumed to secondary plant safety val settin s, and nonal feedwater sup is maintained.

The overpressure protection analyses. (Ref[L and the safety.ana!yses.

Lý ((Ref. are performed using conservative assumptions relative to K)pressure control devices.O BWOG STS B 2.1.2-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 4, Rev. 1, Page 25 of 33 Attachment 1, Volume 4, Rev. 1, Page 26 of 33 B 2.1.2 (O INSERT 8 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.

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

a. Pressurizer rj operated reliefdvalVeg (PORV0., 1 ak -p flow ---- --.b. Stea e ur r.ed a veis_IsPrimary to "- secondary heat c. Control s nback o rand t owe nd _transferrsir a valve.d.. Pressurize rspray valve.0 0 0 0 SAFETY LIMIT The maximum transient pressure allowed in the RCS pressure vessel under the.ASME Code,Section II, is 110%of design pressure.

The rn maximum transient pressure allowed in th RCS piping,, valves; and ANSIfittings under4USAS, Section B3flZ(Ref.n, is.[X of design pressure.4 I ý'--,p s 11 0 0 of Ld 9,1gn pressur, therpforethe SL on maximum allowable RCS pressure is 27.50. psig.Overpressurization of the RCS can result in a breachof the RCPB. If such a breach occurs in conjunction with a fuel claddingfailure, fission products could :enter the containment atmosphere, raising concerns relative to limits on radioactive releases.specified in 10 CFR 1100,"ReactorSite Criteria" (Ref. 7)..APPLICABILITY SL2.1 .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,SLis notlapplicable in MODE 6 because the reactor vessel head closure bolts are notfully'tightened, making it. unlikely that the RCS can be-pressurized.

0 SAFETY LIMIT VIOLATIONS The following SL violation I6 smare applicable to the RCS pressure SL.*.2.2.3 Placing the unit in MODE 3 ensures compliance with 10 CFR 50.36 (d)(1)(i)(A), which requires a shutdown when safety limits are violated.I If the RCS pressure SL is violated when the reactor is in MODE 1 or 2,.the requirement is to restorewcompliance 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 (Ref`0 BWOG STS, ,B 2.1.2-2 Rev. 3.0, 03/31/04 Attachment 1, Volume 4, Rev. 1, Page 27 of 33 Attachment 1, Volume 4, Rev. 1, 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 challengesto Safety systemsis

minimized.

2.2.4 Fviola-te-d)

If the RCS pressure SLis e in MODE 3, 4,or 5, RCS pressure must be restored to within the SL value within 5 minutes.the RCS pressure SL in MODE 3, 4,or.5 is potentially more*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 compoundthe problem by~adding thermal gradient stresses to the, existing pressure;.stre~ss.

REFERENCES

1. 1QCFR5ixA GDC 15, a 1988(UFSAR, Appendix 3D.1.11, Criterion T 15 -Reactor Coolant System Design 2. ASME Boiler and Pressure Vessel Code,Section II., Article NB-7000.nd Appendix 3D.1.24, Criterion 28- _ctivity Limits .ASME Boiler and Pressure Vessel Code, Section X11 BAW-i0043, May 1.972. 0 FSAR, Section [ a1-- 00 U 16.

33, Stad nor Pres1967.7. 10CFR 100.BIAOG STS -B 2.1.2-3 Rev. 3.0, 03/31104 Attachment 1, Volume 4, Rev. 1, Page 28 of.33 Attachment 1, Volume 4, Rev. 1, Page 29 of 33 0 B 2.1.2 O INSERT 10 3. ANSI USAS B31.7 Draft, Nuclear Power Piping, February 1968 with Errata dated June 1968.0 0 Insert Page B 2.1.2-3 Attachment 1, Volume 4, Rev. 1, Page 29 of 33 Attachment 1, Volume 4, Rev. 1, 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 setpoint methodology for derivation of the Trip Setpoints is described in the Background section of the LCO 3.3.1 Bases, under "Trip Setpoints/Allowable Value." 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 for each section referring to an LCO 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.Davis-Besse Page 1 of 2 Attachment 1, Volume 4, Rev. 1, Page 30 of 33 Attachment 1, Volume 4, Rev. 1, Page 31 of 33 JUSTIFICATION FOR DEVIATIONS ITS 2.0 BASES, SAFETY LIMITS (SLs)11. The brackets have been removed and the proper plant specific information/value has been provided.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.0 Davis-Besse Page 2 of 2 Attachment 1, Volume 4, Rev. 1, Page 31 of 33 Attachment 1, Volume 4, Rev. 1, Page 32 of 33 Specific No Significant Hazards Considerations (NSHCs)0 0 Attachment 1, Volume 4, Rev. 1, Page 32 of 33 Attachment 1, Volume 4, Rev. 1, 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. 1, Page 33 of 33