ML20059H576
| ML20059H576 | |
| Person / Time | |
|---|---|
| Site: | Zion File:ZionSolutions icon.png |
| Issue date: | 01/19/1994 |
| From: | Simpkin T COMMONWEALTH EDISON CO. |
| To: | Murley T NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM), Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML20059H580 | List: |
| References | |
| NUDOCS 9401280141 | |
| Download: ML20059H576 (57) | |
Text
,
) Commonwealth Edison e'-
1400 Opus Place Downers Grove. Illinois 60515 d.
January 19,1991 Dr. Thomas E. Murley, Director Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Attn: Document Control Desk
Subject:
Zion Station Units 1 and 2 Application for Amendment to Facility Operating License DPR-39 and DPR-48 NRC Docket Nos. 50-295 and 50-304
Dear Dr. Murley:
Pursuant to 10 CFR 50.90, Commonwealth Edison (CECO) proposes to amend Appendix A, Technical Specification of Facility Operating Licenses, DPR-39 and DPR-48. The proposed amendment revises the minimum Reactor Coolant System temperature required for reactor criticality, in addition to editorial changes in the affected specifications.
A detailed description of the proposed change is presented in Attachment A.
The revised Technical Specification pages are contained in Attachment B.
The proposed changes has been reviewed and approved by both on-site and off-site review in accordance with CECO procedures. CECO has reviewed this proposed amendment in accordance with 10 CFR 50.92(c) and has determined that no significant hazards consideration exists. This evaluation is documented in Attachment C. An Environmental Assessment has been completed and is contained in Attachment D.
CECO is notifying the State ofIllinois of our application for this amendment by transmitting a copy of this letter and its attachments to the designated State Official.
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Dr. Thomas E. Murley January 19,1993 To the best of my knowledge and belief the statements contained herem are
. true and correct. In some respects, these statements are not based on my personal
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knowledge but upon information received from other Commonwealth Edison and contractor employees. Such information has been reviewed in accordance with Company practice and I believe it to be reliable.
Please direct any questions regarding this matter to this office.
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OFFICIAL SEAL ll MARY JO YACK ll NOTARY PUBLIC STATE oF ILLINOIS Sincerel '
E MY COMMISSION EXPmES:11/29/97 Llv:::::::::::::
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/1 T.W. Simpkin Nuclear Licensing Administrator Attachments cc:
C.Y. Shiraki, Project Manager - NRR j
J.D. Smith, Senior Resident Inspector - Braidwood Document Control Desk - NRR Region III Office Office of Nuclear Facility Safety - IDNS l
ZOSR-062-93 2
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ATTACIIMENT A ZION NUCLEAR GENERATING STATION DESCRIPTION AND SAFETY ANALYSIS FOR PROPOSED CIIANGES TO APPENDIX A TECHNICAL SPECIFICATIONS FACILITY OPERATING LICENSES DPR-39 AND DPR-48 LICENSE AMENDMENT REQUEST 93-09 MINIMUM TEMPERATURE FOR CRITICALITY i
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Background.
As a result of the reviews associated with Zion Station's ongoing Technical Specification Improvement Project, the adequacy of the 500 F Minimum Temperature for Criticality specification was questioned since the core neutronic, setpoint, and accident analysis typically assume operation at, or near, no load average Reactor Coolant System Temperature (Tavg) in the analyses of Hot Zero Power (HZP) operations and transients. In December of 1992, Westinghouse confirmed that the Zion Units were not explicitly analyzed for critical a
temperatures as low as 500"F.
In January of 1993, notification was made in accordance with 10CFR21 Section 21.21(a)(1) concerning setpoint methodology and the selection oflimiting core parameters for various design basis transients which can occur at low power.
1 It was CECO's judgement that at temperatures below 530 F the Power Range
.)
Nuclear Instrumentation Low Setpoint Reactor Trip could be nonconservative, since the Power Range channel may be affected by the increased density of the coolant at lower temperatures. Transients which rely on the Power Range Low Setpoint Reactor Trip for primary protection include the Rod Control Cluster Assembly (RCCA) Bank Withdrawal From Suberitical and RCCA Ejection. The remaining low power transients (Boron Dilution During Startup, Steamline Break, and Low Power Feedwater Malfunction) rely on the Power Range Low Setpoint Reactor Trip as backup only.
i Prior to the Part 21 notification, CECO requested that Westinghouse perform a safety evaluation for criticality at a temperature greater than or equal i
to 530 F. A minimum temperature for criticality of 530 F was chosen for the evaluation based on several considerations:
1)
Narrow Range RTDs respond from 530 to 630 F, 2)
At a temperature of greater than or equal to 530 F the Power Range Nuclear Instruments are within their analyzed range of operation, and 3)
The expense and complexity of an analysis at a temperature below 530"F was also prohibitive.
As an interim measure, administrative controls were implemented which established a minimum Tavg of 543 F for reactor criticality. This temperature was based on known available margins in the Zion analyses. The controls were revised in February of 1993 based on an interim evaluation performed by CECO Nuclear Engineering & Techno:ogy Services (NETS). This evaluation conservatively defined a normal operating temperature band down to 540 F and i
transient operations to 530 F. The revised administrative controls provided auditional operating requirements consistent with the interim evaluation. The -
requirements of the administrative controls are in the process of being incorporated into station procedures. The NETS interim evaluation is enclosed as to this attachment.
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The final Westinghouse Evaluation has been received by CECO and has been reviewed by NETS. The evaluation provides justification for revising the Minimum Temperature for Criticality to 530 F and demonstrates that the results and conclusions in the Safety Analyses remain valid at that temperature. The Westinghouse Safety Evaluation (SECL 93-002) is included as Enclosure 2. The CECO NETS review of the Westinghouse Safety Evaluation is included as Enclosure 3.
Change Description Zion Technical Specification 3.2.1.C.2.b requires that the reactor coolant system average temperature be at least 500 F prior to achieving criticality.
However, the licensing basis analysis performed by Westinghouse assumes, as an initial condition for the zero load transients, that the reactor is at a nominal no-load temperature of 547 F. As a result of this discrepancy, a reevaluation of the zero load transients was performed at a reduced RCS average temperature of 530 F. It was concluded by Westinghouse and CECO NETS, that the results and conclusions in the analysis remain valid at the reduced temperature. Based on this evaluation, it is proposed that the minimum temperature for criticality be changed to 530' F.
An additional specification is proposed to require that each reactor coolant system loop temperature (Tavg) be greater than or equal to 530 F while in Mode 2 or 7. An action statement will be added for the case where a loop temperature drops below 530 F. Several administrative changes are also proposed to Section 3.2 of the Technical Specifications to remove specifications which no longer apply due to completion of the EAGLE 21 Modification on Unit 2. Also, requirements of specifications 3.2.1.F and 4.2.1.F referring to the 12% Boric Acid System and heat tracing are deleted due to completion of the 4% Boric Acid System Modifications en i
both units.
I Current Reauirement and Bases 1
Zion Technical Specification 3.2.1.C.2.b states "Immediately prior to startup, the reactor coolant temperature shall be shown to be greater than 500 F." This requirement applies to both Unit 1 and Unit 2. The Minimum Temperature for Criticality specification was established to limit the range and associated uncertainty of the Moderator Temperature Coefficient (MTC) limits. However, no Bases for this requirement are explicitly given in the current Zion Station custom -
specifications. The Bases section of the Improved Standard Technical Specifications (NUREG-1431) for Minimum Temperature for Criticality states that this requirement ensures that:
1)
The Moderator Temperature Coefficient (MTC) is within its analyzed temperature range.
2)
The Trip Instrumentation is within its normal operating range.
3)
The Pressurizer is capable of being in an OPERABLE status with a steam bubble.
4)
The R-eactor Vessel is above its minimum RT temperature.
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Section 3.2 of Technical Specifications presently includes numerous requirements which applied only to Unit 2 until startup from refueling outage Z2R12. With the completion of the EAGLE 21 Modification on both units, those requirements are no longer applicable. The requirements which previously applied only to Unit 1, now apply to both Unit 1 and 2. Requirements in specification 3.2.1.F and 4.2.1.F refer to the 12% Boric Acid system and heat tracing. The 12%
Boric Acid Systems have been replaced by 4% Boric Acid Systems for both Units and heat tracing is no longer required. Specifications currently exist in Technical Specification 3.2.1.F and 4.2.1.F that apply to the 4% Boric Acid Systems.
Revised Requirement and BnseJs The administrative changes to delete requirements for the 12% Boric Acid Systems and heat tracing and the requirements specific to Unit 2 until startup from Outage Z2R12 no longer apply. It is proposed that these requirements be deleted to simplify the Technical Specifications from a human factors perspective. There will be no impact if these specifications are deleted. Corresponding administrative changes to the Bases section are also proposed.
It is proposed that Specification 3.2.1.C.2.a be revised to read "Immediately prior to startup, the reactor coolant temperature shall be shown to be greater than 530 F." A specification (3.2.1.C.2.b) will be added which reads "Each reactor coolant system loop temperature (Tavg) shall be greater than or equal to 530 F." This specification will be applicable while in Modes 2 (Hot Standby) or 7 (Low Power Physics Testing) and will have an Action statement which reads "With a reactor coolant system loop temperature (Tavg) less than 530 F, restore Tavg to within its limit within 15 minutes or be in Mode 3 within the next 15 minutes." The bases section of Technical Specifications Section 3.2 will be revised to include bases for the requirement which are consistent with Standard Technical Specifications. The revised and new requirements and bases as well as the editorial changes to Specification 3.2 are shown in Attachment B.
The requirement for a minimum temperature for criticality is based on the need to ensure that the unit is within analyzed parameters prior to bringing the reactor critical.
The first condition to be satisfied is that the MTC is within its analyzed i
range. For the analysis performed to support the VANTAGE 5 fuel upgrade at Zion Units 1 and 2, a positive MTC limit of +7 pcmrF was assumed. This limit is established in Specification 3.2.1.C.1.a. This requirement is waived during low power physics tests to permit the measurement of the reactor moderator coefficient and other core physics parameters. During the early part of a fuel cycle, the MTC may be slightly positive at coolant temperatures in the power range. The MTC will be most positive at the beginning of the cycle when the Boron concentration is greatest. When control rods are inserted, the temperature at which the MTC becomes negative is lower. Therefore, with the operational control rod program, the MTC is assured to be less than the limit of +7 pcmr F. As a result, the potential for operation with an MTC more positive than the MTC l
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assumed in the Safety Analysis is small at 530 F. This determination is'made based
-on the attached Westinghouse evaluation and the current Technical Specification
- Bases for the MTC limit given in Section 3.2 and the interim evaluation performed by NETS (Enclosure 1).
