ML17264A867

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Proposed Tech Specs,Revising Rcs,Pt & Administrative Control Requirements
ML17264A867
Person / Time
Site: Ginna Constellation icon.png
Issue date: 04/24/1997
From:
ROCHESTER GAS & ELECTRIC CORP.
To:
Shared Package
ML17264A865 List:
References
NUDOCS 9705020089
Download: ML17264A867 (71)
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Attachment II Plant Marked Up Copy of R.E. Ginna Nuclear Power Technical Specifications Included Pages:

5.0-22 9705020089 970424 PDR ADQCK 05000244 P PDR )

Reporting Requirements 5.6 5.6 Reporting Requirements

~

5.6.6 PTLR (continued) c ~ The aoifjtWcighmethVds,=,.viidKCp':::-:deterp$ ne: t+e'CS pressure and ~empe~ra ure andTTOAPA Iimits shal'l be those previously reviewed and approved by the NRC.

C .i.( in NRC letter dated Hay gg,dgggii[iiii!!!!il:,. Ad III 11, 44 I 4 I gy areLs described in the following documents:

1. Letter from R.C. Hecredy, Rochester Gas and Electric Corporation (RGimLE), to Document Control Desk, NRC, Attention: A.R. Johnson, "Application for Facility C.w.i Operating License, Revision to Reactor Coolant System

~

RCS) Pressure and Tem erature Limits Re ort PTLR

" 'Attlclllllltlt'3!VI/

'A,msfstvikt1ve7!Coutp~I't!88'Qll1redmeutsiy Apri'i 2~19r9$ .

2.

Hitigating System Setpoints Limit Curves, 8Yijii'~~r,,";,5lf9,6.

IIAAP-1444 "Hethodology Used to Develop Cold Overpressure and RCS Heatup and Cooldown

".,':.PIP,-'":l1 fiictg oiis':;.!L!,.":::,:::,:.:2::."::::;::".Pe'e8~!3:-

C.< ~ L

d. The PTLR shall be provided to the NRC upon issuance for each C. i.w reactor vessel fluent period and for revisions or supplement thereto.

R.E. Ginna Nuclear Power Plant 5.0-22 Amendment No. g, g

Attachment III Proposed Technical Specifications Included Pages:

5.0-22

Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.6 PTLR (continued)

C. The analytical methods used to determine the RCS pressure and temperature and LTOP limits shall be those previously reviewed and approved by the NRC in NRC letter dated <NRC approval document>. Specifically, the limits and methodology is described in the following documents:

1. Letter from R.C. Hecredy, Rochester Gas and Electric Corporation (RGKE), to Document Control Desk, NRC, Attention: A.R. Johnson, "Application for Facility Operating License, Revision to Reactor-Coolant System (RCS) Pressure and Temperature Limits Report (PTLR)

Administrative Controls Requirements," Attachment VI, April 24, 1997.

2. WCAP-14040-NP-A, "Hethodology Used to Develop Cold Overpressure Hitigating System Setpoints and RCS Heatup and Cooldown Limit Curves," Sections 1, 2, and 4, January 1996.
d. The PTLR shall be provided to the NRC upon issuance for each reactor vessel fluence period and for revisions or supplement thereto.

R.E. Ginna Nuclear Power Plant 5.0-22 Amendment No. g, PP

Attachment IV Ginna Station PTLR, Revision 2

GINNA STATION PTLR Revision 2 RCS PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)

Responsible Hanager Effective Date Controlled Copy No.

R.E. Ginna Nuclear Power Plant RCS Pressure and Temperature Limits Report Revision 2 This report is not part of the Technical Specifications. This report is referenced in the Technical Specifications.

TABLE OF CONTENTS 1.0 RCS PRESSURE AND TEMPERATURE LIMITS REPORT ........................ 2 2.0 OPERATING LIMITS ................................................... 3

2. 1 RCS Pressure and Temperature Limits .......................... 3 2.2 Low Temperature Overpressure Protection System Enable T emperature .................................................. 3 2.3 Low Temperature Overpressure Protection Syste~ Setpoints ..... 3 3.0 REACTOR VESSEL MATERIAL SURVEILLANCE PROGRAM......................

4.0 SUPPLEMENTAL DATA INFORMATION AND DATA TABLES....................... 4 5 .0 REFERENCES ......................................................... 5 FIGURE 1 Reactor Vessel Heatup Limitations ............................ 6 FIGURE 2 Reactor Vessel Cooldown Limitations .......................... 7 TABLE 1 Surveillance Capsule Removal Schedule......... ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 8 TABLE 2 Comparison of Surveillance Material with RG l. 99 Predictions.. 9 TABLE 3 Calculation of Chemistry Factors Using Surveil lance C apsule Data.................................. 10 TABLE 4 Reactor Vessel Toughness Table (Unirradiated)

TABLE 5 Reactor Vessel Surface Fluence Values at 19.5 and 32 EFPY...... 11 TABLE 6 Calculation of ARTS at 24 EFPY............. . 12 PTLR Revision 2

R.E. Ginna Nuclear Power Plant Pressure and Temperature Limits Report 1.0 RCS Pressure and Tem erature Limits Re ort PTLR This Pressure and Temperature Limits Report (PTLR) for Ginna Station has been prepared in accordance with the requirements of Technical Specification 5.6.6.

Revisions to the PTLR shall be provided to the NRC after issuance.

The Technical Specifications addressed in this report are listed below:

3.4.3 RCS Pressure and Temperature (P/T) Limits 3.4.6 RCS Loops - NODE 4 3.4.7 RCS Loops - NODE 5, Loops Filled 3.4.10 Pressurizer Safety Valves 3.4.12 Low Temperature Overpressure Protection (LTOP) System I

PTLR Revision 2

I,I

2.0 OPERATING LIMITS The cycle-specific parameter limits for the specifications listed in Section

=

1.0 are presented in the following subsections. All changes to these limits must be developed using the NRC approved methodologies specified in Technical Specification 5.6.6. These limits have been determined such that all applicable limits of the safety analysis are met. All items that appear in capitalized type are defined in Technical Specification 1. 1, "Definitions."

2. 1 RCS Pressure and Tem erature Limits (LCO 3.4.3 and LCO 3.4. 12)

(Reference 1)

2. 1. 1 The RCS temperature rate-of-change limits are:
a. A maximum heatup of 60'F per hour.
b. A maximum cooldown of 100'F per hour.
2. 1.2 The RCS P/T limits for heatup and cooldown are specified by Figures 1 and 2, respectively.
2. 1.3 The minimum boltup temperature, using the methodology of Reference 2, Section 2.7, is 60'F.

