ML20078S419

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Proposed Tech Specs Section 3.4.9.3. & Associated Bases
ML20078S419
Person / Time
Site: Braidwood  Constellation icon.png
Issue date: 12/16/1994
From:
COMMONWEALTH EDISON CO.
To:
Shared Package
ML20078S413 List:
References
NUDOCS 9412280295
Download: ML20078S419 (18)
v • d • e


Text

.

ATTACHMENT B MARKED UP PAGES FOR PROPOSED CHANGES TO APPENDIX A TECHNICAL SPECIFICATIONS OF FACILITY OPERATING LICENSES NPF-72 AND NPF-77 BRAIDWOOD STATION UNIT 1 REVISED PAGES:

VIII 3/4 4-39*

3/4 4-40 B 3/4 4-16

  • NOTE: THIS PAGE HAS NO CHANGES BUT IS INCLUDED FOR CONTINUITY.

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k nla\twi\lteprev\l2 12 9412280293 941216 PDR ADOCK 05000456 P PDR

. LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION f_AfL TABLE 3.4-2 REACTOR COOLANT SYSTEM CHEMISTRY LIMITS............... 3/44-25 TABLE 4.4-3 REACTOR COOLANT SYSTEM CHEMISTRY SURVEILLANCE j REQUIREMENTS........................................ 3/44-26 1 3/4.4.8 SPECIFIC ACTIVITY........................................ 3/44-27 FIGURE 3.4-1 DOSE EQUIVALENT I-131 REACTOR COOLANT SPECIFIC l l

ACTIVITY LIMIT VERSUS PERCENT OF RATED THERMAL POWER WITH THE REACTOR COOLANT SPECIFIC ACTIVITY

>l *Ci/ GRAM DOSE EQUIVALENT I-131.................. 3/44-29 TABLE 4.4-4 REACTOR COOLANT SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PR0 GRAM.................................... 3/4 4-30 3/4.4.9 PRESSURE / TEMPERATURE LIMITS Re ac tor Cool an t Sys tem. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/44-32 FIGURE 3.4-2a REACTOR COOLANT SYSTEM HEATUP LIMITATIONS APPLICABLE UP TO 32 EFPY (UNIT 1)...... 3/44-33

, FIGURE 3.4-2b REACTOR COOLANT SYSTEM HEATUP s LIMITATIONS APPLICABLE UP TO 16 EFPY (UNIT 2)...... 3/44-34 FIGURE 3.4-3a REACTOR COOLANT SYSTEM C00LDOWN LIMITATIONS APPLICABLE UP TO 32 EFPY (UNIT 1)...... 3/44-35 FIGURE 3.4-3b PEACTOR COOLANT SYSTEM C00LDOWN LIMITATIONS APPLICABLE UP TO 16 EFPY (UNIT 2)...... 3/44-36 TABLE 4.4-5 REACTOR VESSEL MATERIAL SURVEILLANCE PROGRAM - WITHDRAWAL SCHEDULE....................... 3/44-37 Pressurizer.............................................. 3/44-38 Overpressure Protection Systems.......................... 3/44-39 FIGURE 3.4-4a NOMINAL PORY PRESSURE RELIEF SETPOINT VERSUS g RCS TEMPERATURE FOR THE COLD OVERPRESSURE p,M .

PROTECTION SYSTEM APPLICABLE UP T0 '537 EFPY (Unit 1) 3/44-40 FIGURE 3.4-4b NOMINAL PORY PRESSURE RELIEF SETPOINT VERSUS RCS

/

TEMPERATURE FOR THE COLD OVERPRESSURE PROTECTION SYSTEM APPLICABLE UP TO 16 EFPY (UNIT 2).......... 3/4 4-40a ,

I 3/4.4.10 STRUCTURAL INTEGRITY..................................... 3/44-42 3/4.4.11 REACTOR COOLANT SYSTEM VENTS............................. 3/44-43

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a.

REACTOR-COOLANT SYSTEM

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,) OVEURESSVRE PROTECTION SYSTEMS L]ylTING CONDITION FOR OPERATION 3.4.9.3 At least two overpressure protection devices shall be OPERABLE, and each-device shall be either:

a. A residual heat removal (RKR) suction relief valve with a lift setting of less than or equal to 450 psig, or
b. A power operated relief valve (PORV) with a lift setpoint that varies with RCS temperature which does not exceed the limit established in Figure 3.4-4a for Unit 1 (Figure 3.4-4b for Unit 2).