The second condition to be satisfied is that protective instrumentation is within its normal operating range. Westinghouse evaluated the effects of zero power operation (criticality) at temperatures down to 530"F on protection function Process licasurement Accuracy (PMA) terms. The CECO Nuclear Engineering Department (NED) reviewed the revised PMA terms as they apply to the CECO setpoint accuracy calculations and concluded that the Zion Technical Specification setpoints at d Safety Analysis limits are not affected by a minimum critical temperature of 530*F. Enclosure 4 contains the CECO NED analysis of Reactor Trip and Engineered Safety Features Actuation System (ESFAS) Setpoints at a minimum temperature for criticality of 530 F.
The third condition to he satisfied is that the pressurizer is capable of being in an operable status with a steam bubble. Per Zion Technical Specification 3.3.1.D, the pressurizer is required to be operable with at least 150kW heater capacity and a water level less than 92% in Modes 1,2, and 3. During plant heatup (typically in Mode 5) saturated conditions are established in the pressurizer and a bubble is drawn by energizing the heaters and increasing normal letdown. As the RCS pressure and temperature increase, using pump heat from the Reactor Coolant Pumps, the Pressurizer is maintained at saturation conditions. When RCS pressure reaches nominal 2235 psig, the Pressurizer temperature is at the saturation temperature of 653* F, independent of the temperature (Tavg) of the rest of the RCS. As long as system operating pressure at the time of reactor startup remains unchanged (i.e. 2235 psig), the operating temperature in the Pressurizer will also remain unchanged and will not be affected by Tavg. Automatic control systems maintain the Pressurizer at saturation conditions at nominal 2235 psig and 653*F.
Thus, startup at a lower Tavg (530 F) will have no affect on the operating conditions or operability of the Pressurizer at startup. The attached. Westinghouse Safety Evaluation provides justification that, for startup with Tavg at 530* F, the Pressurizer is operable with a steam bubble.
The fourth condition to be satisfied is that the Reactor Vessel is above its Minimum RT3n7. The temperature and pressure changes in the RCS during heatup are limited per the requirements of ASME Section'lII, Appendix G, and 10CFR50, Appendix G. The requirements are specified in the Zion Technical Specifications Section 3.3.2 in the form of heatup and cooldown curves, and are based 'on the -
fracture toughness properties of the ferritic materials in the Reactor Vessel. The requirements also specify the minimum criticality curve, which provides the limiting-temperature and pressure conditions allowed by 10CFR50 Appendix G when the.
core is critical. From the heatup curves, assuming that the RCS is at nominal operating pressure of 2235 psig, the minimum RCS temperature for criticality would be 410 F. Since the proposed Tavg of 530 F is significantly higher than the criticality limit, the required margin between the minimum RTuor and the RCS temperature is maintained. There will be no adverse affect on the reactor vessel for.
startup at Tavg of 530 F.
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The bases for the Action statement is that if the' Tavg of one or more reactor coolant system loops is not within the limit, the limit must be promptly restored, or the plant must be placed in a condition in which the LCO does not apply. This is done by restoring Tavg to 2 530 F within 15 minutes or placing the reactor in Mode 3 within the following 15 minutes. Rapid reactor shutdown can be readily and practically achieved within the allotted time.
Safety Analysis Westinghouse performed the attached Safety Evaluation (Enclosure 2) and determined that the results and conclusions of the Safety Analysis (UFSAR) remain valid if the reactor achieves criticality at 530 F instead of the nominal no-load Tavg of 547 F which was originally assumed in the analysis. Westinghouse determined that five Non-LOCA transients in the UFSAR (Chapter 15) required evaluation for potential effects due to a minimum temperature for criticality of 530 F. They are the following:
1)
Rod Withdrawal From Subcritical (UFSAR 15.4.1) 2)
Rod Ejection (UFSAR 15.4.7) 3)
Zero Power Feedwater Malfunction (UFSAR 15.1.1) 4)
Steamline Break (UFSAR 15.1.5) 5)
Boron Dilution During Startup (UFSAR 151.M The Large Break LOCA (LBLOCA) and Small Break LOCA (SBLOCA) are evaluated at the limiting condition of 102% power. Since the full power Tavg value is unchanged by this proposed amendment, the LBLOCA and SBLOCA analysis are not affected. The long term core cooling and Hot Leg Switchover Analyses do not use Tavg, and therefore, are also not affected.
No effect on the Safety Analysis results from deleting the outdated require.ments in Technical Specifications Section 3.2 as indicated in Attachment B.
These changes are administrative in nature and simply remove requirements which no longer apply. Removal of these specifications will improve the Technical Specifications from a human factors perspective.
Safety Analysis Assumptions Westinghouse made the following assumptions in the evaluations of the Non-3 LOCA transients:
1)
Reactor pressure has been brought up to nominal 2250 psia before l
criticality.
2)
There are no Safety Parameter Interaction List (SPIL) changes as a -
result of reducing the initial temperature (such as the reactivity coefficients, boron concentrations, or peaking factors) 3)
There are no changes to the safety analysis limits due to increased setpoint uncertainty.
4)
The evaluation is valid only for startup operation between 530 F and 547"F. No Mode 1 power operation is assumed in this temperature range.
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The first assumption is consistent with the initial conditions assumed in the
. July 1993 Revision of the UFSAR for the five transients with the exception of the Rod Withdrawal From Suberitical (RWFS) and the Rod Ejection (RE) transients.
Both of these transients are analyzed in the UFSAR from an initial condition of 2210 psia. Westinghouse made the assumption that the Reactor Coolant System is at nominal pressure because this will be true for a normal startup. The automatic Pressurizer Pressure Control System maintains pressure at 2235 psig or the operator maintains pressure with manual control of Pressurizer heaters and sprays.
A normal startup using General Operating Procedure 2, ' Plant Startup', is always conducted at a nominal RCS pressure of 2235 psig (2250 psia).
An initial condition of 2210 psia for the RWFS event generates more severe reaults with respect to DNBR than would be obtained with an initial condition of 2250 psia. Therefore, since the results of the analysis are acceptable with an initial pressure of 2210 psia, in that the DNBR remains greater than the DNBR limit, the RWFS event initiated from 2250 psia will be less severe. The results of the current analysis of record remain bounding for an initial condition of 2250 psia.
An initial condition of 2210 psia for the RE event generates more severe results with respect to DNBR than would he obtained with an initial condition of i
2250 psia. Less DNB would occur in the limiting fuel channels during a RE event initiated from 2250 psia. The current Safety Analysis acceptance criteria also specify that the peak RCS pressure must remain below that which would cause stresses to exceed allowable stress limits. To initiate the event at a higher initial pressure could possibly affect the peak pressure. But, starting from a higher initial pressure will not significantly affect the heat transfer from the fuel or heat generation in the coolant and, consequently, the peak pressure during the RE event will not be significantly affected. Starting from a lower initial temperature (530"F) will decrease the amount of DNB that occurs and will also reduce the pressure peak. The previous discussion is based on the Rod Ejection transient System Overpressure Analysis in the July 1993 update of the UFSAR (Section 15.4.7.4.3).
The second assumption is valid based on the methodology used in performing the SPIL calculations and on the updated evaluation of the Zion Unit 1 and 2 SPIL's based on a temperature for criticality of 530 F.
Hot Zero Power SPIL calculations are performed at the nominal no load temperature. This is appropriate fbr the following reasons:
a.
The worst case assumed temperature for the HZP transient is nominal no-load HZP. It is appro'priate to establish the nuclear design parameters consistent with this condition, b.
Nuclear design establishes neutronic conditions consistent with initial.
conditions assumed in the accident analysis. As nuclear design is modeling pre-event conditions, nominal parameters are typically applied (RCS Temperature at nominal HZP, Pressurizer pressure at 2235 psig, level in its programmed control band, etc.).
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Present nuclear design methods also conservatively ensure 4% trip reactivity
. at HZP and ensure adequate post trip shutdown margin for post trip cooldown to temperatures as low as 525 F (as stated in Enclosure 1, page 5).
Westinghouse evaluated the neutronic portion of the SPIL for Z1C13 for possible impact of a minimum critical temperature of 530 F. No violations were found. Therefore, the previous SPIL remains valid. Enclosure 5 contains the transmittal letter from Westinghouse documenting the results of the SPIL reevaluation. CECO Nuclear Fuel Services (NETS) performed the reevaluation of the neutronic portion of the Z2013 SPIL for possible impact of a minimum critical temperature of 530"F. NETS concluded that there were no violations of the neutronic portion of the Z2C13 SPIL. Therefore, that SPIL remains valid. Enclosure 6 contains the documented results of the NETS reevaluation of the Z2C13 SPIL.
CECO NETS will have administrative controls in place to ensure that all future reloads implementing a 530"F minimum temperature for criticality have the proper neutronic SPIL evaluation. This commitment is being tracked within NETS through the Nuclear Tracking System (NTS).
The third assumption is valid based on the CECO NED analysis of Reactor Trip and ESFAS setpoints at a minimum temperature fbr criticality of 530"F. Refer to Enclosure 4 which concluded that Technical Specification setpoints and Safety Analysis limits are not affected by a minimum temperature for criticality of 530 F.
Increased uncertainty due to lower temperature is accommodated within existing setpoint margins.
The fourth assumption was made due to the applicability of the proposed amendment. The proposed revision to Technical Specification Section 3.2.1.C.2 is applicable to Unit Startup only. Specification 3.2.1.D provides requirements for Reactivity Control & Power Distribution for Power Operation (Mode 1).
Hod Withdrawal From Suberitic_a_1 (UFSAR 15.4.1)
As stated in the UFSAR Section 15.4.1.2, the VANTAGE 5 Reload Transition Safety Report (RTSR), and the attached Westinghouse Safety Evaluation (Attachment 1), the Reactor is assumed to be at hot zero power with the Reactor critical and an average temperature of 547"F. These initial conditions will generate conservative results when compared to the results obtained for a lower initial system average temperature (i.e. 530 F). The higher initial system temperature yields a larger fuel to water heat transfer coefficient, a larget fuel thermal capacity, and a less negative (smaller absolute magnitude) Doppler coefficient. The Doppler feedback effect is the primary mechanism which mitigates the power excursion. At'a higher temperature the less negative Doppler coefficient reduces the Doppler feedback effect, thereby increasing the neutron flux peak during the transient. The high neutron flux peak combined with a high fuel thermal capacity and larger thermal conductivity yield a larger peak heat flux. Starting with a lower average temperature (530 F) will result in increased Doppler power feedback and less thermal conductivity. This will serve to turn the transient around sooner with less severe results. In summary, the results of the RWFS event initiated from 530 F, compared to the event initiated from 547 F, will be less severe. The DNB analysis is inherently less limiting with lower temperatures. The lower temperature also provides less limiting transient neutronics. The conclusions in the UFSAR remain valid.