2.2 Low Tem erature Over ressure Protection S stem Enable Tem erature (LCOs 3.4.6, 3.4.7, 3.4. 10 and 3.4. 12)

(Methodology of Reference 3, Attachment VI, Section 3.4 as calculated in Attachment VII to Reference 3).

2.2. 1 The enable temperature for the Low Temperature Overpressure Protection System is 322'F.

2.3 Low Tem erature Over ressure Protection S stem Set pints (LCO 3,4. 12) 2.3. 1 Pressurizer Power 0 crated Relief Valve Lift Settin Limits (Methodology of Reference 3, Attachment VI as calculated in Reference 4, Attachment IV)

The lift setting for the pressurizer Power Operated Relief Valves (PORVs) is s 411 psig (includes instrument uncertainty).

PTLR Revision 2

3.0 REACTOR VESSEL MATERIAL SURVEILLANCE PROGRAM The reactor vessel material irradiation surveillance specimens shall be removed and examined to determine changes in material properties. The removal schedule is provided in Table 1. The results of these examinations shall be used to update Figures 1 and 2.

The pressure vessel steel surveillance program (Ref. 5) is in compliance with Appendix H to 10 CFR 50, entitled, "Reactor Vessel Radiation Surveillance Program." The material test requirements and the acceptance standard utilize the reference nil-ductility temperature, RT>>, which is determined in accordance with ASTM E208. The empirical relationship between RT>>~ and the fracture toughness of the reactor vessel steel is developed in accordance with Appendix G, "Protection Against Non-Ductile Failure," to section III of the ASME Boiler and Pressure Vessel Code. The surveillance capsule removal schedule meets the requirements of ASTM E185-82.

As shown by Reference 1 (specifically its Reference 51), the reactor vessel material irradiation surveillance specimens indicate that the surveillance data meets the credibility discussion presented in Regulatory Guide 1.99 revision 2 where:

1. The capsule materials represent the limiting reactor vessel material.
2. Charpy energy vs. temperature plots scatter are small enough to permit determination of 30 ft-lb temperature and upper shelf energy unambiguously.
3. The scatter of a,RT>> values are within the best fit scatter limits as shown on Table 2. The only exception is with respect to the Intermediate Shell which is not the limiting reactor vessel material.
4. The Charpy specimen irradiation temperature matches the reactor vessel surface interface temperature within + 25'F.
5. The surveillance data falls within the scatter band of the material database.

4.0 SUPPLEMENTAL DATA INFORMATION AND DATA TABLES 4.1 The RT>>~ value for Ginna Station limiting beltline material is 256.6 F for 32 EFPY per Reference l.

4.2 Tables Table 2 contains a comparison of measured surveillance material 30 ft-lb transition temperature shifts and upper shelf energy decreases with Regulatory Guide 1.99, Revision 2 predictions.

PTLR Revision 2

A" L I

Table 3 shows calculations of the surveillance material chemistry factors using surveillance capsule data.

Table 4 provides the reactor vessel toughness data.

Table 5 provides a summary of the fluence values used in the generation of the heatup and cooldown limit curves.

Table 6 shows example, calculations of the ART values at 24 EFPY for the limiting reactor vessel material.

5.0 REFERENCES

1. WCAP-14684, "R.E. Ginna Heatup and Cooldown Limit Curves for Normal Operation," dated June 1996.
2. WCAP-14040-NP-A, "Hethodology Used to Develop Cold Overpressure Hitigating System Setpoints and RCS Heatup and Cooldown Limit Curves,"

Revision 2, January 1996.

3. Letter from R.C. Hecredy, RG&E, to A.R. Johnson, NRC,

Subject:

"Application for Amendment to Facility Operating License, Revision to Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR) Adminstrative Controls Requirements," dated April 24, 1997 Letter from R.C. Hecredy, RG8E, to A.R. Johnson, NRC,

Subject:

"Application for Amendment to Facility Operating License, "Hethodology for Low Temperature Overpressure Protection (LTOP) Limits," dated February 9, 1996.

5. WCAP-7254, "Rochester Gas and Electric, Robert E. Ginna Unit No. 1 Reactor Vessel Radiation Surveillance Program," Hay 1969.

I PTLR Revision 2

MATERIALPROPERTY BASIS LIMITING MATERIAL: CIRCUMFERENTIAL WELD SA-847 LIMITINGART VALUES AT 24 EFPY: 1/4T, 232'F 3/4T, 196 F 2500 g ~

~ I ' I ~ t ~

6664SSI060666 I ~, I

~ f ~ ~

l m 2250 ~ t ~

.~

~ ~ ~ I

~ IN LEA K TEST L I ICIT i I ~ ~ I i ~ t j

~ 2000 I ~ g

~ ~

1750 UNhCCEPThBLE'PERhTION

~ ~

I I I 1500 S I

CA 1250 hCCEPThBLE OPERATIO.N HBATUP RATE

-. 1000 ~ I ~

UP TO 60 F/Hr'.

HBATUP RATE UP TO IOO F/Hr.

750 500 I ~

250 CRITICALITY I.IMIT EASED Ox INSERVICE HYDROSTATIC TEST TEMPERATURE (SSS F) FOR THE SERVICE PERIOD UP TO Z4 ~ 0 EFPT 0

0 50 100 150 200 250 300 350 400 450 500 Indicated Temperature (Beg.F.)

FIGURE I REACTOR VESSEL HEATUP LIMITATIONS APPLICABLE FOR THE FIRST 24 EFPY (MITHOUT MARGIN FOR INSTRUNENT ERRORS)

PTLR Revision 2

MATERIAL PROPERTY BASIS LIMITINGMATERIAL: CIRCUMFFRENTIALVlELD SA-847 LIMITINGART VALUES AT 24 EFPY 1/4T, 232 F 3/4T, 196 F 2500 ~ ~

5004ZSl00060d ~ i I I ~ I I 2250 l .'

\

i, 'I

~

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~

~ !

I ~

i ~

~

i he

~

t I ' I

~ ~  !

~ W I I i i i i i ~

i I ! I i '. ! I 2000 ~ i I ~

i I ~

~ i I I 1750 UNhCCEPTh3LE OPERATION 1500

! I I i I I

~ I I 1250 hCCEPThBLE OPERhTION 1000 I

750 cooLDo'AN I ~

BhTES P/Hr.

