ML20084U223

Proposed Tech Specs Reflecting Addition of Instrument Uncertainty & New Capsule Surveillance Data

ML20084U223
Person / Time
Site:	Braidwood
Issue date:	06/08/1995
From:	COMMONWEALTH EDISON CO.
To:
Shared Package
ML20084U198	List: ML20084U195 ML20084U223
References
NUDOCS 9506130131
Download: ML20084U223 (18)

Text

.. . . . - - - . - _ . .

i

, LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS

.i SECTION ME TABLE 3.4-2 REACTOR COOLANT SYSTEM CHEMISTRY LIMITS............... 3/44-25 TABLE 4.4-3 REACTOR COOLANT SYSTEM CHEMISTRY SURVEILLANCE REQUIREMENTS..................-......................

3/44-26 3/4.4.8 SPECIFIC ACTIVITY........................................ 3/44-27 FIGURE 3.4-1 DOSE EQUIVALENT I-131 REACTOR C0OLANT SPECIFIC l ACTIVITY LIMIT VERSUS PERCENT OF RATED THERMAL POWER WITH THE REACTOR COOLANT SPECIFIC ACTIVITY -

>l *C1/ GRAM DOSE EQUIVALENT I-131.................. 3/4 4-29 l

-TABLE 4.4-4 REACTOR COOLANT SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PR0 GRAM.................................... 3/44-30 i 3/4.4.9 PRESSURE / TEMPERATURE LIMITS j Reactor Coolant System................................... 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 l s LIMITATIONS APPLICABLE UP TO 16 EFPY (UNIT 2)...... 3/44-34 I FIGURE 3.4-3a REACTOR' COOLANT SYSTEM C00LD0WN LIMITATIONS APPLICABLE UP TO 32 EFPY (UNIT 1)...... 3/44-35 {

FIGURE 3.4-3b REACTOR COOLANT SYSTEM C00LDOWN LIMITATIONS APPLICABLE UP TO 16 EFPY (UNIT 2)...... 3/44-36 I TABLE 4.4-5 REACTOR VESSEL MATERIAL SURVEILLANCE PROGRAM - WITHDRAWAL SCHEDULE....................... 3/44-37 P re s s u r i z e r . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/44-38 Overpressure Protection Systems.......................... 3/44-39 FIGURE 3.4-4a DOMINAL PORY PRESSURE RELIEF SETPOINT VERSUS ,

T,f - RCS TEMPERATURE FOR THE COLD OVERPRESSURE  :

PROTECTION SYSTEM APPLICABLE UP TO #$W EFPY (Unit 1) 3/44-40 l '

FIGURE 3.4-4b NOMINAL PORY PRESSURE RELIEF SETPOINT VERSUS RCS l

TEMPERATURE FOR THE COLD OVERPRESSURE PROTECTION l

2)..........

SYSTEM APPLICABLE UP TO 16 EFPY (UNIT 3/4 4-40a ,

3/4.4.10 STRUCTURAL INTEGRITY..................................... 3/44-42 3/4.4.11 REACTOR COOLANT SYSTEM VENTS............................. 3/44-43 9506130131 950608 PDR ADOCK 05000456 P PDR 7

BRAIDWOOD - UNITS 1 1 2 VIII AMENDMENTNO.p3

i REACTOR COOLANT SYSTEM

/ ;. '0VERPRESSURE PROTECTION SYSTEMS - - -

~

LIMITING CONDITION FOR OPERATION r

At least two everpressure protection d'avices shall be OPERABLE, and 3.4.9.3 each-device shall be either: '

a. A residual heat removal (RHR) 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 overpressure protection devices +

inoperable in MODE 4, restore two overpressure protection devices to l 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 />. l i

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, j 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 Cosmission pursuant to i Specification 6.9.2 within 30 days. The report shall describe the '

circumstances initiating the transient, the effect of the PORVs, RM 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.