APPLICABILITY:

MODES 4, 5, and 6 with the reactor vessel head on.

ACTION:

a. With one of the two required verpressure protection devices inoperable in MODE 4, restore two overpressure protection devices to OPERABLE status within 7 days or depressurize and vent the RCS through at least a 2 square inch vent within the next 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
b. With one of the two required overpressure protection devices inoperable in MODES 5 or 6, restore two overpressure protection devices to OPERABLE status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or vent the RCS through at least a 2 square inch vent within the next 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

I c. With both of the required overpressure protection devices inoperable, depressurize and vent the RCS through at least a 2 square inch vent within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

d. With the RCS vented per ACTIONS a, b, or c, verify the vent pathway at least once per 31 days when the pathway is provided by a valve (s) that is locked, sealed, or otherwise secured in the open position; otherwise, verify the vent pathway every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />,
e. In the event either the PORVs, RHR suction relief valves, or the RCS vents are used to mitigate an RCS pressure transient, a Special Report shall be prepared and submitted to the Conmiission pursuant te Specification 6.9.2 within 30 days. The report shall describe the circumstances initiating the transient, the effect of the PORVs, Rlt suction relief valves, or RCS vents on the transient, and any corrective action necessary to prevent recurrence.
f. The provisions of Specification 3.0.4 are not applicable.

..."'t 3 /t 4-39 AMENDMENT NO. 53

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REACTOR COOLANT SYSTEM BASES PRESSURE / TEMPERATURE LIMITS (Continued) comparison of the steady-state and finite heatup rate data. At any given temperature, the allowable pressure is taken to be the lesser of the three values taken from the curves under consideration.

The use of the composite curve is necessary to set conservative heatup limitations because it is possible for conditions to exist such that over the course of the heatup ramp the controlling condition switches from the inside to the outside and the pressure limit must at all times be based on analysis of the most critical criterion.

Finally, the composite curves for the heatup rate data and the cooldown rate data are adjusted for possible errors in the pressure and temperature sensing instruments by the values indicated on the respective curves.

Although the pressurizer operates in temperature ranges above those for which there is reason for concern of nonductile failure, operating limits are provided to assure compatibility of operation with the fatigue analysis performed in accordance with the ASME Code requirements.

The OPERABILITY of two PORVs, or two RHR suction relief valves, or one -

PORV and one RHR suction relief valve, or an RCS vent opening of at least l 2 square inches ensures that the RCS will be protected from pressure transients t which could exceed the limits of Appendix G to 10 CFR Part 50 when one or more l of the RCS cold legs are less than or equal to 350*F. Either PORV has adequate relieving capability to protect the RCS from overpressurization when the tran-tient is~1imited to either: m the start nf an idla RCP vith the eac

  • e m water temperature of the st'eita' generator le'ss than or equal to 50*F above 'the

. prA cold leg ten eratures..or (2) +he start af a centri'uga.1.ch:rging pu p snd its injection into a water solid RCS.

These two scenarios are analyzed to determine the resulting overshoots assuming a single PORV actuation with a stroke time of 2.0 seconds from full closed to full open. Fi- 3.4-4 (3.4-45) :r: 5:::d ;;:n thi: rlysie :M

= rep-ee r te t M e rie"- eI: lr: $1e *^Pu vari ele retreint such thet, fer t M t m

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3/4.4.10 STRUCTURAL INTEGRITY b

The inservice inspection and testing programs for ASME Code Class 1, 2, and 3 components ensure that the structural integrity and operational readiness  ;

of these components will be maintained at an acceptable level throughout the life of the plant. These programs are in a:cordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR 50.55a(g) except where specific written relief has been granted by the Commission pursuant to 10 CFR 50.55a(g)(6)(i).

l B 3/4 4-16 AMENDMENT NO. 39, 33 l BRAIDWOOD - UNITS 1 & 2 1

P INSERT B Figure 3.4-4a and 3.4-4b are based on this analysis and represent the maximum allowable PORV variable setpoint such that for the overpressurization transients noted, Figure 3.4-4a limits the resulting pressure to 110% of the Appendix G vessel NDT limits,

' Figure 3.4-4b ensures the resulting pressure will not exceed the Appendix G vessel NDT limits.

