ML20078S409
| ML20078S409 | |
| Person / Time | |
|---|---|
| Site: | Braidwood  |
| Issue date: | 12/16/1994 |
| From: | Saccomando D COMMONWEALTH EDISON CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM), Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML20078S413 | List: |
| References | |
| NUDOCS 9412280292 | |
| Download: ML20078S409 (11) | |
Text
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Commonwrith Edison J* -N), Downers 1400 Opus Place -
Grove, Ilknois 60515 December 16, 1994 Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington,.D.C. 20555 Attention: Document Control Desk
Subject:
Request to Amend Technical Specification Section 3.4.9.3 Braidwood Station Units 1 and 2 NRC Docket Nos. 50-456/457
Reference:
D. Saccomando letter to Nuclear Regulatory Commission dated November 30, 1994 transmitting request to apply American Society of Mechanical Engineering (ASME) Code Case N-514 Pursuant to 10 CFR 50.90, Commonwealth Edison Company (Comed) proposes to amend Appendix A, Technical Specifications of Facility Operating Licenses NPF-72 and NPF-77. The proposed amendment requests changes to Technical Specifications Section 3.4.9.3 and the associated bases.
The proposed amendment request primarily consists of changes'to Figure 3.4-4a, " Nominal PORV Pressure Relief Setpoint Versus RCS Temperature for the Cold Overpressure Protection System" for Unit
- 1. The proposed revision extends the duration of applicability for the figure to'8.5 effective full power years, removes the pressure and temperature instrument uncertainties and-administrative limit. The proposed revision takes advantage of a:
10% relaxation of the maximum allowable reactor coolant system pressure in accordance with ASME Code Case N-514. Comed applied for permission to use the criteria of this Code Case in the reference letter.
The amendment request is subdivided as follows:
Attachment A: Description and Safety Analysis of Proposed Changes t Attachment B: Proposed Revision to the Technical Specifications Attachment C: Evaluation of Significant Hazards ;
Considerations Attachment D: Environmental Assessment k nla\bwd\1toprev\1 0
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Document Control Desk- December 16,. 1994 Attachment E: ASME Boiler and' Pressure Vessel Code Section XI Appendix G Figure G-2210-1 Attachment F: ASME Boiler and Pressure Vessel Code Section XI Appendix A Figure A-4200-1 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 The proposed changes have been reviewed and approved by the On-site and Off-site Review Committees in accordance with Comed procedures Comed has reviewed this proposed amendment in accordance with 10 CFR 50.92(c) and has determined that no significant hazards consideration exists.
Comed is notifying the State of Illinois of our application for this amendment by transmitting a copy of this letter and the associated attachment to the designated State Official.
Braidwood Unit 1 is expected to exceed 5.37 Effective Full Power Years at midnight July 16, 1995; therefore, Comed requests that the review and approval of the proposed amendment to be completed by July 16, 1995.
To the best of my knowledge and belief, the statements contained in this document are true and correct. In some respects these statements are not based on my personal knowledge, but_on ,
information furnished by other Comed employees, contractor employees, and/or consultants. Such information has been reviewed in accordance with company practice, and I believe it to be reliable.
Please address any further comments or questions regarding this matter to this office.
Si rely,
,c ~4 e ise M. ccomando Nuclear Licensing Administrator Attachments cc: R. R. Assa, Braidwood Project Manager - NRR S. G. Dupont, Senior Resident Inspector - Braidwood B. Clayton, Branch Chief - Region III Office of Nuclear Facility Safety - IDNS ~~q . -
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ATTACHMENT A DESCRIPTION AND SAFETY ANALYSIS OF PROPOSED CHANGES TO APPENDIX A TECHNICAL SPECIFICATIONS OF FACILITY OPERATING LICENSES NPF-72 AND NPF-77 A. DESCRIPTION OF THE PROPOSED CHANGE 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 EPPY (Unit 1)," of Technical Specification (TS) 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.
B. DESCRIPTION OF THE CURRENT REQUIREMENT The current index page VIII states that Figure 3.4-4a is applicable to 5.37 Effective Full Power Years (EFPY) for Unit 1.
