ML051960247

Improved Technical Specifications, Volume 4, Revision 0, ITS Chapter 2,0, Safety Limits.

ML051960247
Person / Time
Site:	Monticello
Issue date:	06/29/2005
From:	Nuclear Management Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
Download: ML051960247 (25)

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IMPROVED TECHNICAL SPECIFICATIONS MONTICELLO NUCLEAR GENERATING PLANT VOLUME 4 ITS Chapter 2.0, Safety Limits Commltedro Nulear Exce

Attachment 1, Volume 4, Rev. 0, Page 1 of 24 ATTACHMENT I VOLUME 4 MONTICELLO IMPROVED TECHNICAL SPECIFICATIONS CONVERSION ITS CHAPTER 2.0 SAFETY LIMITS Revision 0 Attachment 1, Volume 4, Rev. 0, Page 1 of 24

Attachment 1,Volume 4, Rev. 0, Page 2 of 24 LIST OF ATTACHMENTS

1. ITS Chapter 2.0 Attachment 1, Volume 4, Rev. 0, Page 2 of 24

Rev. 0, Page 3 of 24 Attachment 1,Volume 4, ATTACHMENT I K)

Limits ITS Chapter 2.0, Safety K-,.

4, Rev. 0, Page 3 of 24 Attachment 1,Volume

Attachment 1, Volume 4, Rev. 0, Page 4 of 24 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs)

Attachment 1,Volume 4, Rev. 0, Page 4 of 24

C c C ITS Chapter 2.0 ITS 0 2.0 SAFETY LIMITS LIMITING SAFETY SYSTEM SElTINGS 2.1 2.1 SAFETY UMITS I I 2.1.1 A. Reactor Core Satety Limits Umltl~g Safety Systerp'Settings are Incprporated Into I Sec on, 3of the Techdlcal SpecIflcatZsi,.

~ I I a)

1. With the reactor steam dome pressure < 785 psig CD 2.1.1.1 or core flow < 10% rated core flow. 2) 3 0 Thermal power shall be 5 25% Rated Thermal

-L Power 0

0 2.1.1.2 2. With the reactor steam dome pressure 2785 psig and core flow 10% rated core flow: 2 F X-0 CD 3

0 MOPR shall be a 1.10 for two recirculatlon loop 0 3 operation or 2 1.12 for single recirculation loop ES operation.

(0 2.1.1.3 3. Reactor vessel water level shall be greater than the top of active Irradiated fuel.

CD

0) 2.1.2 B. Reactor Coolant System Pressure Safely Llmt 0e M Reactor steam dome pressure shall be s 1332 pslg.

2.1/2.2 6 06111/02 Amendment No. 20, 47, 8 1, 90,100, 412rtO"r5, 128 Page 1 of 2

C C C ITS 0 ITS Chapter 2.0 2.0 SAFETY UMITS LIMING SAFETY SYSTEM SETTINGS 2.2 SAFETY LIMIT VIOLATIONS 2.2 With any Safety Umit violation, the following actions shall be A) completed within 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />s:

CI CD 2.2.1 A. Restore compliance with all Safety Umits, and 0 3

CD 2.2.2 B. Insert all Insertable control rods. 0 0 -1 0< 0 3

F 0

E, B

)

-N CD to al

0) CD CD 10 0) 0 0

-9' 2.1/2.2 7 06/11/02 Amendment No. Xh 128 Page 2 of 2

Attachment 1, Volume 4, Rev. 0, Page 7 of 24 DISCUSSION OF CHANGES ITS CHAPTER 2.0, SAFETY LIMITS ADMINISTRATIVE CHANGES A.1 In the conversion of the Monticello 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-1433, Rev. 3, "Standard Technical Specifications General Electric Plants, BWR/4' (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 Monticello Page 1 of I Attachment 1, Volume 4, Rev. 0, Page 7 of 24

Attachment 1, Volume 4, Rev. 0, Page 8 of 24 Improved Standard Technical Specifications (ISTS) Markup .

and Justification for Deviations (JFDs)

Attachment 1,Volume 4, Rev. 0, Page 8 of 24

Attachment 1, Volume 4, Rev. 0, Page 9 of 24 SLs 2.0 cTS 2.0 SAFETY LIMITS (SLs) 2.1 2.1 SLs 2.1 A 2.1.1 Reactor Core SLs 2.1 .A.1 2.1.1.1 With the reactor steam dome pressure < 785 psig or core flow < 10%

rated core flow:

THERMAL POWER shall be s 25% RTP.