5.0 0 o zo 40 00 too 250=

0 0 50 100 150 200 250 300 350 400 450 500 Indicated Temperature (Deg.p)

FIGURE 2 REACTOR VESSEL COOLDOWN LIMITATIONS APPLICABLE FOR THE FIRST 24 EFPY (WITHOUT MARGIN FOR INSTRUMENT ERRORS)

PTLR eviSion 2

Table 1 Surveillance Ca sule Removal Schedule Vessel Capsule Capsule Location Lead Fluence Capsule (deg.) Factor Removal Schedule" E19(n/cm )"

1.6 (removed) .5028 77'57 3.00 2.7 (removed) 1.105 1.85 7 (removed) 1.864 247'.99 67'7'370 1.74 17 (removed) 3.746 1.74 TeO

l'eo'b'/A 1.9 Standby NOTES:

(a) Effective Full Power Years (EFPY).

(b) To be determined, there is no current requirement for removal.

(c) Reference l.

I PTLR Revision 2

TABLE 2 Surveillance Haterial 30 ft-lb Transition Temperature Shift 30 lb-ft Transition Temperature Shift Fluence (x 10" n/cm', E > 1.0 Predicted" Heasured" Haterial Capsule HeV)" ('F) ('F) ('F)

.5028 26 25 1.105 32 25 Lower Shell 1.864 37 30 3.746 42

.5028 37 0- 37 1.105 46 Intermediate Shell 1.864 52 52 3.746 59 60 s

]

.5028 135 140 1.105 168 165 Weld Hetal 1.864 191 150 41 3.746 218 205 13

.5028 1.105 90 HAZ Hetal 1.864 100 3.746 95 (a) Reference 1 (including its Reference 51).

1 IC4

~ I'll E

r s

TABLE 3 Calculation of Chemistry Factors Using Surveillance Capsule Data Fluence (x 10'/cm',

~RT F F*h Ropy o )N(~)

Haterial Capsule E ) 1.0 FF ( F ('F) FF VeV)<>

Intermediate .5028 .8081 25 20.2 .6530 Shell Forging 05 1.105 1.0279 25 25.7 1.0566 (Tangential) 1.864 1.1706 30 35.1 1.3703 3.746 1.3418 42 56.4 1.8004 Sum: 137.4 4.8803 Chemistry Factor = 28.2'F Intermediate .5028 .8081 0 0 .6530 Shell 1.105 1.0279 0 0 1.0566 1.864 1.1706 0 0 1.3703 3.746 1.3418 60 80.5 1.8004 Sum: 80.5 4.8803 Chemistry Factor = 16.5'F Weld Metal .5028 .8081 149.7 121.0 .6530 1.105 1.0279 176.4 181.3 1.0566 1.864 1.1706 160.4 187.8 1.3703 3.746 1.3418 219.1 294.0 1.8004 Sum: 854.69 4.8803 Chemistry Factor = 160.7'F NOTES:

(a) Reference 1.

(b) ~RT>>~ for weld material is the adjusted value using the 1.069 ratioing factor per Reference 1 applied to the measured values of Table 2.

PTLR 10 Revision 2

TABLE 4 Reactor Vessel Toughness Table (Unirradiated)"

Naterial Description Cu (%) Ni (%) Initial RT>>('F)

Intermediate Shell .07 .69 20 Lower Shell .05 .69 40 Circumferential Weld .25 .56 -4.8 (a) Per Reference l.

TABLE 5 Reactor Vessel Surface Fluence Values at 19.5 and 32 EFPY" x 10" (n/cm', E ) 1.0 ~ev)

EFPY 0o 15'.47 30'.05 45'969 19.5 2.32 3.49 2.20 1.56

'.45 32 (a) Reference l.

PTLR Revision 2

TABLE 6 Calculation of Adjusted Reference Temperatures at 24 EFPY for the Limiting Reactor Vessel Material Parameter Values Operating Time 24 EFPY Material Circ. Weld Circ. Weld Location 1/4-T 3/4-T Chemistry Factor (CF), 160.7 160.7 (f), 10" (E > 1.0 HeV)" 1.85 .851 F"'luence n/cm Fluence Fact'or FF 1.17 .955 hRTgpy CF x FFy F 188 153,4 Initial RTgpy (I) F -4.8 -4.8 Margin (H), 'F" 48.3 48.3 ART = I + (CFxFF) + H F"" 232 196.9 NOTES:

(a) Value calculated using Table 5 values.

(b) Values from Table 3.

(c) Reference 1.

PTLR 12 Revision 2

Attachment V Redlined Version of LTOP Methodology identifies changes to methodology originally provided in December 8, 1995 RG&E letter to NRC)

LOW TEMPERATURE OVERPRESSURE PROTECTION SYSTEM (LTOPS)

INTRODUCTION The purpose of the LTOPS is to supplement the normal plant operational administrative controls to protect the reactor vessel from being exposed to conditions of fast propagating brittle fracture. The LTOPS also protects the Residual Heat Removal (RHR) System from overpressurizatlon. This has been achieved by conservatively choosing an LTOPS setpolnt which prevents the RCS from exceeding the pressure/temperature limits established by 10 CFR Part 50 Appendix G"'equirements, and the RHR System from exceeding 110% of its design pressure. The LTOPS is designed to provide the capability, during relatively low temperature operation (typically less than 350'F), to automatically prevent the RCS pressure from exceeding the applicable limits. Once the system is enabled, no operator action Is involved for the LTOPS to perform its Intended pressure mitigation function. Thus, no operator action is modelled in the analyses supporting the setpofnt selection, although operator action may be initiated to ultimately terminate the cause of the overpressure event.

The PORVs located near the top of the pressurizer, together with additional actuation logic from the low-range pressure channels, are utilized to mitigate potential RCS overpressure transients.

The LTOPS provides the relief capacity for specific transients which would not be mitigated by the RHR System relief valve. In addition, a limit on the PORV piping is accommodated due to the potential for water hammer effects to be developed in the piping associated with these valves as a result of the cyclic opening and closing characteristics during mitigation of an G<'>

overpressure transient. Thus, a pressure limitmore restrictive than the 10CFR50, Appendix allowable is imposed above a certain temperature so that the loads on the piping from a LTOPS event would not affect the piping integrity.