I AMENDMENT NO. 53 m~*m - UNITS 1 & 2 3/4 4-39

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ATTACHMENT C S EVALUATION OF SIGNIFICANT HAZARDS CONSIDERATIONS FOR PROPOSED CHANGES TO APPENDIX A-TECHNICAL SPECIFICATIONS OF FACILITY OPERATING LICENSES NPF-72 AND NPF-77 lI Commonwealth Edison has evaluated this pr'oposed amendment and determined that it involves no significant hazards '

i considerations. According to Title 10 Code of Federal  :

Regulations Section 50 Subsection 92 Paragraph c I (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 i The Cold Overpressure Protection System Applicable up to 5.37  ;

EFPY (Unit 1)," of Technical Specification (TS) 3.4.9.3. The index page entry associated with Figure 3.4-4a will also be i changed to reflect the changes in Figure 3.4-4a. The fornat of Figure 3.4-4a will be revised to improve readability. Figure 3.4-4a describes the nominal Pressurizer Power Operated Relief Valve (PORV) setpoints for the Low Temperature Overprdssure '

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 at 638 pounds per square inch gauge (psig) to protect PORV downstream piping from water hammer effects during pressurizer solid water conditions and contains allowances for a 60 psig pressure instrument uncertainty, a 13 F temperature instrument uncertainty and a 140F temperature streaming allowance.

Rinla\bwd\bwltops3 13

s The current Figure 3.4-4a also contains allowances for a 50 F thermal transport effect associated with the postulated heat injection transient.

In order to extend the duration of applicability for ,

Figure 3.4-4a to 16 EFPY and remove the administrative limit line  !

it is necessary to revise the current Figure 3.4-4a. l The current 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)," will be replaced with a new Figure 3.4-4a, " Nominal PORV Pressure Relief Setpoint Versus RCS Temperature For The Cold Overpressure Protection System Applicable up to 16 EFPY (Unit 1)."

As the basis for generating the revised Figure 3.4-4a, a revised steady state 10 CFR 50 Appendix G Pressure Temperature (PT) limit curve was generated for 16 EFPY using the data from WCAP 14241,

" Analysis of Capsule X from the Commonwealth Edison Company Braidwood Unit 1 Reactor Vessel Radiation Surveillance Program,"

and WCAP 12685 " Analysis of Capsule U from the Commonwealth Edison Company Braidwood Unit 1 Reactor Vessel Radiation Surveillance Program." These documents were submitted to the NRC on March 21, 1995, and October 22, 1990 respectively. The revised Appendix G PT limits curve also 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, supplemented by a letter dated May 8, 1995.

Finally, a constant 800 psig RCS pressure value was selected to control PORV piping loads due to waterhammer effects from PORV actuation during water solid pressurizer conditions. The pressure values on the revised 10 CFR 50 Appendix G PT limit curve, or the 800 psig PORV discharge piping water hammer load limit, whichever was lower at a given temperature, were then used to develop the revised Figure 3.4-4a.

That portion of the revised Figure 3.4-4a limited by the Appendix G PT limits retains the 60 psig pressure instrument uncertainty, the 13 F temperature instrument uncertainty and the 50 F thermal D

transport effect for heat injection events.

i Kinla h d % 1ttps3 14 I

For that portion of the revised Figure 3.4-4a limited by the 800 s psig PORV discharge piping limit, the 60 psig pressure instrument uncertainty is retained with credit taken for the elevational difference between the PORV LTOPS pressure sensor and.the FORV itself. This elevation difference is approximately 74 feet, so the actual pressure at the PORV discharge will be approximately 32 psig less than the pressure seen by the pressure transmitter.

The 13 F temperature instrument uncertainty and the 50 F thermal transport effect for heat injection events are also retained in this section of Figure 3.4-4a.

The 14 F temperature streaming allowance is not included in .the LTOPS setpoint curve since WCAP 14040, " Methodology used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves," Section 3.2.2, " Pressure limit Selection," assumes LTOPS events are most likely to occur during isothermal conditions in the RCS. Thus, temperature streaming

, would not be a consideration.