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r Rod Ejection (UFFAR 15.4.7)
In the analysis of the Rod Ejection transient, the Reactor is assumed to be at hot zero power with the Reactor critical and at an average temperature of 547*F.
These initial conditions will generate limiting resuus when compared to the results obtained for a lower initial system average temperature (i.e. 530"F). As is stated in UFSAR Section 15.4.7.1, the resultant core thermal power excursion is limited by the Doppler reactivity effect of the increased fuel temperature and terminated by a reactor trip actuated by high neutron flux signals. As in the above discussion, the Rod Ejection event initiated from an initial condition of 530"F would be less severe due to greater Doppler power feedback and the lower heat transfer coeflicient. The conclusions in the UFSAR remain valid for the Rod Ejection transient from an initial condition of 530"F.
e 7
Zero Power Fefdwater Malfunction (UFSAR 15.1.1)
The attached Westinghouse Safety Evaluation (Enclosure 2) provides the analysis of the Zero Power Feedwater Malfunction initiated from a temperature of l
530 F. A specific DNB analysis for the event is not performed. Instead, the reactivity insertion rate is calculated for the feedwater malfunction event and compared to the reactivity insertion rate used in the RWFS event. Starting with a lower coolant average temperature will result in a less limiting cooldown, thus the differential tempe dure which drives the reactivity insertion will be lower. The reactivity insertion rate, will in turn, be lower than that for the RWFS event.
Therefore, the conclusions in the UFSAR remain valid for the Feedwater Malfunction event initiated from 530 F.
Steamline Break Core Response (UFSAR 15.1.5)
The attached Westinghouse Safety Evaluation (Enclosure 2) provides the analysis of the Steamline Break Core Response from an initial condition of 530 F.
Starting with a lower coolant temperature will result in a less limiting cooldown and allow the reactor to reach an equilibrium condition sooner. Steam generator blowdown will not be adversely impacted, thus the time that the engineered safety features setpoints are reached will also not be adversely affected. This results in i
about the same amount of heat transfer capability available on the secondary side for heat extraction from the primary side. However, with the primary side at a lower initial average temperature, the driving force for heat transfer between the secondary and primary will be less and an equilibrium condition will be reached sooner. The reactivity coefficients used in the analysis are not constant with temperature but are valid for a range of temperatures from low (~350 F) to no load i
(547 F). Given all the above, the current steamline break core response event is conservative compared to the results that would be obtained assuming a lower initial reactor coolant average temperature of 530* F. Therefore, the conclusions in the UFSAR remain valid.
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The attached Westinghouse Safety Analysis was completed prior to the July
-1993 issuance of the UFSAR. Before July 1993, the protection that was credited for
{
mitigating the consequences of the Steamline Break event (UFSAR 15.1.5.2) was l
F the High Steam Flow Safety Injection. Since the P-12 Permissive setpoint at Zion could allow both High Steamflow Safety Injection signals to be manually blocked
-(below 540 F) during a startup at 530 F, Westinghouse performed the attached l
analysis at 530"F giving credit for protection to the Low Pressurizer Pressure Safety Injection signal. Acceptable results were obtained crediting the Low Pressurizer SI signal. The July 1993 UFSAR does not reflect the MSLB core -
response re-analysis at the lower critical temperature but does list Low Pressurize.
Pressure SI actuation as one of the functions that provides protection against c steamline rupture (Section 15.1.5.2.1). As such, the 1994 update to the UFSAR wil!
include a discussion regarding mitigation of the consequences of the MSLB event -
crediting both the High Steamflow SI and Low Pressurizer Pressure SI signals.
Both High Steam Flow SI signals and the Low Pressurizer Pressure SI signal are available during startup and critical operations. Per Technical Specification Table 3.4-1, the protection is required to be operable in Modes 1 and 2.
Boron Dilution Durine Startun (UFSAR 15.4.5) l The attached Westinghouse Safety Evaluation provides the analysis of the l
Boron Dilution at Startup event initiated at 530 F. A lower zero power average temperature will impact. the Mode 2 Boron Dilution in the calculation of the specific volume of the RCS. All other inputs remain unaffected. The change in differential boron worth at the lower temperature was found to be negligible in that the current SPIL was determined to be bounding at the lower temperature. The lower temperature will decrease the specific volume (assuming no pressure change). With the lower specific volume (higher density), the concentration of boron (ppm).in the coolant will remain the same but the amount (mass per unit volume) of boron in the coolant will increase due to the constant pressurizer level. As a result, diluting at the maximum volumetric rate given in the UFSAR (208 gpm), the actual amount of boron being diluted from the core will be less. Using the lower specific volume, the :
time to loss of shutdown margin will be greater, giving the operator more time to l
recognize and correct the situation. The results of the analysis remain valid for a -
startup from an initial condition of 530 F.
Conclusion The Minimum Temperature for Criticality specified in Technical Specification 3.2.1.C.2 can be changed to 530"F with no adverse affects or irreversible consequences. Bases for the amendment have been established which are consistent with the Standard Technical Specifications with the one exception noted (P-12). The results and conclusions of the UFSAR for LOCA and Non-LOCA transients remain valid based on the fact that the current analysis results are conservative and bounding for reactor criticality at 530"F. An Evaluation of Significant Hazards Considerations is provided in Attachment C.
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ZOSR-062-93 9
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Enclosures:
1)
NFS Interim Evaluation of PI 92-014 Zion Minimum Temperature for Criticality, NFS:PSS:93-053, February 10, 1993 2)
Westinghouse Safety Evaluation SECL 93-002 Rev. 2, July 27,1995 3)
Nuclear Fuel Services Safety Evaluation Review NFS:RSA:93-052\\NFS:PSS:93-126, August 26,1993 4)
Zion Units 1&2 Minimum Temperature for Criticality Review of Reactor Trip and ESFAS Setpoint Error, February 10,1993 5)
Zion Unit 1 Cycle 13 SPIL Neutronic Evaluation for Reduced T '
Critical, February 23,1993 6)
Zion Unit 2 Cycle 13 SPIL Based on Tcrit of 530 F, Z2013/070, t
July 22,1993 e
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F ENCLOSURE 1' NUCLEAR ENGINEERING & TECIINOLOGY SERVICES NUCLEAR FUEL SERVICES INTERIM EVALUATION MINIMUM TEMPERATURE FOR CTIRICALITY i
LICENSE AMENDMENT REQUEST 93-09 t
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T.
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February 10,1993 NFS:PSS:93-053 l
LNIBlM FVAl.UATION OF PI 92-014
.l ZION MINIMUM TCRIT (T.S. 3.2.1 REACTIVITY CONTROL)
Prepared by:
- /M Lonnie K. Kepfey i
Reviewed by:
M all &
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Wallace W. guk Nuclear Fuel Services Department i
Commonwealth Edison Company i
72 W. Adams St., Room 900 Chicago, Illinois 60603 l
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x February 10,1993-NFS:PSS:93-033 A.
BACKGROUND / POTENTIAL ISSUE DESCRIPTION Zion Regulatory Assurance has formally questioned,in the Zion Tech Spec Improvement Project (TSIP), the adequacy of the 500 f Minimum Temperature Technical Specification as the core and safety analysis are typically performed at HZP (547 f with uncertainties). NFS is evaluating and tracking this issue as P192-014.
Technical Specification 3.2.1, " Reactivity Control", does not allow the reactor to be critical when the Moderator Temperature Coefficient (MTC) is more positive than the MTC Tech Spec limit of +7 pcm/ f. During-physics testing, administrative rod withdraw limits (RWLs) are-established to ensure the MTC Tech Spec limit is satisfied. The subject associated minimum temperature for criticality specification is to limit the range and associated uncertainty of these administrative limits.
The concerns are that:
(1) The 300 f specified in Zion Tech Spec 3.2.1 may not have a rigorous technical bases and therefore be inappropriate for the specification; (2) The Tech Specs may allow operation in a temperature region that is not bounded by the present core and transient analysis of record; and (3) As the Westinghouse Standardized Tech Spec bases for minimum temperature for criticality address issues beyond MTC control, which are not addressed in the Zion Tech Spec bases, Zion's Technical Specification may be inadequate.
B.
EVALUATION Each concern identified above is evaluated below:
- 1. MTC Control Zion's Tech Spec 3.2.1 addresses reactivity control. The Tech Spec bases for the requirement is MTC centrol(originally maintaining a negative MTC). The ieth specs were originally structured to control MTC as follows: 1) Immediately prior to startup, the reactor core conditions lTavg, boron concentration, rod position and burnup) required to maintain MTC within the limit are defined; 2) Administrative limits on these core conditions are established as required; and 3) Immediately -
prior to criticality, the reactor coolant temperature is verified greater than the administrative limits (except during physics testing), and verified greater than 300 f. The specification allows CECO to credit and apply administrative limits to meet the negative MTC requirement, but through a reasonable range of application (to 500 f).
2
February 10,1993 NFS:PSS:93-053 -
The coef ficients applied in the administrative procedure use design reactivity coefficients to extrapolate the MTC physics test results (a test at one point) to a family of curves over a wide range of these parameters.
The uncertainty increases as operating conditions (the most significant of which is temperature) deviate from the conditions at which the MTC test was performed (traditionally HZP).
Previous Zion evaluations (see NFS:RSA:93-024) have established a conservative initial operating minimum of 543'f. The initial startup procedure also requires a minimum RCS temperature of 543 f prior to initial criticality. In discussions with Westinghouse, it has been l
determined that margins in the existing analysis methods have been used to justify operating at seven degrees below nominal and is the i
bases for the Westinghouse Standard Technical Specification " Minimum i
Temperature for Criticality" LCO. A 540 f minimum for normal HZP operations is therefore required and applied in the remainder of this evaluation. A concern is that during subsequent plant operating evolutions (i.e. warming the secondary, physics testing, and initial turbine loading, etc.) RCS temperature may drop to lower temperatures.