3-1

acr %s. I N

Two specific transients have been defined, with the RCS in a water-solid condition, as the design basis for LTOPS. Each of these scenarios assumes no RHR System heat removal capability. The RHR System relief valve (203) does not actuate during the transients. The first transient consists of a heat injection scenario in which a reactor coolant pump in a single loop is started with the RCS temperature as much as 50'F lower than the steam generator secondary side temperature. This results in a sudden heat input to a water-solid RCS from the steam generators, creating an increasing pressure transient. The second transient has been defined as a mass injection scenario into a water-solid RCS as caused by one of two possible scenarios. The first scenario is an inadvertent actuation of the safety injection pumps into the RCS. The second scenario is the simultaneous isolation of the RHR System, isolation of letdown, and failure of the normal charging flow controls to the full flow condition. Either scenario may be eliminated from consideration depending on the plant configurations which are restricted by technical specifications. Also, various combinations of charging and safety injection flows may also be evaluated on a plant-specific basis. The resulting mass injection/letdown mismatch causes an increasing pressure transient.

3.2 LTOPS Setpoint Determination Rochester Gas and Electric and Babcock 8 Wilcox Nuclear Technology (BWNT) have developed the following methodology which is employed to determine PORV setpolnts for mitigation of the LTOPS design basis cold overpressurization transients. This methodology maximizes the available operating margin for setpolnt selection while maintaining an appropriate level of protection in support of reactor vessel and RHR System integrity.

3-2

Parameters Considered The selection of proper LTOPS setpoint for actuating the PORVs requires the consideration of numerous system parameters including:

a. Volume of reactor coolant involved In transient
b. RCS pressure signal transmission delay
c. Volumetric capacity of the relief valves versus opening position, including the potential for critical flow
d. Stroke time of the relief valves (open 6 close)
e. Initial temperature and pressure of the RCS and steam generator
f. Mass input rate into RCS
g. Temperature of injected fluid
h. Heat transfer characteristics of the steam generators
i. Initial temperature asymmetry between RCS and steam generator secondary water
j. Mass of steam generator secondary water
k. RCP startup dynamics I. 10CFR50, Appendix 6"I pressure/temperature characteristics of the reactor vessel
m. Pressurizer PORV piping/structural analysis limitations
n. Dynamic and static pressure differences throughout the RCS and RHRS
o. RHR System pressure limits
p. Loop asymmetry for RCP start cases
q. Instrument uncertainty for temperature (conditions under which the LTOP System is placed into service) and pressure uncertainty (actuation setpoint)

These parameters are modelled in the BWNT RELAP5/MOD2-B&Wcomputer code (Ref. 19) 3-3

which calculates the maximum and minimum system pressures.

Pressure Limits Selection The function of the LTOPS is to protect the reactor vessel from fast propagating brittle fracture.

This has been implemented by choosing a LTOPS setpolnt which prevents exceeding the limits prescribed by the applicable pressure/temperature characteristic for the specific reactor vessel material in accordance with rules given in Appendix G to 10CFR50I". The LTOPS design basis takes credit for the fact that overpressure events most likely occur during isothermal conditions in the RCS. Therefore, it is appropriate to utilize the steady-state Appendix G limit. In addition, the LTOPS also provides for an operational consideration to maintain the integrity of the PORV piping, and to protect the RHR System from overpressure during the LTOPS design basis transients. A typical characteristic 10CFR50 Appendix G cuwe is shown by Figure 3.1 where the allowable system pressure increases with Increasing temperature. This type of curve sets the nominal upper limit on the pressure which should not be exceeded during RCS increasing pressure transients based on reactor vessel material properties. Superimposed on this curve ls the PORV piping limit and RHR System pressure limit which is conservatively used, for setpolnt development, as the maximum allowable pressure above the temperature at which it intersects with the 10CFR50 Appendix G curve.

When a relief valve is actuated to mitigate an increasing pressure transient, the release of a volume of coolant through the valve will cause the pressure increase to be slowed and reversed as described by Figure 3.2. The system pressure then decreases, as the relief valve releases coolant, until a reset pressure is reached where the valve is signalled to close. Note that the pressure continues to decrease below the reset pressure as the valve recloses. The nominal 3-4

II 1

<,fikt'~,t+g+

s

lower limit on the pressure during the transient ls typically established based solely on an operational consideration for the reactor coolant pump P1 seal to maintain a nominal differential pressure across the seal faces for proper film-riding performance. In the event that the available range is insufficient to concurrently accommodate the upper and lower pressure limits, the upper pressure limits are given preference.

The nominal upper limit (based on the minimum of the steady-state 10CFR50 Appendix G requirement, the RHR System pressure limit, and the PORV piping limitations) and the nominal RCP 41 seal performance criteria create a pressure range from which the setpoints for both PORVs may be selected as shown on Figures 3.3 and 3.4. Where there is insufficient range between the upper and lower pressure limits to select PORV setpoints to provide protection against violation of both limits, setpoint selection to provide protection against the upper pressure limit violation shall take precedence.

Mass Input Consideration For a particular mass input transient to the RCS, the relief valve will be signalled to open at a specific pressure setpoint. However, as shown on Figure 3.2, there will be a pressure overshoot during the delay time before the valve starts to move and during the time the valve is moving to the full open position. This overshoot is dependent on the dynamics of the system and the input parameters, and results in a maximum system pressure somewhat higher than the set pressure. Similarly there will be a pressure undershoot, while the valve is relieving, both due to the reset pressure being below the setpoint and to the delay in stroking the valve closed.

The maximum and minimum pressures reached (P>>>< and PQiN) in the transient are a function of the selected setpoint (P,) as shown on Figure 3.3. The shaded area represents an optimum 3-5

range from which to select the setpoint based on the particular mass input case. Several mass input cases may be run at various input flow rates to bound the allowable setpoint range.

Heat Input Consideration The heat input case is done similarly to the mass input case except that the locus of transient pressure values versus selected setpoints may be determined for several values of the initial RCS temperature. This heat input evaluation provides a range of acceptable setpoints dependent on the reactor coolant temperature, whereas the mass input case is limited to the most restrictive low temperature condition only (i.e. the mass injection transient is not sensitive to temperature). The shaded area on Figure 3.4 describes the acceptable band for a heat input transient from which to select the setpoint for a particular initial reactor coolant temperature.

If the LTOPS is a single setpolnt system, the most limiting result Is used throughout.

Final Setpoint Selection By superimposing the results of multiple mass input and heat input cases evaluated, (from a series of figures such as 3.3 and 3.4) a range of allowable PORV setpoints to satisfy both

/

conditions can be determined. For a single setpoint system, the most limiting setpoint is chosen, with the upper pressure limit given precedence if both limits cannot be accommodated.