The revised Figure 3.4-4a also d.oes not contain the administrative limit line at 638 psig.

The TS index page entry associated with Figure 3.4-4a is being changed to reflect the change in the duration of applicability of Figure 3.4-4a, and the format of Figure 3.4-4a is being changed to improve readability.

B. NO SIGNIFICANT RAZARDS 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 revised curve was developed using industry standards and regulations which are recognized as being inherently conservative. Appropriate instrument uncertainties and allowances have been included in the development of the LTOPS curves. The PT and LTOPS curves provide RCS pressure limits to protect the Reactor Pressure Vessel (RPV) from brittle fracture '

by clearly separating the region of normal operations from the region where the RPV is subject to brittle fracture.

Using Regulatory Guide (.G) R 1.99, " Radiation Embrittlement of Reactor Vessel Materials," Revision 2, Braidwood Unit 1 Surveillance Capsule U and Capsule X 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 PT Limit curves, and the heatup and cooldown ranges provides the required assurance that the RPV is protected from brittle fracture.

K inla \twd\tvltops) 15

-_-___-m___ . _ . _ _

No changes to the design of the facility have been made, no new s equipment has been installed, and no existing equipment has been removed or modified. This amendment will not change any system operating modes. The revised LTOPS curve provides assurance that the RPV is protected from brittle fracture.

l The index page and format changes are purely administrative in nature and are designed to reflect the change in the duration of applicability of Figure 3.4-4a and improve the readability of 1 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 l in the probability or consequences of an accident previously ,

l

. evaluated.

2. The proposed change does not create the possibility of a new

, or different kind of accident from any accident previously evaluated.

The use of the new LTOPS curve d'oes not change any postulated accident scenarios. The new LTOPS curve was generated using Braidwood capsule surveillance data and an approved, conservative methodology. No new equipment will be installed, and no existing equipment will be modified. No new system interfaces are created, and no existing system interfaces are modified. The new LTOPS curve provides assurance that the RPV is protected from brittle fracture.

No new accident or malfunction mechanism is introduced by this amendment.

The index page and format changes are purely administrative in l nature and are designed to reflect the change in the duration of applicability of Figure 3.4-4a, and improve the readability of Figure 3.4-4a. These administrative changes will have no effect on any equipment, system, or operating mode.

1 Therefore, the proposed change does not create the possibility of I a new or different kind of accident from any accident previously J evaluated. l

3. The proposed change does not involve a significant reduction l in a margin of safety. I The new LTOPS curve was developed using industry standards and regulatio'ns which are recognized as being inherently  ;

conservative. Appropriate instrument uncertainties and I allowances are included in the development of the new LTOPS curve. This amendment will not change the operational characteristics or design of any equipment or system.

Kinla\bwd\bwltops3 16

s All accident analysis assumptions'and conditions will continue to be met. The RPV is adequately protected from non-ductile failure by the revised LTOPS curve.

] The index page and format changes are purely administrative in nature and are designed to reflect the change in the duration of

~~

applicability of Figure 3.4-4a, and improve the readability of 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.

4 e

K i nle \twd\tMit ops 3 17

l s

ATTACHIGNT D ENVIRONMENTAL ASSESSMENT FOR PROPOSED CHANGES TO APPENDIX A-TECHNICAL SPECIFICATIONS OF i FACILITY OPERATING LICENSES  !

NPF-72 AND NPF-77 Commonwealth Edison Company (Comed) has evaluated this proposed i

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

1. The proposed licensing

• action involves the issuance of an ==and= ant to a license for a reactor pursuant to 10 CFR 50 which changes a requirement with respect 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 16 EFPY (Unit 1)," and makes administrative changes to the index page for Technical Specification 3.4.9.3 to support the revised Figure 3.4-da. The format of Figure 3.4-4a is also being revised to improve readability. .