2.1.A.2 2.1.1.2 With the reactor steam dome pressure 2 785 psig and core flow 2 10%

rated core flow:

MCPR shall be 2 d;for two recirculation loop operation or 2 for single recirculation loop operation.

2.1.A.3 2.1.1.3 Reactor vessel water level shall be greater than the top of active irradiated fuel.

2.1.B 2.1.2 Reactor Coolant System Pressure SL Reactor steam dome pressure shall be E psig. 0 2.2 SL VIOLATIONS 2.2 With any SL violation, the following actions shall be completed within 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />s:

2.2A 2.2.1 Restore compliance with all SLs; and 2.2.B 2.2.2 Insert all insertable control rods.

BWR/4 STS 2.0-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 4, Rev. 0, Page 9 of 24

Attachment 1, Volume 4, Rev. 0, Page 10 of 24 JUSTIFICATION FOR DEVIATIONS ITS CHAPTER 2.0, SAFETY LIMITS

1. The brackets are removed and the proper plant specific information/value is provided.

2. Changes have been made to reflect the current licensing basis value.

Monticello Page 1 of 1

- Attachment 1, Volume 4, Rev. 0, Page 10 of 24

Attachment 1, Volume 4, Rev. 0, Page 11 of 24 Improved Standard Technical Specifications (ISTS) Bases Markup and Justification for Deviations (JFDs)

Attachment 1, Volume 4, Rev. 0, Page 11 of 24

Attachment 1, Volume 4, Rev. 0, Page 12 of 24 Reactor Core SLs B 2.1.1 B 2.0 SAFETY LIMITS (SLs)

B 2.1.1 Reactor Core SLs BASES SSR BACKGROUND GDC 10 .1 -reuires andSLs ensured that specified acceptable fuel design limits are not exceeded during steady state operation, normal operational transients, and anticipated operational occurrences (AOOs).

The fuel cladding integrity SL is set such that no significant fuel damage is calculated to occur if the limit is not violated. Because fuel damage is not directly observable, a stepback approach is used to establish an SL, such that the MCPR is not less than the limit specified in Specification 2.1.1.21for [both Gener l Electric Compiny (GE) and Advanced NucleaTFuel CorporatiorY(ANF) fuel. MCPR greater than the specified limit represents a conservative margin relative to the conditions required to maintain fuel cladding integrity.

The fuel cladding is one of the physical barriers that separate the radioactive materials from the environs. The integrity of this cladding barrier is related to its relative freedom from perforations or cracking.

Although some corrosion or use related cracking may occur during the life of the cladding, fission product migration from this source is incrementally cumulative and continuously measurable. Fuel cladding perforations, however, can result from thermal stresses, which occur from reactor operation significantly above design conditions.

While fission product migration from cladding perforation is just as measurable as that from use related cracking, the thermally caused cladding perforations signal a threshold beyond which still greater thermal stresses may cause gross, rather than incremental, cladding deterioration. Therefore, the fuel cladding SL is defined with a margin to the conditions that would produce onset of transition boiling (i.e.,

MCPR = 1.00). These conditions represent a significant departure from the condition intended by design for planned operation. The MCPR fuel cladding integrity SL ensures that during normal operation and during AOOs, at least 99.9% of the fuel rods in the core do not experience transition boiling.

Operation above the boundary of the nucleate boiling regime could result in excessive cladding temperature because of the onset of transition boiling and the resultant sharp reduction in heat transfer coefficient.

Inside the steam film, high cladding temperatures are reached, and a cladding water (zirconium 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. l BWRI4 STS B 2.1.1-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 4, Rev. 0, Page 12 of 24

Attachment 1, Volume 4, Rev. 0, Page 13 of 24 B 2.1.1 0 INSERT I USAR Section 1.2.2 (Ref. 1) requires the reactor core and associated systems to be designed to accommodate plant operational transients or maneuvers that might be expected without compromising safety and without fuel damage. Therefore, 0 INSERT IA The reactor vessel water level SL ensures that adequate core cooling capability is maintained during all MODES of reactor operation. Establishment of Emergency Core Cooling System initiation setpoints higher than this SL provides margin such that the SL will not be reached or exceeded.