-[
Recent core designs have not been able to satisfy the MTC limit at I
temperatures much below HZP, even with the credit for rod withdrawal limits. As a result, operation significantly below HZP lat BOLl has been restricted by the rod withdrawallimits. These limits have, for previous cycles, crossed the Tech Spec rod insertion lines at temperatures significantly above 500 f. Later in core life, the unit has significant decay heat which makes it easier to control HZP temperature, and in the event of an increase in secondary heat removal, MTC reactivity feedback
(
will quickly restore RCS temperature.
i With the present positive MTC tech spec limit (+7) and existing core designs, there is ample margin to the MTC limit. Even with a large BOL increase in secondary heat removal, the potential to operate with an MTC more positive than the MTC assumed in the safety analysis is small. The actual concern with MTC control has therefore largely been eliminated.
1 As the reactor can theoretically be critical at temperatures significantly below HZP, and still comply with the MTC Tech Spec, the additional TSIP concerns are addressed below.
==
Conclusion:==
for MTC control the Tech Soee is adeouate and is not a technical or saf ety issue.
i 1
February 10,1993 NFS:PSS:93-053
- 2. Safety Analysis, Core Analysis, and Setpoint Analysis a) Safety Analysis - The preliminary assessment is that PI 92-014 does not constitute a safety issue for two reasons. First, the safety analysis is not required to assume a transient initiated from significantly below the nominal Tavg band (in this case below 540 f ). Secondly, the preliminary evaluation indicates that even if the analysis was required it would not generally lead to results more limiting than the present analysis of record and in all cases would not exceed the present accident analysis acceptance criteria.
Each transient assumes initial operation within normal operating bands with worst case uncertainties -- Temperature on the programmed Tavg band ( 5.5*f), pressure within the Pressurizer pressure control band (2235 40 psi), level within the limits of the' control band (15%
uncertainty), etc. Administrative controls, operating procedures, and control systems establish reasonably conservative initial conditions.
Infrequent operation significantly below the established operating bands is itself a normal operating transient (ANSI Class II), and the analysis does not normally assume in the analysis of Class II, III, or IV events an ongoing Class 11 event as an initial condition. Nevertheless,if the accident analyses were performed at lower initial RCS temperatures, the results are generally not more limiting.
An evaluation is in progress to support continuous, zero power, critical operation to an RCS temperature as low as 530 f. The major analyzed events for HZP are the overpower events: Rod Withdraw From Subcritical (RWFS), Rod Ejection, and Steamline Break. The Zero Power Feedwater Malfunction and Boron Dilution events are also discussed below.
For the " Rod Withdraw from Subcritical," the UFSAR and VANTAGE 5 RTSR states "The reactor is assumed to be just critical at hot zero power (547*F). This assumption is more conservative than that of a Imver initial system temperature." The DNB analysis is inherently less limiting with lower temperature. The lower temperature also provides less limiting transient neutronics.
Much of the " Rod Ejection Accident"is performed generically for all Westinghouse units and should be largely insensitive to a plant's tech spec temperature requirements. The event is extremely rapid (peak flux of approximately 350% of nominalis attained in 0.16 seconds). The maximum power level and local fuel temperatures are so far above initial conditions that the event is insensitive to non-neutronic initial conditions. Similar to RWFS, the lower temperature would result in less limiting transient neutronics. The reduced initial temperature is not expected to adversely affect the peak transient RCS pressure.
4
i 3
February 10,1993 NFS:PSS:93-053 The "Steamline Break" analysis maximizes the transient cool down, and the assumed re craticality. As such, the results are largely insensitive to initial temperature. Sensitivity studies indicate that the event initiated from a lower temperature will typically have a smaller cool down and should be less limiting.
The "Zero Power Feedwater Malfunction" is also impacted by reduced temperature operations; however, this event applies the bounding reactivity insertion from the RWFS transient, and applies a bounding most negative MTC. The change in initial RCS and steam generator mass do not significantly affect the results.
The Boron Dilution at Startup event is not adversely impacted by the initial temperature change.
b) Neutronics analysis - Most HZP Safety Parameter Interaction List-(SPIL) calculations are performed at the nominal no load temperature.
This is appropriate as follows:
i
- 1. The worst case assumed temperature for the HZP transient is l
nominal no load - HZP. It is appropriate to establish the nuclear design parameters consistent with this condition.
2.
Nuclear design establishes neutronic conditions consistent i
with the initial conditions assumed in the accident analysis. As nuclear design is modeling pre-event conditions, nominal-parameters are typically applied [RCS temperature at nominal a
HZP, Pressurizer pressure at 2235 psig,levelin its programmed i
control band, etc.l.
Present nuclear design methods also conservatively ensure 4% trip reactivity at HZP and ensure adequate post-trip shutdown margin for post-trip cool down to temperatures as low as 525af, Also note that the preliminary nuclear design SPIL cvaluation to support continuous operation to 530 f demonstrates all SPIL parameters are satisfied down to 525*f.
c) The Setpoint analysis -- The EAGLE-21 NIS setpoint analysis does not explicitly address critical operation below HZP; however, the HFP setpoint analysis should be bounding over zero power operations to a temperature as low as 530 f. The necessary revision to the setpoint analysis to confirm that the impact can be accommodated within i
existing analysis margins is being developed. The preliminary results confirm that extending the setpoint analysis to address critical operation as low as 530*f is accommodated within the present analysis setpoints and allowable values. Though preliminary, this conclusion is t
I 5
February 10,1993
^
NFS:PSS:93-053 appropriate for application in this subject evaluation for determination of safety significance.
Westinghouse has provided a preliminary evaluation of the impact of reduced temperatures on the protection functions. There is no impact to the High range trip, and the impact on the Low range trip is small and can be accommodated within existing setpoint margins. In addition, for an initial startup after refueling, the increased setpoint uncertainty
~i from reduced RCS temperature effects is secondary to the known uncertainties associated with the significant " flux-to-power" relationship change which occurs from the end of the previous cycle to the beginning of the next and the large inherent uncertainty at zero power from not being able to calibrate against a calorimetric until the plant reaches the power range. The uncertainty is knmvn and apprapriate precautions in an initial startup are in place.
The key compensatory action is reducing the high flux trip setpoint: for zero power events, the Low Range trip provides a conservative trip function when compared to the High Range trip. The HZP " Rod Fjection" and " Rod Withdraw from Subcritical" events take credit for the Low Range, High Flux Trip which is not typically reduced. This is acceptable because the initial startup procedure specifies a boron concentration adequate to ensure the reactor is subcritical with all control rods withdrawn. Subsequently, a dilution to critical from a near all rods out condition is performed. This significantly reduces the potential for Rod Ejection and Rod Withdraw From Subcritical events. More important is that the events are insensitive to trip setpoint error. The rise in nuclear flux is so rapid that 1) the effect of errors in the trip setpoint on the actual time at which the rods begin to fall is negligible, and 2) the power q
excursion is mitigated by Doppler reactivity feedback prior to the trip.
The analysis assumes the reactor trip function terminates the event, but after the peak nuclear power has been attained and reduced.
in summary, for an initial startup after refueling, the Low Range trip provides conservative protection with respect to the High Range Trip, the uncertainty in the Low Range trip is small and can be accommodated in existing setpoint margins. The uncertainty introduced from a reduced RCS temperature is secondary to the existing uncertainty in the trip from the power calorimetric and EOL/BOL flux change effects are presently addressed by existing compensatory startup provisions and trip setpoint sensitivities.
For startups performed after the power ascension from refueling has been completed, the NIS setpoint uncertainty is less than during the initial startup because a NIS calibration has been completed.
Westinghouse has determined that the Low Range trip, once scaled to full power (Tcold of 530 f),is conservative for zero power reduced temperature operations above 530 f.
6 l
February 10,1993 NFS:PSS:93-053 The NIS Low Range trip is the primary trip protection for overpower and overcooling events from HZP. Westinghouse in its preliminary-evaluation of continuous reduced temperature operation has also identified potential impacts to the NIS Intermediate Range trip, NIS l
Source Range trip, Feedwater Flow, High Steam Flow, Steam Flow / Feed Flow Mismatch, Steam Flow in Two Steamlines - High, Steam Generator Water Level, and Pressurizer Water Level. The preliminary results of the i
evaluation for continuous low power operation shows the increased uncertainty is accommodated within existing setpoint margins.
t
==
Conclusion:==
NFS does not consider transient zero nower oneration at reduced temnerature(continuous oneration above 540 f and transient cool down to 530 f) a Reactivity Manauement or safety issue thitt reouires additional compensatory actions.
- 3. Westinghouse Standardized Tech Spec Bases I
The bases section for Westinghouse Standard Tech Specs (7_lon has custom Specs), for minimum temperature for criticality, states that this specification is required to ensure:
- 1. The moderator temperature coefficient is within its analyzed temperature range [ addressed abovel;
- 2. The trip instrumentation is within its normal operating range:
- 3. The Pressurizer is capable of being within its normal operating range; NDT emperature;.-
- 4. The reactor vesselis above its minimum RT t
and i
- 5. The plant is above the cool down permissive, P-12 INote: allows 1
Si on High Steam Flow" to be blocked below 540 fl.
j r
These are Tech Specs that ensure plant operation within nominal design parameters. It is a subset of like parameters [RCS temperature, Pressurizer pressure, Pressurizer level, steam generator pressure, steam j
generator level, etc.l that define normal operating conditions which
'j become the initial conditions for transient operations. The parameters are assured of being in their operating range by Tech Specs, control systems, or other documents (UFSAR, plant procedures, etc.).
Zion's narrow range temperature indication is on scale at 530 f. The preliminary evaluation of Westinghouse setpoints described above ensures trip instrumentation is within its analyzed range at temperatures above 530 f.
7
February 10,1993 NFS:PSS:93-053 Technical Specification 3.3.1.d ensures the Pressurizer is operable prior to criticality.
Technical Specification 3.3.2 ensures the reactor vessel is above its minimum RTNDT emperature; and t
Tech Spec Table 3.4.1 requires the unit to be borated to shutdown conditions prior to blocking Safety injection on High Steam How. The wording of Tech Spec Table 3.4.1 may not be adequate to ensure SI actuation on liigh Steamflow is available prior to criticality. This i
evaluation requires that RCS temperature be above 540 f (P-12) prior to.
initial criticality. The issue must be addressed in the review and I
approval of the future evaluation for continuous operation above 530 f.
The standard Minimum Temperature for Criticality specification and the Zion specification differ in the selection of the limit and what remedial actions are required should the limit be exceeded. The standard Tech Spec limit is established at the bottom end of the normal operating band.
in a transient, Standard Tech Specs allow 15 minutes to restore temperature above the minimum level, or be in Hot Standby within an additional 15 minutes, but do not specify a lower limit for the allowable range of the transient (the only limit for off-nominal operation is a second, lower temperature limit, for physics testing). The Zion Tech Spec establishes a limit significantly below the normal operating temperature (established by administrative procedures) that cannot be exceeded, and provides no allowable period of deviation even in physics testing. These administrative level of assurance questions (the appropriate limit, its objectives, and the appropriateTechnical..