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. The current Figure 3.4-4a contains a 60 pounds per square inch gauge (psig) pressure instrument uncertainty, a 13 F temperature instrument uncertainty, a 14'F temperature streaming allowance, and a 508F thermal transport effect allowance associated with the postulated heat injection transient. Figure 3.4-4a also contains an administrative limit line at 638 psig which limits the maximum setting of the PORV in the LTOPS mode to protect the PORV discharge piping from water hammer effects on PORV opening and closing with the RCS and pressurizer in a water solid condition.
The current bases for TS 3.4.9.3 states that either PORV has adequate relieving capability to protect the RCS from overpressurization when the transient is limited to either: (1) the start of an idle Reactor Coolant Pump (RCP) with the secondary water temperature of the steam generator less than or equal to 50 F above RCS cold leg temperatures or (2) the start of a centrifugal charging pump and 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.
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Figure 3.4-4a (3.4-4b) are based upon this analysis and represent-the maximum allowable PORV variable setpoint such that, for the ,
two overpressurization transients noted, the resulting pressure '
will not exceed the Appendix G reactor vessel NDT limits. I 1
C. BASES FOR THE CURRENT REQUIREMENT l
The setpoints provided for the LTOPS are selected such that the i pressure peaks resulting from design basis overpressure events I are limited to values less than those specified by Appendix G of i Title 10 Code of Federal Regulations Part 50 (10 CFR 50), which j incorporates by reference American Society of Mechanical Engineers (ASME)Section XI Appendix G. NUREG-0800, " Branch l Technical Position MTEB 5-2," (NUREG-0800 BTP 5-2) also provides I guidance in this area. )
In March of 1994, Comed submitted an amendment request for a new Figure 3.4-4a to replace a figure that expired at the end of 4.5 EFPY.
In June of 1994, a supplemental request was submitted to the Nuclear Regulatory Commission (NRC). The June 1994 supplemental submittal adjusted tre curve submitted in the March 1994 request to account for a 50'F thermal transport effect on the postulated heat injection transient which had inadvertently left out of the March 1994 curve. In July of 1994, the curve was revised again at the NRC's request to account for a 60 psig pressure instrument uncertainty. This necessitated reducing the duration of applicability of Figure 3.4-4a from 32 EFPY to 5.37 EFPY, and adding a 638 psig administrative limit line to protect the PORV discharge piping from water hammer effects on PORV opening and closing with the RCS and pressurizer in a water solid condition.
D. NEED FOR REVISION OF THE REQUIREMENT Currently, Figure 3.4-4a is valid until Braidwcod Unit 1 reaches 5.37 EFPY. In addition, the current Figure 3.4-4a contains an administrative limit line, and contains allowances for pressure and temperature instrument uncertainties. In order to extend the duration of applicability for Figure 3.4-4a, 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 bases for TS 3.4.9.3, and the TS index page will be revised to reflect the changes to Figure 3.4-4a.
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E.. DESCRIPTIOtt OF THE REVISED REQUIREMENT 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 8.5 EFPY (Unit 1) . "
l' The pressure and temperature instrument uncertainties and allowances will be removed from the revised Figure 3.4-4a. The revised Figure 3.4-4a will still account for the 50 F thermal
- transport effect on the postulated heat injection transient.
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 ASME Code Case N-514. Comed applied for permission to use the triteria of ASME Code Case N-514 in the determination of LTOPS setpoints via letter dated November 30, 1994.
The current bases for TS 3.4.9.3 will be replaced with Insert B.
Insert B reads:
" 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 for 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 instrument uncertainties and does not account for the flow induced pressure difference between the RCS pressure sensor and the Reactor Vessel
. beltline."
The index will be revised to reflect the new duration of applicability for Figure 3.4-4a.
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1 F. BASES FOR THE REVISED REQUIREMENT Pressure-Temperature (P-T) limits and LTOPS setpoints are I
, established to protect the reactor pressure vessel (RPV) from j nonductile failure due to overpressurization. The determinatica of the P-T limits and LTOPS setpoints is based on very conservative assumptions and methodologies. In addition to the
. basic conservative assumptions and methodologies used to develop the P-T and LTOPS curves, the current Figure 3.4-4a contains a 60 1 psig pressure' instrument uncertainty, a 13"F temperature instrument uncertainty, a 14 F temperature streaming allowance, and a 500F thermal transport effect allowance for the postulated heat injection transient.