The selection of the setpolnts for the PORVs considers the use of nominal upper and lower pressure limits. The upper limits are specified by the minimum of the steady-state cooldown curve as calculated in accordance with Appendix 8 to 10CFR50I'I or the peak RCS or RHR 3-6

I g

System pressure based upon piping/structural analysis loads. The lower pressure extreme is specified by the reactor coolant pump P1 seal minimum differential pressure performance criteria. Uncertainties in the pressure and temperature instrumentation utilized by the LTOPS are accounted for consistent with the methodology of Reference 2.0. Accounting for the effects'f instrumentation uncertainty imposes additional restrictions on the setpoint development, which is already based on conservative pressure limits such as a safety factor of 2 on pressure stress, use of a lower bound Kfcurve and an assumed ~IT flaw depth with a length equal to 1~8 times the vessel wall thicknes 3.3 Application of ASME Code Case N-514 Ere e d:'8::".I't6'L'id:,tran rt -: I!1I! -:l...,,!r.':i",,:e-::

tc;t't,OW~Ot'the:::Preeeureq deter~1ned<t~aSStf+SPPendec;8" ~"~-allewe

, paragraph G-2215, of section xt of the AsME code"t.QYt~te,:spp1RVgfog@fASME"::Code!Casa'N.".:.St'8'":lnclsaeae:::the.

JOJeletfttg,::,ntarglh)1A:::tl'l8~fSQIOtl~!OfifltstprBssule-tJ88lperatutst!Ilnttt;".,Oulpseirrh~WIK~,

hsi'L!tCp'Sile'nagfedercods,:case;N-".ste:.requfreet Lfg%~!o:bs;:effecthretst coolantaetnpelatureeffeesdfen

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'RooeF.:::.Orgg~epgant tefnP crater ee;OOrree goading Ltd,a::. reaoforrtr Seeel~mitalp tetnPetaiure;:."::-:.Ot! 8 I dfetenoe rroln ethfnefdtr.:,trees e~t'Suds ceil'ee~ahteniftTtetr't+~80%F~

whichever is greater. RTNpT is the highest adjusted reference temperature for weld or base 3-7

metal in the beltline region at a distance one-fourth of the vessel section thickness from the vessel inside surface, as determined by Regulatory Guide 1.99, Revision 2.

3.4 Enable Temperature for LTOPS The enable temperature is the temperature below which the LTOPS system is required to be operablei bTrhe:Sfn~na L70:3:egabfeltsntpeinture le,,eetabliihed.uefnng::Ihe:ff fdinne;prOV¹d:::byASliilegtf Cede.'.Case,,NS O',::;:Fhe,A'8MB!Code!CsiY%.',<,'.:.i6~@ris.'en'::"en+N(RCF':,.qu,.d~teBpeYa~FN nnrreegnndfnffi'O~ihetreantnrbTee'See!!ltrrii:eetil!i'eiiiPiiiiiiire!r'll'RT: sj,"LSgsePNiggtfeP,';-The e&QeaeWT Whinheeer iS greater aS deeCrtbed In SeCtiOn 3.3e!Tliialdaffnlt7nn"..I'SYafenr!euPPOited!~[i!then titreebngbouestgwneds6roup~ihsafnnaTenabfe'ternpeinture federerrnfned~as(IITianr+807paf; 3-8

The RCS cold leg temperature limitation for starting an RCP is the same value as the LTOPS enable temperature to ensure that the basis of the heat injection transient is not violated. The Standard Technical Specifications (STS) prohibit starting an RCP when any RCS cold leg temperatures is less than or equal to the LTOPS enable temperature unless the secondary side water temperature of each steam generator is less than or equal to 50'F above each of the RCS cold leg temperatures.

3-9

Figure 3.1 TYPICALAPPENDIX G P/T CHARACTERISTICS (g 2500

~~ 2000 I z

~O 0

O 1500 0 'FNR U 1000 IMPOSED PORV PIPING LIMIT I-9Cl 500 100 IMPOSED RHRS z PIPING LIMIT 0

0 100 200 300 400 500 lNDICATEDCOOLANT TEMPERATURE, 'F 3-10

Figure 3.2 TYR ICAL: RRESSUR 2'TRANSIENT

.:(1.'REL'IEF VAVLECYCLE):.

8EVPOINT-------------

RESET

~ Uride 3-11

Figure 3.3

""SAP'03N3':: >>':

DET.ERMIINATIQN

'(MASS INPUT):

AVP

'APPENDIX:G SIAXIMUMt;IMIT'

MAX

,'CP& SEAL':::

PERFORMANCE CRITERIA;;.;;;

SETPOINT RANGE PORV SETPOIN7):PSlG the PORV discharge The maximum pressure limit is the rginimum of the Appendix G limit, piping structural analysis limit, or the RHR system limit 3-12

Figure 3.4

'(HEAT:INPUT) -"

'APPENDIX:G SIAXIMUMt;IMIT'.

Pex--------

I I

I I

RCP A: SEAL

PERFORMANCE CR1TERlA SETPOINT RANGE:

PORV SETPOINT):PSIG The maximum pressure limit Is the minimum of the Appendix G limit, the PORV discharge piping structural analysis limit, or the RHR system limit 3-13

4.0 REFERENCES

NUREG 1431, "Standard Technical Specifications for Westinghouse Pressurized Water Reactors",

Revision 0, September, 1992.

2. U.S. Nuclear Regulatory Commission, "Removal of Cycle-Specific Parameter Limits from Technical Specifications", Generic Letter 88-16, October, 1988.
3. U.S. Nuclear Regulatory Commission, Radiation Embrittlement of Reactor Vessel Materials, Re ulato Gufde1.99 Revislon2, May,1988.
4. Code of Federal Regulations, Title 10, Part 50, "Fracture Toughness Requirements for LIght-Water Nuclear Power Reactors", Appendix G, Fracture Toughness Requirements.
5. ASME Boiler and Pressure Vessel Code,Section XI, "Rules for Inservlce Inspection of Nuclear Power Plant Components", Appendix G, Fracture Toughness Criteria For Protection Against Failure.
6. R. G. Soltesz, R. K. Disney, J. Jedruch, and S. L Ziegler, Nuclear Rocket Shielding Methods, Modification, Updating and Input Data Preparation. Vol. 5-Two-Dimensional Discrete Ordinates Transport Technique, WANL-PR(LL)434, Vol. 5, August 1970.
7. ORNL RSIC Data LIbrary Collection DLC-76 SAILOR Coupled Self-Shielded, 47 Neutron, 20 Gamma-Ray, P3, Cross Section Library for Light Water Reactors.