2. This proposed license amend ==nt request involves no significant hazards considerations.

As demonstrated in Attachment C, this amendment request involves no significant hazards considerations.

3. There is no significant change in the types or significant increase in the amounts of any effluent that mm.y be released offsite.

This amendment request will not result in the installation of any new equipment, or the modification'of any existing equipment. i Kinla\bwd\bwltops) 18 i

1

-- -. .__-_-_______--_-____\

s No changes will be made in the mode of operation of any plant system or equipment. I No new release paths or mechanisms will be created by this amendment.

Thus, there is no significant change in.the types or significant increase in the amounts of any effluent that may be released off site.

4. There is no significant increase in individual or cumulative occupational radiation estposure.

No new equipment will be installed, and no existing equipment will be modified. No new operating modes or procedures are created. Plant operation will remain unchanged.

Thus, there is no significant increase in individual or r:umulative 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.

Kinla\bwd\bwltops3 19

._ 3 ATTACHBGENT E y i

Graphs and Tables 1

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105 621 - 561 617 583 583 '

110 621 561- 617 583 583 -'

110 798 730 810 776 770 115 822 782 838 004 770 120 849 789 888 790 770 '

125 878 818 000 822 770 130 910 856 935 857 770 135 944 884 973 885 770 140 981 I 921 1013 935 770 145 1020 900 1056 978 770  ;

150 1082 1002 1103 1025 770 155 1100 1048 1153 1975 770 i 100 1154 10 5 1208- 11 3 770 165 1200 1140 1264 11 5 770 i 170 1285 12N 1328 1248 770 1 175 1325 1285 1392 1314 770  :

ISO 1390 13 5 1483 1385 770 185 1458 1388

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200 1688 16N 1803 1725 770 l' 205 1791 1731 1904 18 3 770 210 1880 18 3 2011 1933 770 i 215 1984 1934 2127 2046 770 i 220 2100 2000 2127 2049 770 '

, 225 2228 21 5 2168 2008 770 230 2354 2284 2294 2216 770 235 2490 24 5 2430 2352 770 1 240 2633 2573 2573 2495 770  ;

245 2788 27 3 2728 2848 770 l 250 2947 2887 2887 2000 770

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Table 2 .

PORV Setpoint and Transient Pressures instrument Enor Instrument Adjusted Appendix Appendix G Error G with Code Case Adjusted for Adjusted N-514 Reduced by (2)

Unadjusted Instrument Appendix G Appropriate DP Nominal Transient Peak RCS Appendix G Error with Code Value and Linuted PORV Pressure Transient RCS

..Ieme__ _ _.valu_e_ . _ap.est _C_ ass.ttSt.4 ...p. P!ain_g.u_rnit__ sesto_ int __ ov3rshogt_j_ _P_ress_u,re ,,

deg F psig psig psig psl0

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98i0 psig psig 70 621 561 617 583 553

• 30 583 100 621 561 617 583 553 30 583 120 849 789 868 770 554 28 582 150 1062 1002 1103 770 560 34 594 200 1699 1639 1803 770 819 52

' 871 250 2947 2887 2887 770 697 73 770 300 (4) (4) (4) 770 682 88 770 Notes:

1. When the Appendix G with Code Case N-514 exceeds 770 (800- 30 psi elevation adjusted instrument uncertainty) psig, the 770 psig PORV piping pressure limit govems and is used for LTOP analysis.

2. The nominal PORV setpoint value from the proposed Tech Spec Figure 3.4-4a.

This value represents the highest aBowed PORV setpoint. Actual PORV setpoints may be less than the nominal value given.

3. Differential pressure effeds are the pressure corrections to account for flow induced pressure differences between the reactor vessel .oenterEne and the location of the pressure transmitter. Differe,ntial pressure effects are not applicable for the 770 (800 - 30 psi uncertainty) psig PORV piping pressure limit since the reactor vessel center line is no longer the pressure limiting location in the reactor coolant system.