The proposed revision to Figure 3.4-4a contains only 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 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.
i As the basis for generating the revised Figure 3.4-4a, a revised l steady state 10 CFR 50 Appendix G curve was generated based on j 8.5 EFPY, excluding pressure and temperature instrument i uncertainties, and incorporating a 10% relaxation of the maximum i allowable RCS pressure limit. To determine the flow induced differential pressure effects between the RCS loop piping and the reactor vessel midplane, the following cases were considered:
- 1. 4 RCPs and 0 Residual Heat Removal Pumps (RHR)' operating with RCS temperature greater than or equal to 350 F. l J
The revised 10 CPR 50 Appendix G limit curve pressure limits were then adjusted down by the appropriate amount for each temperature range. Also, 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 pressure limit curve, or the 800 psig PORV piping water hammer load limit, whichever was lower at a given temperature, were then used to develop the revised Figure 3.4-4a.
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The pressure and temperature instrument uncertainties are being removed as the regulations governing 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 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. These margin terms include:
- 1) A safety factor of 2 is applied to the membrane stress intensity factor (pressure). For example, a P-T limit for an allowable pressure of 400 psig is actually based on the stress intensity resulting from a pressure of 800 psig.
- 2) ASME BPV Section XI Appendix G allows the sum of the pressure stress intensity factor (multiplied by a safety factor of two) and the thermal stress intensity factor to be no higher than the reference stress intensity factor Ka as shown in ASME BPV Section XI Appendix G Figure G-2210-1 (Attachment E). The Ka value at a given temperature is the lower bound of all available static, dynamic, and crack arrest fracture toughness data, and is identical to the lower bound crack arrest Ku values shown in ASME BPV Section XI Appendix A Figure A-4200-1 (Attachment F) . This is considerably more conservative than using the crack initiation toughness, Ka (see Attachment F) which ASME BPV Section XI Appendix E permits for evaluating the effects of actual overpressure events on the operability of the RPV, or the actual fracture toughness of the RPV limiting material, which would be expected to fall above the Ka curve.
- 3) A 20 margin on mean predicted shift is included in the Regulatory Guide (RG) 1.99, " Radiation Embrittlement of Reactor Vessel Materials," Revision 2, method for determination of a conservative, upper bound value of adjusted reference temperature. For Braidwood Unit 1 at 32 EFPY with an assumed flaw having a depth equal to 1/4 of the thickness (1/4T) of the RPV, this results in the addition of 56"F to the adjusted reference temperature used in the calculation of P-T limits.
- 4) Stress intensity factors are calculated on the basis of an assumed flaw in the wall of the RPV with a depth equal to 1/4T.
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The degree of conservatism associated with this requirement can be seen in ASME Section XI Appendix E, which permits the use of a 1.0" initiation crack size.
This is considerably smaller than 1/4T, and is consistent with industry experience in performing In Service Inspections of RPV beltline regions.
Figure 5 of J.N. Chirigos and T.A. Meyer, " Influence of Material Property Variations on the Assessment of Structural Integrity of Nuclear Components," Journal of Testing and Evaluation, Vol. 6, No. 5, September 1978, (Attachment G) illustrates, in general, the conservatism typical of P-T limits developed in accordance with 10 CFR 50 Appendix G for a pressurized water reactor at an end-of-life neutron fluence. At the lower end of the operating temperature range, where the concern for protection against nonductile fracture is highest, a margin of at least 600 psig exists above the 10 CFR 50 Appendix G P-T limit if the sum of the actual pressure and thermal stress intensity factors is not allowed to exceed the Ku curve. Even greater margins exist when more realistic flaw sizes and the estimated actual available beltline material toughness is taken into account.
The ASME Section XI Appendix E Paragraph E-1200 Acceptance Criteria can be used to illustrate the conservatism of 10 CFR 50 Appendix G specifically for Braidwood Unit 1. For Braidwood Unit 1, with a 32 EFPY 1/4T adjusted reference temperature of 159 F (56'F of which is RG 1.99, Rev. 2 margin), the maximum allowable pressure for a pressurized thermal transient occurring at 214 F is 2485 psig (the design pressure). The 10 CFR 50 Appendix G steady-state pressure limit at the same temperature is 935 psig, a margin of 1550 psig. The random pressure and temperature instrument uncertainties of 60 psig and 13 F applied in the past to P-T limits are therefore very small when compared with the margin between the 10 CFR 50 Appendix G P-T limits and the ASME Section XI Appendix E allowable values. Additionally, the original Braidwood LTOPS setpoint development took credit for the fact that LTOPS events are most likely to occur with the RCS at isothermal conditions. And, for isothermal pressure transients, the margin available from ASME Section XI Appendix E is even higher.