ASME Boiler and Pressure Vessel Code,Section III, "Rules for Construction of Nuclear Power Plant Components", Division 1, Subsection NB: Class 1 Components.

Branch Technical Position MTEB 5-2, "Fracture Toughness Requirements", NUREG4800 Standard Review Plan 5.3.2, Pressure-Temperature Limits, July 1981, Rev. 1.

10. ASTM E-208, Standard Test Method for Conducting Drop-Weight Test to Determine Nil-Ductility Transition Temperature of Ferritic Steels, ASTM Standards, Section 3, American Society for Testing and Materials.
11. B8W Owners Group Report BAW-2202, "Fracture Toughness Characterization of WF-70 Weld 4-1

Material", BBW Owners Group Materials Committee, September 1993.

12. Letter, Clyde Y. Shiraki, Nuclear Regulatory Commission, to D. L Farrar, Commonwealth Edison-Company, 'Exemption from the Requirement to Determine the Unirradiated Reference Temperature in Accordance with the Method Specified In 10 CFR 50.61(b) (2) (i) (TAC NOS. M84546 and M84547), Docket Nos. 50-295 and 50404, February 22, 1994.
13. Code of Federal Regulations, Title 10, Part 50, "Fracture Toughness Requirements for Light-Water Nuclear Power Reactors, Appendix H, Reactor Vessel Material Surveillance Program Requirements.
14. Timoshenko, S. P. and Goodier, J. N., Theo of Elasticit, Third Edition, McGraw-Hill Book Co.,

New York, 1970.

15. ASME Boiler and Pressure Vessel Code,Section XI, "Rules for Inservice Inspection of Nuclear Power Plant Components", Appendix A, Analysis of Flaws, Article A@000, Method For K, Determination.
16. WRC Bulletin No. 175, PVRC Recommendations on Toughness Requirements for Ferritic Materials",

Welding Research Council, New York, August 1972.

17. ASME Boiler and Pressure Vessel Code Case N-514,Section XI, Division 1, "Low Temperature Overpressure Protection", Approval date: February 12, 1992.
18. Branch Technical Position RSB 5-2, "Overpressurization Protection of Pressurized Water Reactors While Operating at Low Temperatures", NUREG4800 Standard Review Plan 5.2.2, Overpressure Protection, November 1988, Rev. 2.
19. BWNT, "RELAPS/MOD2, An Advanced Computer Program for Light-Water Reactor LOCA and Non-LOCA Transient Analysis," BAW-1 0164P-A.
20. Instrument of America (ISA) Standard 67.04-1994.

4-2

Attachment VI Final Version of LTOP Methodology (Replaces methodology originally provided in December 8, 1995 RG&E letter to NRC which in turn replaced methodology provided in Section 3 to WCAP-14040)

LOW TEMPERATURE OVERPRESSURE PROTECTION SYSTEM (LTOPS)

INTRODUCTION The purpose of the LTOPS Is to supplement the normal plant operational administrative controls to protect the reactor vessel from being exposed to conditions of fast propagating brittle fracture. The LTOPS also protects the Residual Heat Removal (RHR) System from overpressurization. This has been achieved by conservatively choosing an LTOPS setpoint which prevents the RCS from exceeding the pressure/temperature limits established by10 CFR Part 50 Appendix GI'I requirements, and the RHR System from exceeding 110% of its design pressure. The LTOPS is designed to provide the capability, during relatively low temperature operation (typically less than 350'F), to automatically prevent the RCS pressure from exceeding the applicable limits. Once the system is enabled, no operator action is involved for the LTOPS to perform its intended pressure mitigation function. Thus, no operator action is modelled in the analyses supporting the setpoint selection, although operator action may be initiated to ultimately terminate the cause of the overpressure event.

The PORVs located near the top of the pressurizer, together with additional actuation logic from the low-range pressure channels, are utilized to mitigate potential RCS overpressure transients.

The LTOPS provides the relief capacity for specific transients which would not be mitigated by the RHR System relief valve. In addition, a limit on the PORV piping is accommodated due to the potential for water hammer effects to be developed in the piping associated with these valves as a result of the cyclic opening and closing characteristics during mitigation of an overpressure transient. Thus, a pressure limitmore restrictive than the 10CFR50, Appendix GI'I allowable is imposed above a certain temperature so that the loads on the piping from a LTOPS event would not affect the piping integrity.

3-1

0 II iE

'I I

Two specific transients have been defined, with the RCS in a water-solid condition, as the design basis for LTOPS. Each of these scenarios assumes no RHR System heat removal capability. The RHR System relief valve (203) does not actuate during the transients. The first transient consists of a heat injection scenario in which a reactor coolant pump in a single loop is started with the RCS temperature as much as 50'F lower than the steam generator secondary side temperature. This results in a sudden heat input to a water-solid RCS from the steam generators, creating an increasing pressure transient. The second transient has been defined as a mass injection scenario into a water-solid RCS as caused by one of two possible scenarios. The first scenario is an inadvertent actuation of the safety injection pumps into the RCS. The second scenario is the simultaneous Isolation of the RHR System, isolation of letdown, and failure of the normal charging flow controls to the full flow condition. Either scenario may be eliminated from consideration depending on the plant configurations which are restricted by technical specifications. Also, various combinations of charging and safety injection flows may also be evaluated on a plant-specific basis. The resulting mass injection/letdown mismatch causes an increasing pressure transient.

3.2 LTOPS Setpoint Determination Rochester Gas and Electric and Babcock & Wilcox Nuclear Technology (BWNT) have developed the following methodology which is employed to determine PORV setpoints for mitigation of the LTOPS design basis cold overpressurization transients. This methodology maximizes the available operating margin for setpoint selection while maintaining an appropriate level of protection in support of reactor vessel and RHR System integrity.

3-2

Parameters Considered The selection of proper LTOPS setpoint for actuating the PORVs requires the consideration of numerous system parameters including:

a. Volume of reactor coolant involved in transient
b. RCS pressure signal transmission delay
c. Volumetric capacity of the relief valves versus opening position, including the potential for critical flow
d. Stroke time of the relief valves (open & close)
e. Initial temperature and pressure of the RCS and steam generator
f. Mass input rate into RCS
g. Temperature of injected fluid
h. Heat transfer characteristics of the steam generators
i. Initial temperature asymmetry between RCS and steam generator secondary water 1

J. Mass of steam generator secondary water

k. RCP startup dynamics I. 10CFR50, Appendix Gt'I pressure/temperature characteristics of the reactor vessel
m. Pressurizer PORV piping/structural analysis limitations
n. Dynamic and static pressure differences throughout the RCS and RHRS
o. RHR System pressure limits
p. Loop asymmetry for RCP start cases
q. Instrument uncertainty for temperature (conditions under which the LTOP System is placed into service) and pressure uncertainty (actuation setpolnt)

These parameters are modelled in the BWNT RELAP5/MOD2-B&Wcomputer code (Ref. 19) 3-3

Sr';

which calculates the maximum and minimum system pressures.