Specification required remedial actions) should be addressed by the station.
Omelusion: The nresent Tech Snecs adeouately address the issues presented in the Standard Tech Snec bases with two excentions. One. a Minimum Temnerature for Criticality Snecification limit of 530*f is annronriate to ensure trin instrurnentation is in within its analyzed rance. Two. Si on Steam Flow must be enabled nrior to criticality. This could be addressed by imnroved wordinc in either Tech Snec Table 3.42 or Tech Snec 3.2.1. a conservative tech snec internretation. or a P setnoirit chance. The station should address the administrative level of assurance ouestions described above. In the interim. until the evaluation allowine continuous oneration at reduced temocrature is finalized and annroved. should the RCS temnerature fall below 540 f the.
temnerature should be restored to above 540*f in a reasonable neriod.
1 Should temocrature fall below 530. it should be restored to 530 f within 15 minutes. A rnanual reactor trin. and the annronriate operatinn action snecified in the operatina Reactor Trin nrocedure for nost-trin reduced temnerature imnact on shutdown marcin 13 nnm/ f below 5251 should 8
February 10,1993 NFS:PSS:93-053 be initiated if the 13 minute action reouirement is not satisfied. or RCS temnerature (Tavn) falls belmv the nresent 500of Technical Snecification.
This evaluation is continnent upon havine the protection function
" Safety iniection sLXteam Flow in two Steamlines" enabled whenever the' reactor is critical.
i C.
DETERMINATION OF NO UNREVIEWED SAFETY OUESTIONS As demonstrated in the technical evalyation above, PI 92-014 will not produce any unreviewed safety questions per 10CFR50.59. Specifically:
a) Does the PI increase the probability of occurrence of an accident previously evaluated in the safety analysis report?
P192-014 does not involve an increase in the probability of occurrence of an accident previously evaluated in the safety analysis report. The PI does not change any fuel, core, or plant design requirement and does not affect any other plant system or component. The increased uncertainty in the protection setpoints is small and can be accommodated by existing setpoint margin.
b) Does PI 92-014 increase the consequences of an accident previously evaluated in the safety analysis report?
PI 92-014 does not involve an increase in the consequences of an accident previously evaluated in the safety analysis report. An increase in secondary heat removalis addressed in the present safety analyses of record. The increased uncertainty in the protection setpoints is small and preliminary information indicates it is accommodated within -
existing setpoint margin. Therefore the licensing basis acceptance criteria for all analysis that could be impacted by PI-014 are satisfied.
As such, the conclusions and therefore the consequences of accidents previously analyzed remain unchanged, c) Does PI 92-014 increase the probability of a malfunction of equipment important to safety which has been previously evaluated in the safety analysis report?
PI 92-014 does not involve an increase in the probability of a malfunction of equipment important to safety which has been previously analyzed in the safety analysis report, as modified and approved by the VANTAGE 5 submittal. The impact of P192-014 on core reactivity control, the accident analysis, core analysis, and setpoint analysis have been addressed in this evaluation. The Pl does not impact the design or operation of any other plant equipment important to sa fety.
9
February 10,1993 NFS:PSS:93-053 d) Does PI 92-014 increase the consequences of a malfunction of equipment important to safety which has been previously evaluated in the safety analysis report?
1 P192 014 does not involve an increase in the consequences of a malfunction of equipment important to safety which has been previously analyzed in the safety analysis report as modified and approved in the VANTAGE 5 submittal. The impact of PI 92-014 on core reactivity control, the accident analysis, core analysis, and setpoint analysis have been evaluated. The Pi does not adversely impact the design or operation of any other plant equipment. As such, the accident -
l analysis, including the analysis of the consequences of malfunctions of equipment important to safety, is unchanged.
c) Does PI 92-014 create the possibility for an accident of a different type than any previously evaluated in the safety analysis report?
PI 92-014 does not create the possibility of a new or different kind of accident from any accident previously evaluated. This is based on the fact that the method and manner of reactor and plant operation is unchanged.
O Does PI 92-014 create the possibility for a different type of malfunction of equipment important to safety from any previously evaluated in the safety analysis report?
i P192-014 does not create the pcssibility of a new or different type of a malfunction of equipment important to safety from any previously
+
analyzed in the safety analysis report as modified and approved by the VANTAGE 5 submittal. The increased setpoint uncertainty is absorbed within existing setpoint margins. The PI does not change any other fuel, core, or plant design requirement and does not affect any other plant 3
system or component.
g) Does PI 92-014 reduce the margin of safety as defined in the basis for any Technical Specification?
PI 92-014 does not involve a reduction in a margin of safety. This i
evaluation maintains a normal operating temperature range consistent with the existing accident analysis. In addition, the preliminary results from the ongoing evaluation of lower temperature critical operation required to close PI 92-014 indicate that the Zion safety analysis bounds zero power critical operation at reduced temperatures (above 530 f).
r Existing plant operating procedures and administrative restrictions established in this evaluation ensure these analysis assumptions are satisfied. Adequate shutdown margin is ensured down to 530 f. The l
core initial conditions are consistent with current licensing basis design criteria and therefore maintain Technical Specification margins.
t 10
February 10.1993 NFS:PSS:93-053 '
D, CONCLUSIONS I
The preceding evaluation demonstrates that startup and operation of Zion Units I & 2 in Cycle 13 while the analysis to close PI 92-014 is being finalized poses no unreviewed safety questions. This is contingent upon
]
operating temperature restrictions defined above.
1 j
u
' l
'l 1
-i i
i 11 i
.i
.l 1)
.I 1
ENCLOSURE 2
)
.WESTINGIIOUSE SAFETY EVALUATION SECL 93-002 i
REV. 2 MINIMUM TEMPERATURE FOR CTIRICALITY LICENSE AMENDMENT REQUEST 93-09 i
i e
f f
f I
b E
ww
-i,,-
m f,_ ~^'
s L
Westinghouse Energy Systems Electric Corporation 3:" 355
"*F ""**a n22n355 CWE-93-161 ET-NSL-OPL-1-93-392 July 27,1993 Mr. D. L. Shamblin Commonwealth Edison Company 1400 Opus Place, Suite 500 Downers Grove, IL 60515 Commonwealth Edison Company Zion Units I and 2 Minimum Temoerature for Criticality - Revised SECL
Dear Mr. Shamblin:
Enclosed please find safety evaluation SECL 93-002 Rev. 2, " Minimum Temp which demonstrates that operation of Zion Units 1 and 2 with a minimum temp 530' F does not constitute an unreviewed safety question or require a change to the Zio specifications. The revised safety evaluation identifies the changed sections. The conclus original safety evaluation remain unchanged.
If you have any questions, do not hesitate to call.
Very truly yours, b
B. S. Humphries, Manager Commonwealth Edison Projects Enclosure Customer Projects Department B. E. Rarig/cid ec:
M. Lohmann T. Joyce W. Kuk J. Ballard K. Noms J. Johnson CLDeh4 BEL A
SECL NO.93-002 REV. 2 Customer Reference No(s).
Westinghouse Reference No(s).
WESTINGHOUSE NUCLEAR SAFETY SAFETY EVALUATION CHECK LIST (SECL) 1.) NUCLEAR PLANT (S):
Zion Units 1 and 2 2.) SUBJECT (TITLE):
Minimum Temocrature for Criticality 3.) The written safety evaluation of the revised procedure, design change or modification require 10CFR50.59(b) has been prepared to the extent required and is attached. If a safety evaluation is not required or is incomplete for any reason, explain on Page 2.
Parts A and B of this Safety Evaluation Check List are to be completed only on the basis of the safety evaluation performed.
CHECK LIST - PART A - 10CFR50.59(a)(1) 3.1) Yes 1 No _ A change to the plant as described in the FSAR?
3.2) Yes _ No 1 A change to procedures as described in the FSAR7 3.3) Yes _ No 1 A test or experiment not described in the FSAR7 3.4) Yes _ No 1 A change to the plant technical specifications?
(See Note on Page 2.)
4.) CHECK LIST - PART B - 10CFR50.59(a)(2) (Justification for Part B answers must be include on page 2.)
4.1) Yes _ No 1 Will the probability of an accident previously evaluated in the FSAR be increased?
1 4.2) Yes _ No 1 Will the consequences of an accident previously evaluated in the FSAR be increased?
4.3) Yes _ No 1 May the possibility of an accident which is different than any already evaluated in the FSAR be created?
t 4.4) Yes _ No 1 Will the probability of a malfunction of equipment important to safety i
previously evaluated in the FSAR be increased?
i 4.5) Yes _ No 1 Will the consequences of a malfunction of equipment importart to safety previously evaluated in the FSAR be increased?
4.6) Yes _ No 1 May the possibility of a malfunction of equipment impcrtant to safety different than any already evaluated in the FSAR be created?
4.7) Yes _ No 1 Will the margin of safety as described in the bases to any technical specification be reduced?
Page1
SECL 93-002 REV.2 NOTES:
If the answer to any of the above questions is unknown, indicate under 5.) REMARKS and explain below.
if the answer to any of the above questions in Part A (3.4) or Part B cannot be answered in the negative, based on written safety evaluation, the change review would require an application fo license amendment as required by 10CFR50.59(c) and submitted to the NRC pursuant to 10CFR50.90.
5.) REMARKS The answers given in Section 3, Part A, and Section 4, Part B, of the Safety Evaluation Checklist, are based on the attached Safety Evaluation.
FOR UFSAR UPDATE Section:
Pages:
Tables:
Figures:
6.) SAFETY EVALUATION APPROVAL LADDER:
6.1) Prepared by (Nuclear Safety):
M Date:
7-2C. - %
B. E. Rarig 6 6.2) Nuclear Safety Group Manager:
C10ith Date: 7-21o ' D '
}.)d. I.ivingsthn Page 2 axm.
\\
SECL 93-002 Rev. 2 i
ZION UNITS 1 & 2 MINIMUM TEMPERATURE FOR CRITICALITY 3
i SAFETY EVALUATION
1.0 BACKGROUND
Zion Technical Specification Section 3.2.1-C specifies that immediately accident analyses use a zero-load temperature of Zion Unit 2 Cycle 13, Commonwealth Edison has requested that the Zion lic accidents analyzed at zero-load conditions (zero-load temperature of 547'F) be eva address the impact of assuming the plant anamt criticality at 530*F.
The bases section of Standard Tech Specs (Zion has custom specs) for mim criticality, states that this limitation is required to ensure:
- 1. He moderator temperature coefficient is within its analyzed temperature range.