Figure 3.4-4a does not'contain allowances.fcr' random instrument uncertainties and was developed using the guidance contained.in ASME Code Case N-514 in addition to the Appendix G requirements.

Figure 3.4 -4a also accounts for the flow induced pressure difference between the RCS pressure sensor and the Reactor Vessel beltline.

Figure 3.4-4b contains allowances for temperature Jnstrument uncertainties and does not account for the flow induced pressure difference between the RCS pressure sensor and the Reactor Vessel-beltline.

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ATTACHMENT C 1

EVALUATION OF SIGNIFICANT HAZARDS CONSIDERATIONS FOR PROPOSED CHANGES TO APPENDIX A TECHNICAL SPECIFICATIONS OF FACILITY OPERATING LICENSES NPF-72 AND NPF-77 Commonwealth Edison has evaluated this proposed amendment and determined that it involves no significant hazards considerations. According to Title 10 Code'of Federal Regulations Section 50 Subsection 92 Paragraph c (10 CFR 50.92 (c)), a. proposed amendment to an operating license involves no significant hazards considerations if operation of the facility in accordance with the proposed amendment would not:

1. Involve a significant. increase in the probability or consequences of an accident previously evaluated; or
2. Create the possibility of.a new.or different kind of accident from any accident previously evaluated; or
3. Involve a significant reduction in a margin of safety.

A. INTRODUCTION Commonwealth Edison (Comed) proposes to revise Figure.3.4-4a,

" Nominal PORV Pressure Relief Setpoint Versus RCS Temperature For The Cold Overpressure Protection System Applicable up to 5.37' EFPY (Unit 1)," of Technical Specification 3.4.9.3 and the bases associated with TS 3.4.9.3. The index page entry associated with Figure 3.4-4a will also be changed to reflect the changes in Figure 3.4-4a. Figure 3.4-4a describes the nominal Pressurizer Power Operated Relief Valve (PORV) setpoints for the Low Temperature Overpressure Protection System'(LTOPS) as a function of reactor Coolant System (RCS) temperature.

Currently, Figure 3.4-4a is valid until Braidwood Unit 1 reaches 5.37 Effective Full Power Years (EFPY). In addition, the current Figure 3.4-4a contains an administrative limit line to protect PORV downstream piping from water hammer effects during pressurizer solid water conditions, and contains allowances for pressure and temperature instrument uncertainties. The current Figure 3.4-4a also contains allowances for a 50'F thermal transport effect associated with the postulated heat injection  ;

transient.

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1

In order to extend the duration of applicability for Figure 3.4-4a to 8.5 EFPY, remove the pressure and temperature instrument uncertainties, and remove the administrative limit line it is necessary to revise the. current Figure 3.4-4a. The revised Figure 3.4-4a will retain the 50'F thermal transport effect allowance. Additionally, the revised curve accounts for the flow induced pressure difference between the pressure transmitter in the RCS loop piping and the reactor vessel midplane, and takes advantage of a 10% relaxation of the maximum allowable RCS pressure in accordance with American Society of Mechanical Engineers (ASME) Code Case N-514. Comed applied for permission to use the criteria of ASME Code Case N-514 in the determination of LTOPS setpoints via letter dated November 30, 1994.

In. addition the index and bases pages will be revised to reflect the changes to Figure 3.4-4a.

The pressure and temperature instrument uncertainties are being removed as the regulations governing Pressure-Temperature (P-T) limits and LTOPS setpoints; 10 CFR 50 Appendix G, ASME Boiler and Pressure Vessel (BPV) Code Section XI Appendix G, and NUREG 0800 Branch Technical Position MTEB 5-2 (NUREG 0800 BTP 5-2), do not require margins for random instrument uncertainties. Further, compared with the margins inherent in the P-T limits and LTOPS setpoints developed in accordance with these regulations, random pressure and temperature instrument uncertainties are insignificant. With respect to the 800 psig limit employed to protect the PORV discharge piping from water hammer effects, conservatisms in the application of this limit also support removal of pressure and temperature instrument uncertainties and allowances while still providing adequate PORV discharge piping protection.

l The imposition of margin in addition to.that already shown to exist results in several negative effects with no concurrent ,

increase in protection against nonductile failure of the Reactor 1 Pressure Vessel (RPV). This unnecessary margin reduces operating l flexibility by reducing the margin between the maximum allowed l RCS pressure and the minimum RCS pressure for reactor coolant  !

pump (RCP) operation. Excess margin increases the likelihood of LTOPS actuation at lower setpoints and reduces the amount of separation of the setpoints for the PORVs used in the LTOPS.