Insert Page B 2.1.1-1 Attachment 1, Volume 4, Rev. 0, Page 13 of 24

Attachment 1, Volume 4, Rev. 0, Page 14 of 24 Reactor Core SLs B 2.1.1 BASES APPLICABLE The fuel cladding must not sustain damage as a result of normal SAFETY operation and AOOs. The reactor core SLs are established to preclude ANALYSES violation of the fuel design criterion that ag MCPR limit is to be 0 established, such that at least 99.9% of the fuel rods in the core would not be expected to experience the onset of transition boiling.

• The Reactor Protection System setpoints (LCO 3.3.1.1, "Reactor Protection System (RPS) Instrumentation"), in combination with the other LCOs, are designed to prevent any anticipated combination of transient conditions for Reactor Coolant System water level, pressure, and THERMAL POWER level that would result in reachin the MCP limit. 0 2.1.1.1r Fuel Claddina IntearitvlrGeneral Electric-GriRanv (GE) Fuelli 0D GE critical power correlations are applicable for all critical power calculations at pressures 2 785 psig and core flows 2 10% of rated flow.

For operation at low pressures or low flows, another basis is used, as follows:

Since the pressure drop in the bypass region is essentially all elevation head the core pressure drop at low power and flows will always be > 4.6 psi. Analyses (Ref. 2) show that with a bundle flow (

of 28 x 03 lb/hr, bundle pressure drop is nearly independent of

/bundle power and has a value of 3.5 psi. Thus, the bundle flow with 6Ea4. psi driving head will be > 28 x 103 lb/hr. Full scale ATLAS test (

Isa tan ato80Qsialindicate that the 785 psm fuel assembly critical power at this flow is approximately 3.35 MWt.

With the design peaking factors, this corresponds to a THERMAL POWER > 50 % RTP. Thus, a THERMAL POWER limit of 25% RTP for reactor pressure < 785 psigis conservative. O0%rare So 2.1.1.b F el Cladding Intearitv dvanced Nuclear Fu Con oration NF) Fuell Theuse f the XN-3 correlation is valid for critical power calculations at pressure > 580 psig and bund e mass fluxes > 0.25 106 lb/hr-ft2 (Ref. 3). For operation at low ressures or low flows, he fuel cladding integrity SL is established by limiting condition on c re THERMAL POWE , with the following b sis:

Provided that the water level n the vessel downco er is maintained above he top of the active fel, natural circulation i sufficient to ensure a minim m bundle flow for all uel assemblies that h e a relatively high powet and potentially can a proach a critical heat ux condition. For the ANF ,x9 fuel design, the m nimum bundle flow is 30 x 103 lb/hr. For the ANF 8x8 fuel design, the inimum bundle flow is 28 x 103 lb/hr. For all desi ns, the coolant minim bundle flow and m ximum flow area are BWR/4 STS B 2.1.1-2 Rev. 3.0, 03/31/04 Attachment 1, Volume 4, Rev. 0, Page 14 of 24

Attachment 1, Volume 4, Rev. 0, Page 15 of 24 Reactor Core SLs B 2.1.1 BASES APPLICABLE SAFETY ANALYSES (continued) such that t mass flux is always 0.25 x 106 lb/hr-ft2. ull scale critical power test taken at pressures own to 14.7 psia mdi te that the fuel assembly ritical power at 0.25 106 lb/hr-ft2 is appro imately 3.35 MWt.

At 25% SP, a bundle power f approximately 3.35 Wt corresponds to a bundl radial peaking facto of> 3.0, which is sig ficantly higher than 0.

the ex cted peaking factor. Thus, a THERMAL P WER limit of 25% P for reactor ress tes < 785 si is con rvative.

2.1.1.2M MCPR [GEk'uellI 0D The fuel cladding integrity SL is set such that no significant fuel damage is calculated to occur if the limit is not violated. Since the parameters that result in fuel damage are not directly observable during reactor operation, the thermal and hydraulic conditions that result in the onset of transition boiling have been used to mark the beginning of the region in which fuel damage could occur. Although it is recognized that the onset of transition boiling would not result in damage to BWR fuel rods, the critical power at which boiling transition is calculated to occur has been adopted as a convenient limit. However, the uncertainties in monitoring the core operating state and in the procedures used to calculate the critical power result in an uncertainty in the value of the critical power. Therefore, the fuel cladding integrity SL is defined as the critical power ratio in the limiting fuel assembly for which more than 99.9% of the fuel rods in the core are expected to avoid boiling transition, considering the power distribution within the core and all uncertainties.