- 2. He trip instrumentation is within its normal operating range.
- 3. The pressurizer is capable of being in an OPERABLE status with a steam bubble.
- 4. He reactor vessel is above its mimmum RT,or temperamre.
- 5. He plant is above the cooldown steam dump permissive, P-12.
For the analyses performed to support the VANTAGE 5 fuel upgrade at Zion Un positive MTC limit (+7) was assumed. As a result, the potential for operation with an more positive than the MTC assumed in the safety analysis is small.
Trip instrumentation will be outside of its normal operating temperature range w an2mm criticality at 530*F. The effect on the instrumentation uncertainty has been evalu is discussed in Section 3.2.
The effect on the pressurizer and reactor vessel at the reduced temperature for c been evaluated, as described in Section 3.4.
He P-12 permissive is an automatic enabling signal for SI on High Steam Flow a during a starmp (increasing power) condition. Section 3.1.6 addresses the unav side signal for SI acmation in the temperature range 530*F to 540*F, and the use of the L Pressurizer Pressure signal for SI actuation in this teraperature range.
2.0 LICENSING BASIS Zion Technical Specification Section 3.2.1-C (2.b) specifies that immMi*1y prior to reactor coolant temperature shall be shown to be greater than 500*F.
Chapter 15 of the Zion FSAR describes the accidents which are analyzed at zero-load con (Modes 2-3). These events are potentially impacted by a reduction in the muumum t for criticality from 547'F to 530*F:
i m,.
Page 3
SECL 93-002 Rev. 2 Rod Withdrawal from Suberitical (FSAR 15.4.1)
Rod Ejection (Zero Power)
(FSAR 15.4.1)
Zero Power Feedwater Malfunction(FSAR 15.1.1)
Steamline Break Core Response (FSAR 15.1.5)
Boron Dilution Mode 2 - Startup (FSAR 15.4.5) 3.0 EVALUATION 3.1 Non-LOCA Evaluation ne following assumptions are used in the evaluations:
The reactor pressure has been brought up to nominal (2250 psia) b a.
b.
There are no SPIL changes as a result of reducing the initial temperature reactivity coefficients, boron concentrations, or peakmg factors).
There are no changes to the safety analysis limjts due to increased c.
d.
His evaluation is only valid for Mode 2 (zero power) operation between 530' 547'F. No Mode I power operation is assumed in the temperamre range.
3.1.1 Rod Withdrawal from Suberitical (FSAR 15.4.1)
In the analysis of the rod withdrawal from suberitical event the reactor is assum zero power with the reactor critical and at an average temperamre of 547'F. These initial conditions will generate conservative results when compared to the results obtained initial system average temperature (i.e., 530'F). He initial temperature of 547'F wh compared to 530*F gives a larger fuel to water heat transfer coefficient, larger the water and a less negative Doppler coefficient, all of which reduce the Dopp feedback. Rus, starting with a lower average temperature at 530'F will result in in Doppler power feedback. His will serve to turn the transient around sooner with le results. Therefore the conclusions in the FSAR remain valid.
3.1.2 Rod Ejection (Zero Power),(FSAR 15.4.1)
In the analysis of the rod ejection (zero power) event the reactor is assumed to be power with the reactor critical and at an average temperature of 547'F. These initial con will generate conservative results when compared to the results obtained for a lower in symm average temperature (i.e.,530*F). De initial temperature of 547'F when compare 530'F gives a larger fuel to water heat transfer coefficient, larger heat capacity of t a less negative Doppler coefficient, all of which reduce the Doppler power feedba starting with a lower average temperamre at 530*F will result in increased Doppler feedback. This will serve to turn the transient around sooner with le the conclusions in the FSAR remain valid.
Page 4
SECL 93 002 Rev. 2 3.1.3 Zero Power Feedwater Malfunction (FSAR 15.1.1)
A specific DNB analysis for the zero power feedwater event is not reactivity insertion rate used in the RWFS event. reactiv vent and compared to the the RWFS event will be limiting. Starting with a lower cIf the rate is lower, the in a less limiting cooldown, thus the differential temperature whi h d conclusions in the FSAR remain validrate will be lower. The re c
. Therefore the 3.1.4 Steamline Break Core Response (FSAR 15.1.5)
Starting with a lower coolant average temperamre will result in a le not be adversely impacted, thus the time the ng cooldoivn and r owdown will also not adversely affected; His restJts in about the same amoun n s are reached is available on the secondary side for hea'. extraction from the primary er capability the secondary and primary will be less and an eq
. However, with the e etween n t on will be reached sooner.
His event is analyzed as a cooldown event, thus heat transfer a steam generator tubes remain covered. While y for a longer s ong as the greater mass inventory on the secondary side, this additional mass will result in transfer capability reduced long before the conservative model in h and heat temperature of 530*F will not significantly impact t e analysis has predicted a average results.
He reactivity coefficients used in the analysis (stuck rod coefficie temperamre but are valid for a range of temperatures from low n with Given the above, the current steamline break core response eve results that would be obtained assuming a lower initial reactor c pared to the 530'F. Therefore the conclusions in the FSAR remain valid.
ra ure of In the zero power steamline break core response calculation, SIis coincident with Low Steam Pressure. He block of P-12 at tempera:u g
team Flow blocks SI actuation from the normally credited secondary side signa so criticality of 540*F (which bounds the cooler tem ow, low secondary side signal. While normally the accumulators would inject b rom a leg as the RCS pressure dropped below the accumulator actuation pre e cold
,1-that the Westinghouse model was not conservative with regard to the Zi as a concern accumulator configuration. Derefore, as a conservative approach, the acc on plant-specific actuation from the Low Pressurizer Pressure sig re not um Page 5
SECL 93-002 Rev. 2 3.1.5 Boron Dilution (Mode 2 - Startup) (FSAR 15.4.5)
A lower zero power average temperature will impact the Mode 2 boro of the specific volume of the RCS. All other inputs (i.e., boron worth u at on The lower temperature will decrease the specific volume (assuming Using this new lower specific volume, the time from alarm to loss of s re).
calculated, will be greater. Thus the lower initial average temperature conditions. Derefore the conclusions of the FSAR remain valid.
3.1.6 P-12 Permissive The P-12 permissive is an automatic enabling si SIis actuated by High Steam Flow coincident wit temperatures below 540*F also blocks SI actuation from the normally credited secon minimum temperature of criticality of 530'F and ary side The reanalysis credited S1 actuation from the Low Pressyrizer Pressu results were obtamed. Therefore, the P-12 pernussive which enables the High coincident with Low-Low Tavg or Low Steam Pressure signal is not re steamline break core response accident in the temperature range of 530 licensing basis accident analyzed in Mode 2.
3.2 Setpoint Effects and Evaluation Westinghouse evaluated the effects of zero power operation (criticality) at tem 530*F on protection function PMA terms. He results of Westinghouse's evaluation transmitted to Commonwealth Edison in Reference 1. For some functions critica temperature introduces conservative errors which need not be accounted for
. For other functions, additional, or revised errors were identified which should be included i protection function uncertamry calculations.
Commonwealth Edison reviewed the revised PMA uncertamty terms identified they apply to the CECO setpoint accuracy calculations. Based on this review CE that the Zion Technical Specification setpoints and safety analysis limits are n reduction in mmimum critical temperature to 530*F. Reference 2 contains the results of Commonwealth Edison's evaluation.
3.3 SPIL Evaluation Commonwealth Edison has evaluated the neutronic portion of the Zion 2 Cycle 13 S contained in Refercoce 3 for possible impact of the revised mmimum critical temperatu 530'F. No violations were found. Therefore, the redesigned SPIL remamt valid. Referen documents the results of Commonwealth Edison's evaluation.
Page 6
SECL 93@2 Rev. 2 impact of the ~ revised minimum critical temperature for possible Westinghouse's evaluation.Derefore, the redesigned SPIL remain u s of The neutronic-only SPIL evaluation performed for Cycle 13 operat Terit = 530'F for future cycles. Herefore, a reload specific SPIL ev o necessarily bound performed for Terit = 530'F.
s ould be 3.4 Mechanical Evaluation 500*F prior to startup. However, the current bad temperature of 547'F, which plant operators anticipate will s uses a zero-Therefore, an evaluation is being conducted to support startup of Z r or to startup.
530 *F.
a Tavg of Westinghouse has performed evaluations to providejust@ cation that fo reactor startup:
vg = 530*F at
- The pressurizer is capable of being in an OPERABLE stams w
- Re reactor vessel is above its mmimum RT m.
water level less than 92% in operating Modes 1, ac ty and a the steam bubble in the pressurizer is typically drawn in Mode 5 He pr. D of about 325 psi. With saturated water condition essurizer heaters are essure increased, with the decreasing water level in the pressurizer allowing steam bubble in the upper portion of the pressurizer.
rming the As the RCS pressure and temperature increases, using pump heat from pumps, the pressurizer is maintamed at saturation conditions. When RCS ant psig, the pressurizer temperature is the saturation temperature of a temperature (Tavg) of the rest of the RCS. As lottg as the system operating of reactor startup remsim unchanged, i.e. 2235 psig, the operating t ressure at the time will also remam unchanged, and will not be affected by Tavg. Rus startu will have no effect on the operating conditions or the operability of the pre avg r up.
changes in the RCS during heatup are limited pe Appendix G (Reference 7), and 10CFR50, Appendix G (Reference 8) are based on the fracture toughness properties of
. He requirements are addition to RCS temperature and pressure limits relative to systems heamp
. In requirements also specify the mmimum criticality curve, which provide l
Page 7
SECL 93-002 Rev. 2 and pressure conditions allowed by 10CFR50, Appendix G when the core is critical. From t Zion heatup curves, assuming that the RCS is at normal operating pressure of 2235 psi minimum RCS temperature for criticality would be about 410*F. Since the proposed 530'F for reactor startup is significantly greater than the criticality limit, the required between the minimum RTm and the RCS temperature is maintained, nere would be no adverse affect on the reactor vessel for startup at Tavg = 530*F.
3.5 LOCA Evaluation Large Break and Small Break LOCA are performed at limiting conditions of 102% of Since the Tavg value previously supported for full power (561.4*F with +5/-10 uncerta conditions, the LBLOCA and SBLOCA analyses are not affected. He longterm core coo and hot leg switchover analyses do not use Tavg, and therefore, are not affected.
4.0 DETERMINATION OF UNREVIEWED SAFETY OUESTION 1.
Will the probability of an accident previously evalugted in the FSAR be increased?