This endangers RCP seals as LTOPS actuation at setpoints determined by the current methodology may cause RCS pressure to drop below the minimum needed to maintain proper RCP seal differential pressure.

k r nia\twi\ltoprevitt 14 l

The original Byron /Braidwood LTOPS setpoint development took credit for the fact that LTOPS events are most likely to occur with the RCS at isothermal conditions. Thus no. temperature streaming effects would occur. Therefore, the temperature streaming allowance is being removed from the revised Figure 3.4-4a.

B. NO SIGNIFICANT HAZARDS ANALYSIS

1. The proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

The new LTOPS curve will not change any postulated accident scenarios. The Appendix G limits, which is the bases for the revised curve was developed using industry standards and regulations which are recognized as being inherently conservative. The LTOPS curve provide RCS limits to protect the RPV from brittle fracture by clearly separating the region of normal operations from the region where the RPV is subject to brittle fracture.

Using the Regulatory Guide (RG) 1.99, " Radiation Embrittlement of Reactor Vessel Materials," Revision 2, Braidwood Unit 1 Surveillance Capsule U results, and the' requirements of Appendix G to 10 CFR 50, as modified by the guidance in ASME Code Case N-514, a new LTOPS curve was prepared. This new curve, in conjunction with the P-T Limit curves, and the heatup and cooldown ranges provides the required assurance that the RPV is protected from brittle fracture. Removal of pressure and temperature instrument uncertainties and allowances still results in retention of adequate margin to P-T limits. No changes to the design of the facility have been made and no new equipment has been added or removed. The revised LTOPS curve provides assurance that the RPV is protected from brittle fracture.

The bases and index page changes are purely administrative in nature and are designed to reflect the changes in Figure 3.4-4a.

These administrative changes will have no effect on any equipment, system, or operating mode.

Thus, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

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2. The proposed change does not create the possibility of a new or different kind of accident fram any accident previously evaluated.

The use of the new LTOPS curve does not change any postulated accident scenarios. The new curve does not represent any appreciable change in the current methodologies; it merely provides assurance that the RPV is protected from brittle fracture. Removal of pressure and temperature instrument uncertainties and allowances still results in retention of adequate margin to P-T limits. No new accident or malfunction mechanism is introduced by this amendment and no physical plant changes will result from this amendment.

The bases and index page changes are purely administrative in nature and are designed to reflect the changes in Figure 3.4-4a.

These administrative changes will have no effect on any equipment, systam, or operating mode.

Tbcietore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. The proposed change does not involve a significant reduction in a margin of safety.

The new LTOPS curve was generated using conservative methodology with capsule surveillance data. The Appendix G limits, which is the bases for the new LTOPS curve were developed using industry standards and regulations which are recognized as being inherently conservative. Removal of pressure and temperature instrument uncertainties and allowances still results in retention of adequate margin to P-T limits. The use of the new LTOPS limits would not change any postulated accident scenarios.

The bases and index page changes are purely administrative in nature and are designed ta reflect the changes in Figure 3.4-4a.

These administrative changes will have no effect on any equipment, system, or operating mode.

Thus, the proposed change does not involve a significant reduction in a margin of safety.

Therefore, based on the above evaluation, Comed has concluded that these changes involve no significant hazards considerations.

I ki nla\twi\Itoprev\l6 1b

ATTACHMENT D ENVIRONMENTAL' ASSESSMENT FOR PROPOSED CHANGES TO APPENDIX A TECHNICAL SPECIFICATIONS OF FACILITY OPEr TNG LICENSES NPF-72 & _ JPF-77 Commonwealth Edison Company (Comed) has evaluated this proposed license amendment request against the criteria for identification of licensing and regulatory actions requiring environmental assessment in accordance with Title 10, Code of Federal  ;

Regulations, Part 51, Section 21 (10 CFR 51.21). Comed has determined that this proposed license amendment request meets the criteria for a categorical exclusion set forth in .'