Pressure Limits Selection The function of the LTOPS is to protect the reactor vessel from fast propagating brittle fracture.

This has been implemented by choosing a LTOPS setpolnt which prevents exceeding the limits prescribed by the applicable pressure/temperature characteristic for the specific reactor vessel material in accordance with rules given in Appendix G to 10CFR50I". The LTOPS design basis takes credit for the fact that overpressure events most likely occur during isothermal conditions in the RCS. Therefore, it is appropriate to utilize the steady-state Appendix G limit. In addition, the LTOPS also provides for an operational consideration to maintain the integrity of the PORV piping, and to protect the RHR System from overpressure during the LTOPS design basis transients. A typical characteristic 10CFR50 Appendix G curve is shown by Figure 3.1 where the allowable system pressure increases with increasing temperature. This type of curve sets the nominal upper limit on the pressure which should not be exceeded during RCS increasing pressure transients based on reactor vessel material properties. Superimposed on this curve is the PORV piping limit and RHR System pressure limit which is conservatively used, for setpoint development, as the maximum allowable pressure above the temperature at which it intersects with the 10CFR50 Appendix G curve.

When a relief valve is actuated to mitigate an increasing pressure transient, the release of a volume of coolant through the valve will cause the pressure increase to be slowed and reversed as described by Figure 3.2. The system pressure then decreases, as the relief valve releases coolant, until a reset pressure is reached where the valve is signalled to close. Note that the pressure continues to decrease below the reset pressure as the valve recloses. The nominal 3-4

Q6 p~

I I 1

tv~'=<<fj

lower limit on the pressure during the transient is typically established based solely on an operational consideration for the reactor coolant pump ¹1 seal to maintain a nominal differential pressure across the seal faces for proper film-riding performance. In the event that the available range is insufficient to concurrently accommodate the upper and lower pressure limits, the upper pressure limits are given preference.

The nominal upper limit (based on the minimum of the steady-state 10CFR50 Appendix 8 requirement, the RHR System pressure limit, and the PORV piping limitations) and the nominal RCP ¹1 seal performance criteria create a pressure range from which the setpoints for both PORVs may be selected as shown on Figures 3.3 and 3.4. Where there is insufficient range between the upper and lower pressure limits to select PORV setpolnts to provide protection against violation of both limits, setpolnt selection to provide protection against the upper pressure limit violation shall take precedence.

Mass Input Consideration For a particular mass input transient to the RCS, the relief valve will be signalled to open at a specific pressure setpolnt. However, as shown on Figure 3.2, there will be a pressure overshoot during the delay time before the valve starts to move and during the time the valve is moving to the full open position. This overshoot is dependent on the dynamics of the system and the input parameters, and results in a maximum system pressure somewhat higher than the set pressure. Similarly there will be a pressure undershoot, while the valve is relieving, both due to the reset pressure being below the setpoint and to the delay in stroking the valve closed.

The maximum and minimum pressures reached (P>>and P~,) in the transient are a function of the selected setpoint (Ps) as shown on Figure 3.3. The shaded area represents an optimum 3-5

range from which to select the setpoint based on the particular mass input case. Several mass Input cases may be run at various input flow rates to bound the allowable setpoint range.

Heat Input Consideration The heat input case is done similarly to the mass input case except that the locus of transient pressure values versus selected setpoints may be determined for several values of the initial RCS temperature. This heat input evaluation provides a range of acceptable setpolnts dependent on the reactor coolant temperature, whereas the mass input case is limited to the most restrictive low temperature condition only (i.e. the mass injection transient is not sensitive to temperature). The shaded area on Figure 3.4 describes the acceptable band for a heat input transient from which to select the setpoint for a particular initial reactor coolant temperature.

If the LTOPS is a single setpolnt system, the most limiting result is used throughout.

Final Setpoint Selection By superimposing the results of multiple mass input and heat input cases evaluated, (from a series of figures such as 3.3 and 3.4) a range of allowable PORV setpoints to satisfy both conditions can be determined. For a single setpoint system, the most limiting setpoint is chosen, with the upper pressure limit given precedence if both limits cannot be accommodated.

The selection of the setpolnts for the PORVs considers the use of nominal upper and lower pressure limits. The upper limits are specified by the minimum of the steady-state cooldown curve as calculated in accordance with Appendix G to 10CFR50'I or the peak RCS or RHR 3-6

System pressure based upon piping/structural analysis loads. The lower pressure extreme is specified by the reactor coolant pump 41 seal minimum differential pressure performance criteria. Uncertainties in the pressure and temperature instrumentation utilized by the LTOPS are accounted for consistent with the methodology of Reference 2.0. Accounting for the effects of instrumentation uncertainty imposes additional restrictions on the setpolnt development, N

which is already based on conservative pressure limits such as a safety factor of 2 on pressure stress, use of a lower bound K R curve and an assumed ~/~T flaw depth with a length equal to 1~8 times the vessel wall thickness.

3.3 Application of ASME Code Case N-514 ASME Code Case N-514I' allows LTOPS to limit the maximum pressure in the reactor vessel to 110% of the pressure determined to satisfy Appendix G, paragraph G-2215, of Section XI of i the ASME Code"'. The application of ASME Code Case N-514 increases the operating margin in the region of the pressure-temperature limit curves where the LTOPS is enabled. Code Case N-514 requires LTOPS to be effective at coolant temperatures less than 200'F or at coolant temperatures corresponding to a reactor vessel metal temperature, at a 1/4t distance from the inside vessel surface, less than Ropy + 50 F, whichever is greater. RTD~ is the highest adjusted reference temperature for weld or base metal in the beltline region at a distance one-fourth of the vessel section thickness from the vessel Inside surface, as determined by Regulatory Guide 1.99, Revision 2.

3-7

Enable Temperature for LTOPS The enable temperature is the temperature below which the LTOPS system is required to be operable.