No. The revised mimmum temperamre for criticality will not introduce a new accident initiator that could change the probability of a given transient to occur, herefore, the probability of accidents previously evaluated in the FSAR will not be increased.
2.
Will the consequences of an accident previously evaluated in the FSAR be increased?
No. He safety evaluation demonstrates that all applicable acceptance criteria will be met with the revised mmimum temperature of criticality. Additionally, no new limiting single failure is introduced by the increased mmimum temperature. Herefore, there is not potential for an increase in the doses.
3.
May the possibility of an accident which is different than any already evaluated in the FSAR be created?
No. A revised mmimum temperature of criticality will not introduce a new accident initiator mechanism. Dus no new accident will be created.
4.
Will the probability of a malfunction of equipment important to safety previously evaluated in the FSAR be increased?
No. A revised mmimum temperamre of criticality will not affect the design or operation of plant equipment important to safety. Herefore, the probability of a malfunction of equipment in.portant to safety will not be increased.
Page 8 m
)
SECL 93 002 Rev. 2 I
5.
Will the consequences of a malfunction of equipment important of safety previously
,I evaluated in the FSAR be increased?
No. The revised minimum temperature of criticality is such that malfunction ofim equipment is not affected. Therefore, no increase in the consequences already evaluated i FSAR will result. There is no possibility of increasing the doses as a result of oper the revised minimum temperature of criticality.
6.
May the possibility of a malfunction of equipment important to safety different than already evaluated in the FSAR be created?
No new failure mode has been cr2ated as a result of the revised mimmum temperature,of criticality. Therefore, the possibility of a malfunction of important safety equipment no evaluated in the FSAR has not been created.
7.
Will the margin of safety as defined in the BASES to any technical specification be reduced No, as discussed in the attached safety evaluation, the revised mmimum temperature o will not invalidate any of the conclusions presented in the FSAR accident analyses. For all the FSAR accidents, the DNB design basis, primary and secondary pressure limits, and dose release limits continue to be met. Thus, there is no reduction in the margin to safety.,
5.0 CONCLUSION
He safety analyses continue to be conservative with respect to a lower zero power average temperature (530*F). The conclusions in the FSAR remam valid. Startup at a lower Tavg will have no effect on the operating conditions or the operability of the pressurizer or the reactor vessel.
6.0 REFERENCE
- 1) Westinghouse letter CWE-93-116 "PMA Terms for Reduced Minimum Critical Temperamre of $30*F", dated February 9,1993.
- 2) Commonwealth Edison letter CHRON # 198051, " Zion Units 1 & 2 Minimum Temperature for Criticality, Review of Reactor Trip and ESFAS Setpoint Error", dated February 10, 1993.
- 3) Westinghouse letter 92CW*-G-0111, " Zion Unit 2 Cycle 13 Redesigned SPIL Transmittal",
dated October 23,1992.
- 4) Commonwealth Edison letter Z2Cl3/070, " Updated Evaluation of Zion Unit 2 Cycle 13 SPIL Based on T of 530*F", dated July 22,1993.
1
- 5) Westinghouse letter 93CW*-G-0023. " Zion Unit 1 Cycle 13 SPIL Neutronic Evaluation for Reduced T
, dated February 23,1993.
Page 9
==
SECL 93-002 Rev. 2
- 6) Zion Station Technical Specifications. Section 3.3.1.D.
- Pressurizer."
7)
ASME Boiler & Pressure Vessel Code. Appendix G, " Protection Against Nonductile Failure.*
- 8) 10CFR50, Appendix G, " Fracture Toughness Requirements.*
9)
Zion Station Technical Specifications, Section 3.3.2, " Pressurization and S e
t Page 10
j ENCLOSURE 3 i
NUCLEAR FUEL SERVICES SAFETY EVALUATION REVIEW 8/26/93 MINIMUM TEMPERATURE FOR CRITICALITY LICENSE AMENDMENT REQUEST 93-09 t
P s
August 26,1993 NFS:RSA:93-052 SEP 011993 NFS:PSS:93-126 I
Mr. AnthonyBroccolo
Subject:
NFS Review of Westinghouse's Zion Minimum Temperature for Criticality Safety Evaluation
References:
1)
Westinghouse Ltr, CWE-93-161, ET-NSL-OPL-I-93-392, from B. S. Humphries to D. L. Shamblin, " Zion Units 1 and 2 Minimum Temperature for Criticality - Revised SECL", dated July 27,1993.
2)
Westinghouse Ltr, CWE-93-131, ET-NSL-OPL-I-93-173, from B. S. Humphries to D. L. Shamblin, "Ziori Units I and 2 Minimum Temperature for Criticality - SECL Transmittal", dated March 30,1993.
3)
Westinghouse Ltr, CWE-93-116, ET-NSL-OPL-I-93-069, from B. S. Humphries to Dr. T. Rieck, "PMA Terms for Reduced Minimum Critical Temperature of 530 *F", dated February 9,1993.
4)
CECO Ltr, ENC Chron No. 198051, from G. Wagner to W. Poirier, " Zion Units 1 & 2 Minimum Temperature for Criticality Review of Reactor Trip and ESFAS Setpoint Error", dated February 10,1993.
5)
Westinghouse Ltr, 93CW'-G-0023, from M. J. Weber to Dr. T. Rieck,
" Zion Unit 1 Cycle 13 SPIL Neutronic Evaluation for Reduced Terit", dated February 23,1993.
6)
NFS Ltr, Z2Cl3/070, from R. J. Chin to M. J. Weber, " Updated Evaluation of the Zion Unit 2 Cycle 13 SPIL Based on Tcrit of 530 *F", dated July 22, 1993.
7)
NFS Ltr, NFS:RSA:92-157, J. E. Ballard to R. A. Chrzanowski, " Zion Minimum Critical Temperature", dated December 21,1992.
8)
Zion Technical Specifications through Amendment 147 (DPR-39) and Amendment 135 (DPR-48), Specification 3.2.1.C.2.b.
9)
NFS Ltr, NFS:RSA:93-059, J. E. Ballard to M. Rauckhorst, " Zion Increased FW Isolation Time Evaluation", dated May 6,1993.
August 26,1993 NFS:RSA:93-052 NFS:PSS:93-126 9
NFS has reviewed the revised Westinghouse (E) Minimum Temperature for Criticality Safety Evaluation (Reference 1/ Attachment B) which addresses the resulting impact of assum that Zion achieves criticality and initiates Mode 2 operation with the Reactar Coolant System (RCS) 2530 *F. The original Westinghouse safety evaluation (Reference 2) was issued in March 1993 but required revision due to an incorrect assumption regarding the Safety Injection (SI) logic available for Main Steam Line Break (MSLB) protection with the low Tavg permissive P-12 actuated. This evaluation provides an assessment which includes the effects on the Zion licen basis accident analyses V toi protection functions based on a minimum critical temperature (Tcrit) range from 530 *
- F.
It was determined that the current safety analysis results, reactor protection setpoints, the pressurizer operability and the RTmyr temperature requirements all remain valid for a reduced critical and zero load (Mode 2) temperature of 530 F. This means that zero load operation in Mode 2, per Zion Technical Specificatior alid for an indicated temperature between 530 *F to 547 F based on narrow range RTD ii
,ntation. In addition, prior to rolling the turbine, reactor heatup during Mode 2, no load operation using nuclear heating from 530 F to the nominal Tavg of 547 F is permitted. Power operation in Mode 1 is assumed to occur only with Tavg 2 543 *F.
Further, based on the information provided in Table 3.4-1 of the Zion Technical Specifications with regard to SI actuation signals, it is NFS' interpretation that SI actuation initiated from secondary side trip funct'
-hall remain available when approaching criticality and subsequently for no load operation duri.
reactor start-up. To that end, in order to be consistent and to meet the intended requirement delineated in the Tech Specs, Zion Station must verify that the SI actuation on High Steam Line Flow is not blocked below the P-12 permissive prior to start-up.
Details of NFS' review are included as Attachment A. Ifyou have any questions on the NFS review, please contact Wallace Kuk (G. O. ext 3855) or Jerry Ballard (G. O. ext 3839) of i
my staff.
i Mk fs errance A. Rieck Nuclear Fuel Services Manager W
TAR:WWK:pc ep.wwornuma l
i
August 26,1993 NFS:RSA:93-052 i
NFS:PSS:93-126 Attachments cc:
K. A. Ainger S. A. Kaplan R. A. Chrzanowski T. W. Simpkin R. J. Niederer J.R. Johnson-J. E. Ballard A. W. Wong D. R. Redden K. W. Norris B. L. Manges TAR Directs PSS File - Zion Min. Tcrit PSS-CF NFS-CF k
i l
1
August 26,1993 NFS:RSA:93-052 NFS:PSS:93-126 Attachment A
- 1. BACKGROUND Section 3.2.1.C.2.b of the Zion Technical Specifications specifies that, prior to a unit startup, the reactor coolant temperature shall be verified to be greater than 500 *F; ho~ wever, the zero-load (Modes 2 and 3) transients analyzed in the Zion UFSAR, Chapter 15, assume an initial RCS temperature of 547 'F. The uncertainty evaluations for setpoints could not be extrapolated to temperatures below 540 F. To resolve this Tech Spec inconsistency, E and CECO /NFS performed a rigorous evaluation to determine the impact of critical operation between 547 F and 530 F. Zion informed NFS that 530 'F was an acceptable temperature for operational purposes since the selection of 530 'F to be the minimum critical temperature (Tcrit) is partly due to the narrrow range RTD Tavg response span of 530 *F to 630 'F.
- 2. EVALUATION
SUMMARY
The areas evaluated include the initial conditions and input assumptions for the safety analysis, the reactor protection setpoint uncertainty efTects, and mechanical evaluations all based on a minimum Tcrit of 530 *F. Each of these areas is summanzed below:
A. Reactor Protection Setooint Evaluation The E evaluation identified which reactor protection setpoints were affected by a minimum Tcrit of 530 'F. E then generated revised Process Measurement Accuracy (PMA) terms [ based on reduced RCS temperature effects] which are used in setpoint uncertainty calculations. The revised setpoint PMA results were documented in Reference 3. CECO's Electrical / Instrumentation and Control Engineering (El&C) group ra-=%h+ed the maximum reactor protection setpoint uncertainty values using the revised PMA terms EI&C has determined that the revised setpoint uncertainties did not exceed the maximum total allowance and the existing Zion Tech Spec reactor protection setpoints and the safety analysis limits were not affected by a minimum Tcrit of 530 "F. The El&C's setpoint results were documented in Reference 4.