The MCPR SL is determined using a statistical model that combines all the uncertainties in operating parameters and the procedures used to calculate critical power. The probability of the occurrence of boiling transition is determined using the approved General Electric Critical Power correlations. Details of the fuel cladding integrity SL calculation are given in Reference 2. Referencel2lif includes a tabulati uncertainties used in the determination of the MCPR SL and of the nominal values of the parameters used in the MCPR SL statistical analysis.

2.1.1.2b M PR FANF Fuel The MCP SL ensures sufficie t conservatism in the perating MCPR limit that in the event of an A 0 from the limiting co dition of operation,

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ax IeddLPU.U-/O Uo LIIn THIIS0 IUeI IUU A+XPP AE19h V-%APAZ in6-^

III Lime Uore WUUIU UvrApeLeu LU avulu boiling ,ransition. The margi, between calculated boiling transition (i.e.,

MCP R= 1.00) and the MCFIR SL is based on a detailed statistical BWR/4 STS B 2.1.1-3 Rev. 3.0, 03/31/04 Attachment 1, Volume 4, Rev. 0, Page 15 of 24

Attachment 1, Volume 4, Rev. 0, Page 16 of 24 Reactor Core SLs B 2.1.1 BASES APPLICABLE SAFETY ANALYSES (continued) procedure hat considers the unc rtainties in monitoring the core operating tate. One specific uniertainty included in th SL is the uncertaint inherent in the XN-3 ritical power correlati n. Reference 3 describes he methodology use in determining the M PR SL.

The XN- critical power correla on is based on a signi cant body of practical est data, providing a igh degree of assuran e that the critical power, a evaluated by the cor elation, is within a sm IIpercentage of the actual cr cal power being esti ated. As long as the ore pressure and flow are ithin the range of validity of the XN-3 correl tion, the assumed reactor onditions used in defi ing the SL introduce nservatism into the limit be ause bounding high r dial power factors and bounding flat local peakin distributions are use to estimate the numb of rods in boiling 0 transiti n. Still further conse atism is induced by th tendency of the XN-3 c rrelation to overpredi t the number of rods i boiling transition.

These onservatisms and thinherent accuracy of e XN-3 correlation provid a reasonable degree of assurance that ther would be no transition boiling in the core uring sustained opera on at the MCPR SL.

If boili g transition were to opcur, there is reason t believe that the integrity of the fuel would n be compromised. Sionificant test data accu ulated by the NRC a d private organization indicate that the use of a oiling transition limitat on to protect against c adding failure is a very conservative approach. M ch of the data indicate that BWR fuel can survie for an extended pe iod of time in an envir nment of boiling trannition -

2.1.1.3 Reactor Vessel Water Level IDuring MODES 1 and 2 the reactor vessel water level is required to be above the top of the active fuel to provide core cooling capability. With fuel in the reactor vessel during periods when the reactor is shut down, 0

consideration must be given to water level requirements due to the effect of decay heat. If the water level should drop below the top of the active irradiated fuel during this period, the ability to remove decay heat is reduced. This reduction in cooling capability could lead to elevated cladding temperatures and clad perforation in the event that the water level becomes < 2/3 of the core height. The reactor vessel water level SL has been established at the top of the active irradiated fuel to provide a point that can be monitored and to also provide adequate margin for effective action.

BWR/4 STS B 2.1.1-4 Rev. 3.0, 03/31/04 Attachment 1, Volume 4, Rev. 0, Page 16 of 24

Attachment 1, Volume 4, Rev. 0, Page 17 of 24 Reactor Core SLs B 2.1.1 BASES SAFETY LIN1ITS The reactor core SLs are established to protect the integrity of the fuel prevent cla-dbarri-er tthe release of radioactive materials to the environs.

SL 2.1.1.1 and SL 2.1.1.2 ensure that the core operates within the fuel 0D design criteria. SL 2.1.1.3 ensures that the reactor vessel water level is greater than the top of the active irradiated fuel in order to prevent

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L APPLICABILITY SLs 2.1.1.1, 2.1.1.2, and 2.1.1.3 are applicable in all MODES.