The ASME Code explicitly recognized the amount of margin in the 10 CFR 50 Appendix G P-T limits in the 1993 addenda. ASME Code Section XI Appendix G paragraph G-2215 incorporated the guidance contained in ASME Code Case N-514, and allows LTOPS setpoints to exceed the usual P-T limits by 10% as a standard practice.
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 RPV.
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a This unnecessary margin reduces operating flexibility by reducing the margin between the maximum allowed RCS pressure and the minimum RCS pressure for RCP operation. Excess margin increases the likelihood of LTOPS actuation at lower setpoints, and reduces the amount of separation of the setpoints for the Power Operated Relief Valves (PORV) 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.
The original Braidwood LTOPS setpoint development took credit for the fact that LTOPS events are assumed to occur with the RCS at isothermal conditions. Thus no temperature streaming effects would occur. Therefore, the temperature streaming allowance of 14 F is being removed from the revised Figure 3.4-4a.
Temperature and pressure instrument uncertainties have not been applied to the 800 psig PORV discharge piping limit in the development of the revised Figure 3.4-4a. The 800 psig value was chosen to protect the PORV discharge piping from water hammer effects if the PORV were to actuate with the RCS and the pressurizer in a water solid condition. Under these conditions, water hammer is a concern only when the PORV initially opens or closes. Once the PORV discharge piping is water solid water hammer is no longer a concern until the PORV closes and interrupts the stream.
To address this water hammer concern, the PORV setpoints of the revised Figure 3.4-4a are selected to ensure that the peak pressure reached during a PORV opening event is less than or equal to 800 psig. The PORV opening event was chosen as the most limiting since the pressure at the next potential water hammer opportunity; PORV closure, will be approximately 20 psig less than the PORV opening setpoint.
As a further conservatism, the pressure transmitters used for the LTOPS are calibrated without a head correction for the elevation difference between the pressure transmitter and the top of the pressurizer. 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. Finally, the 800 psig PORV discharge piping limit is constant regardless of temperature.
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t Thus, since-the method of choosing PORV setpoints limits the peak pressure reached on a PORV opening event to less than or equal to 800 psig, the conservatism of not correcting for elevational differences between the LTOPS pressure transmitters and the top of the pressurizer further limits pressure at the PORV discharge, and the 800 psig PORV discharge piping limit is independent of temperature, it is unnecessary to account for pressure and temperature instrument uncertainties in the application of this limit.
The index page is being changed to reflect the change to Figure
~ 3.4-4a. The bases page change.provides clarification with regard to the various uncertainties which were excluded when developing-Figure 3.4-4a (Unit 1) and Figure 3.4-4b (Unit 2).. Please note that the inputs to the Unit 1 and Unit 2 curves differ because the Unit 2 curve was previously approved using a different criteria.
G. IMPACT OF THE PROPOSED CHANGE The new low temperature overpressure protection 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. 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 curve provides assurance that the RPV is protected from brittle fracture.
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, system, or operating mode.
! Thus, this proposed amendment will have no negative effect on any
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system or operating mode.
H. SCHEDULE REQUIREMENTS Comed requests that this proposed amendment be approved prior to July 10, 1995. Unit 1 is currently predicted to reach 5.37 EFPY at midnight July 16, 1995.
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l I. IDENTIFICATION AND DISCUSSION OF IRREVERSIBLE CONSEQUENCES The new low temperature overpressure protection 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. 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 curve provides assurance that the RPV is protected from brittle fracture.
No new accident or malfunction mechanism is introduced by this l
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.
Also, the change clarifies the various inputs used for the Unit 1 and Unit 2 curves. These administrative changes will have no effect on any equipment, system, or operating mode.
There is no significant change in the types or significant increase in the amounts of any effluent that may be released offsite, and there is no significant increase in individual or cumulative occupational radiation exposure.
Thus, no irreversible consequences have been identified, f
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