The Glnna LTOPS enable temperature is established using the guidance provided by ASME XI Code Case N-514. The ASME Code Case N-514 supports an enable RCS liquid temperature corresponding to the reactor vessel 1/4t metal temperature of RTNp~ + 50 F or 200'F, whichever is greater as described in Section 3.3. This definition ls also supported by the Westinghouse Owner's Group. The Ginna enable temperature is determined as (RTNpY + 50 F)

+ (instrument error I~I) + (metal temperature difference to 1/4 T).

The RCS cold leg temperature limitation for starting an RCP is the same value as the LTOPS enable temperature to ensure that the basis of the heat injection transient is not violated. The Standard Technical Specifications (STS) prohibit starting an RCP when any RCS cold leg temperatures is less than or equal to the LTOPS enable temperature unless the secondary side water temperature of each steam generator is less than or equal.to 50'F above each of the RCS cold leg temperatures.

3-8

Figure 3.1 TYPICALAPPENDIX G P/T CHARACTERISTICS (g 2500 2000

~

z.

~1500

~

0 O

EL' oF/HR

~U 1000 IMPOSED PORV Cl PIPING LIMIT I-Q 500 100 IMPOSED RHRS Cl PIPING LIMIT 0

0 100 200 300 400 500 I NDIGATED COOLANT TEMPERATURE,

'F 3-9

P Figure 3.2 TYR ICAL'RESSURE:TRANSIENT

"(1'; R EL'I EF,',VAVLE CYCLE):;",":

RESE7 3-10

Figure 3.3

, '. SETPO)NT::.:":

DET.ERMIINATION:

"(MASS INPUT):

'APPENDIX'G MAXIMUM l.'IMIT'CP

& 'SEA'L':::

PERFORMANCE

'CRrrE8%;:::;:

SETPOINT RANGE:

PORV SETPOINT):PSIG The maximum pressure limit is the minimum of the Appendix G limit, the PORV discharge piping structural analysis limit, or the'RHh system limit 3-11

Figure 3.4

-.; SE FPQ) NT::

DETER MIIMATION:

(HEAT:INP. UT)

'APPENDIX:G MAXIMUM I.'IMIT'-------------

P ue--------

P L)Ã I

I RCR N: SEAL::;

PE%'.QRMANCE

'CRrrERtA::::::

SETPOINT. RANGE:

p.

S P,ORV SETPOIN7):PSlG The maximum pressure limit is the minimum of the Appendix G limit, the PORV discharge piping structural analysis limit, or the RHR system limit 3-12

NUREG 1431, "Standard Technical Specifications for Westinghouse Pressurized Water Reactors",

Revision 0, September, 1992.

2. U.S. Nuclear Regulatory Commission, "Removal of Cycle-Specific Parameter Limits from Technical Specifications", Generic Letter 88-16, October, 1988.
3. U.S. Nuclear Regulatory Commission, Radiation Embrittlement of Reactor Vessel Materials, Re ulato Guide 1.99 Revision 2, May, 1988.
4. Code of Federal Regulations, Title 10, Part 50, "Fracture Toughness Requirements for Light-Water Nuclear Power Reactors", Appendix G, Fracture Toughness Requirements.

ASME Boiler and Pressure Vessel Code Section XI, 'Rules for Inservice Inspection of Nuclear Power Plant Components", Appendix G, Fracture Toughness Criteria For Protection Against Failure.

6. R. G. Soltesz, R. K. Disney, J. Jedruch, and S. I Ziegier, Nuclear Rocket Shielding Methods, Modification, Updating and Input Data Preparation. Vol. 5-Two-Dimensional Discrete Ordinates Transport Technique, WANL-PR(LL)<34, Vol. 5, August 1970.

ORNL RSIC Data LIbrary Collection DLC-76 SAILOR Coupled Self-Shielded, 47 Neutron, 20 Gamma-Ray, P3, Cross Section Library for Light Water Reactors.

ASME Boiler and Pressure Vessel Code,Section III, "Rules for Construction of Nuclear Power Plant Components", Division 1, Subsection NB: Class 1 Components.

Branch Technical Position MTEB 5-2, "Fracture Toughness Requirements", NUREG4800 Standard Review Plan 5.3.2, Pressure-Temperature Limits, July 1981, Rev. 1.

10. ASTM E-208, Standard Test Method for Conducting Drop-Weight Test to Determine Nil-Ductility Transition Temperature of Ferritic Steels, ASTM Standards, Section 3, American Society for Testing and Materials.
11. B&W Owners Group Report BAW-2202, "Fracture Toughness Characterization'of WF-70 Weld Material", B&W Owners Group Materials Committee, September 1993.

4-1

u. Letter, Clyde Y. Shlraki, Nuclear Regulatory Commission, to D. L. Farrar, Commonwealth Edison Company, "Exemption from the Requirement to Determine the Unirradiated Reference Temperature in Accordance with the Method Specified in 10 CFR 50.61(b) (2) (i) (TAC NOS. M84546 and M84547)", Docket Nos. 50-295 and 50404, February 22, 1994.
13. Code of Federal Regulations, Title 10, Part 50, "Fracture Toughness Requirements for Light-Water Nuclear Power Reactors", Appendix H, Reactor Vessel Material Surveillance Program Requirements.
14. Tlmoshenko, S. P. and Goodier, J. N., Theo of Elastlcit, Third Edition, McGraw-Hill Book Co.,

New York, 1970.

15. ASME Boiler and Pressure Vessel Code,Section XI, "Rules for Inservice Inspection of Nuclear Power Plant Components", Appendix A, Analysis of Flaws, Article A-3000, Method For g Determination.
16. WRC Bulletin No. 175, "PVRC Recommendations on Toughness Requirements for Ferritlc Materials",

Welding Research Council, New York, August 1972.

17. ASME Boiler and Pressure Vessel Code Case N-514,Section XI, Division 1, "Low Temperature Overpressure Protection", Approval date: February 12, 1992.
18. Branch Technical Position RSB 5-2, "Overpressurization Protection of Pressurized Water Reactors While Operating at Low Temperatures", NUREG4800 Standard Review Plan 5.2.2, Overpressure Protection, November 1988, Rev. 2.
19. BWNT, "RELAPS/MOD2, An Advanced Computer Program for Light-Water Reactor LOCA and Non-LOCA Transient Analysis," BAW-10164P-A.
20. Instrument of America (ISA) Standard 67.04-1994.

4-2

Attachment VII LTOP Enable Temperature Calculation 1

(First use of LTOP enable temperature methodology)

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