SAFETY LIMIT Exceeding an SL may cause fuel damage and create a potential for VIOLATIONS radioactive releases in excess of 10 CFR 100, 'Reactor Site Criteria,"

limits (Ref. 4). Therefore, it is required to insert all insertable control rods and restore compliance with the SLs within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. The 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> Completion Time ensures that the operators take prompt remedial action and also ensures that the probability of an accident occurring during this period is minimal.

REFERENCES 1.1 1 FR50, pendix 00DC10.

_____~~~~I3 Z _l..t.2-2-UA.Sion 1

0D

• General Electric Standard Applicaion

2. NEDE-24011 -P-A Kla st ap oved r ision . for Reactor Fuel' (revision specifed 6nA)

Specilication 5.6.3)

3. IXN-Mz524(A). Re sion 1. Noye ber 1983.

l4 10CRNE0DE.B 1152ePes, RevsnerI Electi 2001.

0

4. 10 CFR 100. , sundle D~esigns, Revision 8, April 200 ! .

BWR/4 STS B 2.1.1-5 Rev. 3.0, 03/31/04 Attachment 1,Volume 4, Rev. 0, Page 17 of 24

Attachment 1, Volume 4, Rev. 0, Page 18 of 24 RCS Pressure SL B 2.1.2 B 2.0 SAFETY LIMITS (SLs)

B 2.1.2 Reactor Coolant System (RCS) Pressure SL BASES BACKGROUND The SL on reactor steam dome pressure protects the RCS against overpressurization. In the event of fuel cladding failure, fission products are released into the reactor coolant.. The RCS then serves as the primary barrier in preventing the release of fission products into the i atmosphere. Establishing an upper limit on reactor steam dome pressure ensures continued RCS integriLy,/Tccordiing to 10 CFR 50, Appendix A, CGDC 14, "Reactor C o lnt ressure Boundary," and GDC 5,"Reactor Coolant System Desigji "(Ref. 1), the reactor coolant pres ure boundary (RCPB) shall be designed with sufficient margin to ensure that the design conditions are not e eeded dun q no mal operation and anticipated operational occurrences (AOOs). 3for the pressure vessel anticipated operational occurrences (ADOs)

During normal operation and~ RCS pressure is limited from exceeding the design pressure by more than 10%, in accordance with Section III of the ASME Code (Ref. 2)y To ensure system integrity, all (S)

RCS components are hydrostatically tested at 125% of design pressure, in accordance with ASME Code requirements, prior to initial operation when there is no fuel in the core. Any further hydrostatic testing with fuel in the core may be done under LCO 3.10.1, "Inservice Leak and Hydrostatic Testing Operation." Following inception of unit operation, RCS components shall be pressure tested in accordance with the requirements of ASME Code, Section Xl (Ref. 3). Ireadorcoolant pressureboundary(l Overpressurization of the RCS could result in a breach of the4RCPB, (6 reducing the number of protective barriers designed to prevent radioactive releases from exceeding the limits specified in 10 CFR 100, "Reactor Site Criteria" (Ref. 4). If this occurred in conjunction with a fuel cladding failure, fission products could enter the containment atmosphere.

APPLICABLE The RCS safety/relief valves and the Reactor Protection System Reactor SAFETY Vessel Steam Dome Pressure - High Function have settings established ANALYSES to ensure that the RCS pressure SL will not be exceeded.

The RCS pressure SL has been selected such that it is at a pressur

'below which it can be shown that the integrity of the system is not endangered. The reactor pressure vessel is designedlto Section III of the ASME, Boiler and Pressure Vessel Code, 1 ditions including summeror1966 1335 1 Addenda through thel[winte76f19721 (Ref. 5), which permits a maximum estransient of 110%, 1375 psig, odesign pressure 1250 psig.

I psig, f as measured in the reactor steam dome, is BWR/4 STS B 2.1.2-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 4, Rev. 0, Page 18 of 24

Attachment 1, Volume 4, Rev. 0, Page 19 of 24 B 2.1.2 Q INSERT 2 According to USAR Section 4.2.1 (Ref. 1), the reactor vessel design pressure of 1250 psig was determined by an analysis of margins required to provide a reasonable range for maneuvering during operation, with additional allowances to accommodate transients above the operating pressure without causing operation of the safety/relief valves. In addition, the reactor vessel was also designed for the transients which could occur during the design life.