4. Appendix G values were not calculated for temperatures in excess of 250 deg F since the Appendix G pressure limit at 250 deg F is 2947 psig which exceeds the normal operating pressure of 2235 psig.

5. Pressure instrument unartainty is accounted for as follows Appendix G Limits - Appendix G lirnits are adjusted down 60 psig prior to applying the 10% allowance of Code Case N 514.

800 psig Limit - The PORV piping limit is adjusted down 60 psig for instrument uncertainty and then adjusted up 30 psig due to the 74 ft (appmx.) elevational difference between the pressure instrument location and the top of the pressurizer.

The net adjusted PORV piping limit becomes 770 psig.

Figure 1 '

Braidwood Unit 1 Appan/Hx G and Code Case N-514 Limits at 16 nFY 3500 I

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ATTACHMENT '

F REVIEW OF QUESTIONS AND RESPONSES APRIL 20, 1995 REQUEST FOR ADDITIONAL INFORNATION 1

1. Provide the revised Appendix G curves based on 8.5 effective l full power years (EFPY) for the reactor heatup and cooldown  !

process.

See Figures 1 and 2 and Table 1 of this attachment.

1

2. Provide the curves of pressure / temperature (P/T) limits which are developed in accordance with ASME Code Case N-514. l See Figures 1 and 2 and Table 1 of this attachment.

3. It is our understanding tha't the instrument uncertainties .

have not been incorporated in the propcsed figure 3.4-da '

(power operated relief valve (PORV) setpoints for low-temperature overpressure protection (LTOP) applicable up to 8.5 EFPY) as well as in the proposed P/T limits in Item 2 above. This design is not acceptable to the staff. The margins in the Appendix G curves can not be used to justify  ;

the elimination of the instrn==nt uncertainties in the LTOP '

setpoints. Other wise the P/T limits will not be adequately protected by these setpoints. Please provide a revised Figure 3.4-da for staff review.

See Insert A of Attachment B of this amendment request.

4. Provide a discussion on the change made for the l

administrative limit to protect the PORY discharge piping fross water ha - r effects. l See the discussion in Section F, " Bases of the Revised l

Requirement," of Attachment A of this amendment request. '

5. Technical Specification (Ts) 3.4.9.3 is effective at 350'F <

(Mode 4 and below). The proposed PORV setpoints cover )

4 0 0'F . What is the actual enable temperature for LTOP7

• l Discuss the basis for this LTOP enable temperature in light I of Appendix G requirements. I l

Per Braidwood Station operating procedures and TS, the LTOP system is physically enabled at a Reactor Coolant System (RCS) temperature of 350'F (entry into Mode 4) .

Kinla\bwd\bwltops) 21 i

F 7 1

At Braidwood, the RHR inlet relief valve capacity was verified to ~

l be sufficient to provide overpressure protection for the RHR system for the mass input from two centrifugal 1 charging pumps operating and discharging into the RCS in an unthrottled condition while letdown is isolated. l I

The LTOP design basis mass injection transient postulates one i centrifugal charging pump injecting to the RCS in an unthrottled i condition with RC3 letdown isolated. Thus, the preceding l discussion shows that, at Braidwood,.the RHR inlet relief valve is capable of procecting the RHR system during the design basis LTOP mass injectior event.

2. RHR pressure relief valves may not be able to relieve their  ;

rated capacity due to a greater than estimated back-l pressure.

The Westinghouse valve design data used to procure the Braidwood relief valves assumes that the valve discharge pressure is 100 psig and that the discharge flow is two phase at the maximum temperature. The Braidwood RHR relief valves discharge to the Recycle Holdup Tanks. This provides a backpressure of only 12 psig versus 100 psig. Thus, at Braidwood, the valve backpressure is significantly less than the estimated backpressure and issua 2 is not a concern.

3. Inconsistencies may exist in RHR relief valve design basis documentation.

A review was conducted of the Westinghouse design information ar.d the Braidwood Updated Final Safety Analysis Report and no

~

inconsistencies were identified.

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