10 CPR 51.22 (c) (9) . This determination is based upon the following:

1. The proposed licensing action involves the 'enuance of an amendment to a license for a reactor y a nt to 10 CFR 50 which changes a requirement with r.;spect to installation or use of a facility component located within'the restricted area, as defined in 10 CFR 20, or which changes an inspection or a surveillance requirement. This proposed license amendment request replaces Figure 3.4-4a, " Nominal PORV Pressure Relief Setpoint Versus RCS Temperature For The Cold Overpressure protection System Applicable up to 5.37 .

EFPY (Unit 1)," of Technical Specification 3.4.9.3 with a new Figure 3.4-4a, " Nominal PORV Pressure Relief Setpoint Versus RCS Temperature For The Cold Overpressure protection System Applicable up to 8.5 EFPY (Unit 1)," and makes administrative changes to the ,

bases and index pages for Technical Specification l 3.4.9.3 to support the revised Figure 3.4-4a. 1

2. this proposed license amendment request involves no ,

significant hazards considerations; j

3. there is no significant change in the types or  :

significant increase in the amounts of any effluent that may be released offsite; and

4. there is no significant increase in individual or cumulative occupational radiation exposure.

Therefore, pursuant to 10 CFR 51.22 (b), neither an environmental impact statement nor an environmental assessment is necessary for this proposed license amendment request.

kt nia\tnal\itoprevut 17

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i

ATTACEMENT E American Society of Mechanical Engineers Boiler and PIcssure Vessel Code Section XI Appendix G Figure G-2210-1  ;

l 18

4 APPENDIX G - NONMANDATORY c.22 t4.3 G 22I2 160 I- I /

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  1. 1, - 26.73 1.223 eno (0.0146 (T- ,Typ7 + 1600, en 140 -

_ # f, = refe rence stre se intensity factor, kai6

,T!

  • tempersture et which Kf , is permitted,0F  !

,jg _

nr ,T -,e,e, n,i ooetait, tee,e,e,o,e /

100

>. 90

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4 r M

60

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40 20 0

-240 - 200 -100 -120 -40 -40 0 40 60 120 - 100 200 240 Temperature Retetive to M TNOT U T-M Typf f ,09 FIG. G 22101 I .

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  • O ATTACHMENT F American Society of Mechanical Engineers Boiler and Pressure Vessel Code Section XI Appendix A Figure A-4200-1 i

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3 flg. A4$51 1992 SECTION XI- DIVISION I 220 200 in -

g380 Kg a

140 3

w 120 xu 2

f100 t

,e m -

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40 -

20 -. aruar I i I i I I 1 I 0

I i I I 80 120 100 200

- 80 - 40 0 40 Ir-aruos.s A92 FIC. A42%1 LOWER BOUND K. AND K. TEST DATA FOR SA 533 GRADE B CLASS 1, SA 506 CLASS 2, AND SA 508 CLASS 3 STEELS 392

ATTACHMENT G J.N. Chirigos and T. A. Meyer

" Influence of Material Property Variations on the Assessment of Structural Integrity on Nuclear Components" Journal of Testing and Evaluation Volume 6, Number 5, September 1978 Figure 5 l

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' Indicated System Pressure (PSIG) 8 4800 m Tw sir 1

" M g.o.5" i Fl.w g, 4400

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4000 -

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  • Ki r 3600 - I i I I 3200 - w xic f I 2800 e . i ir F w o.5 tach Flew (Kau +K ITI 2400 - (K ui ' "lT3 g

f l A,,.adin o,

- N. ser. y F.cier 2000 Ki c Cv"'- / (Ki u + xivl

,,, 1 lach Flaw g i 1600

'"lM * "873

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8 k(#"" 2 Kgg + Kgy) 1200 -

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= / ,/ re.iw se i.

Method A 800 - / / r e .i scane u.thod s p,

400 -

! ico 200 soo 400 Soo soo 7ao soo

~

I l l I I I I 0 200 400 600 800 indicated System Temperature (*F)

PIG. S-Reector nssel cooldown limit curws. ATIAt. masimum 3)T/h(60Ylh)llpoig = 6.Vkl'n;Iin. = 2Mmm: T = (4'- 321/l.8l.

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