Insert Page B 2.1.2-1 Attachment 1, Volume 4, Rev. 0, Page 19 of 24

Attachment 1, Volume 4, Rev. 0, Page 20 of 24 RCS Pressure SL B 2.1.2 BASES APPLICABLE SAFETY ANALYSES (continued) se equivalent to 1375 psig at the lowest elevation of the RCS. The RCS is designed to the USAS Nuclear Power Piping Code, Sectin B31.

03-[@ Editiorq, including Addenda throughj[July, 1970] (Ref. 6), for the 0 design piping, which permits a maximum pressure transient I 4 500 psigf of design pressures of 11250 psigjpr-strcion piping51nd2l' charge piind. The RCS pressure SL is selected to be.

the lowest transient overpressure allowed by the applicable codes.

}0 SAFETY LIMITS The maximum transient pressure allowable in the RCS pressure vessel under the ASME Code, Section 1II,is 110% of design pressure. The maximum transient pressure allowable In the RCS piping, valves, and INSERT 5 fittNings ~is@ of design pressures of 1`250 psig for-smon Dipingnd E 11500 osiag F-dishargE~_ The most limiting of these allowances is 3 i/o of thesu onl pipingfdesign pressure; therefore, the SL on 11332 communicating with the maximum allowable RCS pressure is established at n3251psig as vessel steam space measured at the reactor steam dome.

APPLICABILITY SL 2.1.2 applies in all MODES.

SAFETY LIMIT Exceeding the RCS pressure SL may cause immediate RCS failure and VIOLATIONS create a potential for radioactive releases in excess of 10 CFR 100, "Reactor Site Criteria," limits (Ref. 4). Therefore, it is required to insert all insertable control rods and restore compliance with the SL within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

The 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> Completion Time ensures that the operators take prompt remedial action and also assures that the probability of an accident occurring during this period is minimal.

t7l\

REFERENCES 1. IOCFR50,Append jx DRe GODC 15,and GDC281 LO

2. ASME, Boiler and Pressure Vessel Code, Section 1II, Article NB-7000.

3. ASME, Boiler and Pressure Vessel Code,Section XI, Article IW-5000.

4. 10CFR100. - H95

5. ASME, Boiler and Pressure Vessel Code, Section I1I, R]13EditiorQ, 0 Addendal[winte 1f972 um o19

6. ASME, USAS, Nuclear Power Piping Code, Section B31.1, [i 9 Editiorn, Addenda J[July 1970.

s 1 BWR/4 STS B 2.1.2-2 Rev. 3.0, 03/31/04 Attachment 1, Volume 4, Rev. 0, Page 20 of 24

Attachment 1, Volume 4, Rev. 0, Page 21 of 24 B 2.1.2 INSERT 3 1110 psig for piping communicating with the vessel steam space INSERT 4 1136 psig for piping communicating with the bottom of the vessel INSERT 5 1110 psig for piping communicating with the vessel steam space and 1136 psig for piping communicating with the bottom of the vessel Insert Page B 2.1.2-2 Attachment 1,Volume 4, Rev. 0, Page 21 of 24

Attachment 1, Volume 4, Rev. 0, Page 22 of 24 JUSTIFICATION FOR DEVIATIONS ITS CHAPTER 2.0 BASES, SAFETY LIMITS

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

2. The Bases has been modified to reflect the fuel used at Monticello. The Monticello reactor core does not contain Advanced Nuclear Fuel Corporation (ANF) Fuel.

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

4. A description of the reactor vessel water level SL has been added, consistent wih the Background description of the other SLs.

5. Typographical/grammatical error corrected.

6. Editorial change made for clarity.

Monticello Page 1 of 1 Attachment 1, Volume 4, Rev. 0, Page 22 of 24

Attachment 1, Volume 4, Rev. 0, Page 23 of 24 Specific No Significant Hazards Considerations (NSHCs)

Attachment 1, Volume 4, Rev. 0, Page 23 of 24

• Attachment 1, Volume 4, Rev. 0, Page 24 of 24 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS CHAPTER 2.0, SAFETY LIMITS There are no specific NSHC discussions for this Specification.

Monticello Page 1 of 1 Attachment 1, Volume 4, Rev. 0, Page 24 of 24