B. Core Reload Neutronic Safety Parameter Interaction List (SPIL) Review W performed the ZlCl3 core reload neutronic SPIL evaluation and NFS pedormed the Z2Cl3 SPIL evaluation for the potential effects due to the revised minimum Tcrit of 530 *F. The evaluations determined that all SPIL limits were still met and both Zion core reload designs remained valid. It should be noted that Z2Cl3 had an 1
August 26,1993 NFS:RSA:93-052 NFS:PSS:93-126 updated evaluation performed for the Main Steamline Break transient. This re-evaluation was based on the new statepoints generated for the minimum Tcrit condition of 530 F and the revised feedwater isolation time of 90 seconds (Reference 9). Reference document the E ZlCl3 and the NFS Z2Cl3 SPIL evaluations, respectively.
The core reload neutronic SPIL evaluations, performed for Cycle 13, Units 1 & 2, cannot be verified as bounding for all subsequent core reload cycles for a minimum Terit of 530 *F. Therefore, a cycle specific SPIL evaluation is required for each future core -
reload which credits a minimum Tcrit of 530 F. Administrative controls will be placed in the applicable NFS reload procedures to ensure that all future core reload cycles implementing a 530 *F minimum Tcrit have the proper neutronic SPIL evaluation. This administrative change to the design procedures, to perform the required additional evaluation at 530 *F, will be tracked within NFS through NTS (# 901-150-93-00404).
C. Non-LOCA Evaluation E determined that the fol. lowing Zion UFSAR, Chapter 15 transients required evaluation for potential effects due to a minimum Tcrit of 530 *F (Reference 1):
- 1. Rod Withdrawal from Subcritical (UFSAR 15.4.1)
- 2. Rod Ejection at HZP (UFSAR 15.4.7)
- 4. Steamline Break Core Response (UFSAR 15.1.5)
The E evaluation assumed that all core reload SPIL limits were still valid, the existing UFSAR safety analysis limits were not affected by any increased setpoint uncertainty, and that the RCS pressure was increased to the nominal operating 2235 psig value before criticality. Additionally, E specified that the evaluation was only applicable for Mode 2 transients between 530 *F and 547 *F, as no Mode I power operation was assumed in this temperature range.
E concluded that for each transient evaluated above, the combination of the initial conditions and reactivity feedback effects associated with a lower Tcrit of 530 F would generate less severe results than at the currently assumed 547 "F temperature. Thus, the Zion non-LOCA UFSAR results remain valid and bounding for a minimum Terit of 530 *F.
D. LOCA Evaluation f
The Large Break LOCA (LBLOCA) and Small Break LOCA (SBLOCA) are evaluated at the limiting condition of 102% power. Since the full power Tavg value is 2
August 26,1993 NFS:RSA:93-052 NFS:PSS:93-126 unchanged, the LBLOCA and SBLOCA analyses are not affected. The long term core cooling and hot leg switch over analyses do not use Tavg, and therefore, are also not affected.
E. P-12 Permissive Consideration In the original safety evaluation (Reference 2), Westinghouse analyzed the minimum critical temperature Main Steam Line Break (MSLB) core response assuming that the P-12 permissive allows manual blocking of the High Steam Flow coincident with the Lo-Lo Tavg SI signal. CECO determined however that the P-12 permissive allows manual blocking ofboth the High Steam Flow coincident with Lo-Lo Tavg and the High Steam Flow coincident with Low Steam Line Pressure SI signals when Tavg is less than 540 *F. Consequently, a minimum Tcrit of 530 F could allow Mode 2 operation with both these SI functions still manually blocked. W reanalyzed the zero power (MSLB) core response at 540 *F (which bounds the cooler temperature of 530 F), crediting only the Low Pressurizer Pressure SI signal, and determined this case, which credited an alternate SI signal, still generates acceptable results. Therefore, the P-12 permissive does not affect the minimum critical temperature of 530 *F.
F. Pressudzer Ooerability for Reactor Startup During startup, when RCS pressure reaches the nominal operating pressure of 2235 psig, the Pressurizer temperature is maintained at the corresponding saturation temperature of about 653 *F, independent of the RCS Tavg. As long as the RCS operating pressure during reactor startup remains unchanged, the Pressurizer operating temperature will also remain unchanged and is therefore, not affected by Tavg. Thus, reactor startup at a lower Tavg of 530 *F will have no effect on Pressurizer operability.
G. Reactor Vessel Above Minimum RTET Assuming that the RCS is at the nominal operating pressure of 2235 psig, and based on the Zion heatup curves, the minimum RCS temperature allowed for criticality would be about 410 *F. Since the esaluated minimum startup temperature of 530 F is significantly greater than 410 *F, the required margin between the minimum RTET and the RCS temperature is maintained. There will be no adverse effect on the reactor vessel for reactor startup at 530 *F.
3
August 26,1993 NFS:RSA:93-052 NFS:PSS:93-126
- 3. UFSAR UPDATES UFSAR Section 15.1.5.2 should be updated to brieDy discuss that the MSLB core response event has been analyzed to a minimum critical temperature of 530 F. This case credits the Low Pressurizer Pressure SI protection function since the P-12 Permissive logic actuated below 540 F allows manual block of both the High Steam Flow coincident with Lo-Lo Tavg and High Steam Flow coincident with Low Steam Line Pressure SI functions which are available for the HZP 547'F case. However, the MSLB event analyzed at the HZP 547 F conditions continues to be the most limiting case so these are still the applicable results as presented in the current UFSAR Section 15.1.5.2. This UFSAR brief update will be provided by NFS and is being tracked through NTS (# 930-150-93-00401.)
- 4. CONCLUSIONS NFS concurs with the assumptions t.nd methods used in the W safety evaluation for a minimum Terit of 530 *F (Reference 1). The current safety analysis results, reabtor protection setpoints, the pressurizer operability and the RTmyr temperature requirements all remain valid for a reduced critical and zero load (Mode 2) temperature of 530 F. Reactor heatup during Mode 2 operation using nuclear heating from 530 'F to the nominal Tavg of 547 F is also permitted.
However, Mode 1 operation is assumed to occur with Tavg 2 543 'F based on the previous NFS evaluation of the RCS temperature uncertainty (Reference 7).
There are currently two outstanding issues regarding the evaluation of a minimum Tcrit of 530 F. The first issue is that all future cycles currently require a reload specific neutronic SPIL evaluation to be performed for zero power conditions at 530 F. This additional evaluation is tracked administratively within NFS through NTS (# 901-150-93-00404). Second, Section i
3.2.1.C.2.b of the Zion Tech Specs must be revised to be consistent with the actual design basis minimum Tcrit of 530 F instead of the currently licensed Tcrit 2 500 'F. This commitment is tracked by the NLA as NTS# 930-150-93-00403.
1 l
l 4
1 ENCLOSURE 4 l
]
ZION UNITS 1&2 REVIEW OF REACTOR TRIP AND ESFAS SETPOINT ERROR FEBRUARY 10,1993 1
1 MINIMUM TEMPERATURE FOR CRITICALITY LICENSE AMENDMENT REQUEST 93-09 l
I I
i l
)
)
~
n
/
\\ Comm:n::calth Edis:n t
C J 1400 Opus Placs
( C J Downers Grove, filinois 6o515
'O February 10,1993 in reply refer to CHRON#
19805.$
Mr. W. Poirier, Project Manager Commonwealth Edison Projects Westinghouse Electric Corporation PO Box 355 Pittsburgh, PA 15230-0355 i
Subject:
Zion Units 1 & 2 Minimum Temperature for Criticality Review of Reactor Trip and ESFAS Setpoint Error
Reference:
1)
Westinghouse letter, J.A. Mitchell to K.W. Norris, (CECO), Zion Minimum Temperature for Criticality, RFQ No. 9315019 CECO Function #592281208 2)
CWE-93-116, PMA terms for Reduced Minimum Critical Temperature of 530'F, 2/9/93.
3)
WCAP 12523, Bases Document for Wastinghouse Setpoint Methodology for Protection Systems, Commonwealth Edison Company Zion / Byron /Braidwood Units, October 1990 As part of the analysis required to support startup of the Zion stations at a Tavg j
equal to 530 degrees ' %atinghouse identified and calculated the affected uncertainty terms for each of the following protection functions:
1.
NIS Reactor Trips (Power, Intermediate, Source Ranges) 2.
Steam Generator Level (all) 3.
Steam Flow 4.
Feedwater Flow 5.
Pressurizer Level Functions NED has reviewed the revised uncertainty terms identified in reference 2 as they apply to the CECO setpoint accuracy calculations. Based on this review, CECO concludes that the following Tech Spec setpoints and safety analysis limits are not effected by the reduction in minimum critical temperature to 530 degrees F. The following setpoints were reviewed:
1.
Power Range Neutron Flux, Low 2.
Power Range Neutron Flux, P-8 Permissive 3.
Intermediate Range Neutron Flux i
February 10,1993 Page 2 4.
Source Range Neutron Flux 5.
Steam Flow /Feedwater Flow Mismatch 6.
Steam Flow in Two Steamlines High 7.
Steam Generator Narrow Range Water Level, Low-Low, Low and High-High 8.
Pressurizer Water Level High The attached table lists the setpoint values and Safety Analysis Limits supported by the CECO calculations and are provided for completion of the minimum critical temperature review by Westinghouse.
if you have further questions, please call Pete Wicyk at (708)663 7264.
M Pete VandeVisse~
Instrument Engineer A
YW l&C Supe [rvisor Pete Wicy Engineering and Construction
(,
[ G. Wagner Electrical /l&C. Design Supervisor Attachments cc:
R. Tuetken T. Joyce (S. Kaplan M. Lohmann T. Rieck J. Ballard S. Stimac NEDCC
v ATTACHMENT 2/10/93-ZION MINIMUM TEMPERATURE FOR CRITICALITY Page 1 The following setpoints and Safety Analysis Limits are supported by CECO calculations.
Setooint CECO Setooint Value Safety Analysis Limit Calculation
- 2. Power Range NED I-EIC-0063 28% RTP Neutron Flux, P-8 not used in safety analysis Permissive
- 3. Intermediate NED-I EIC-0062 25% RTP Range Neutron Flux not used in safety analysis
- 4. Source Range NED-1-EIC-0061 1E+ 5 cps not used in safety Neutron Flux analysis
- 5. Steam NED-I EIC-0059 60% steam flow not used in safety Flow /Feedwater analysis Flow Mismatch
- 6. Steam Flow in NED-I-ElC-OO59 NOTE 1 NOTE 2 Two Steamlines High
- 7. S/G Narrow NED-I-ElC-0055 Low-Low Low-Low Range Water Level 10% level 0 % level Low Low not used in 25% level safety analysis High High 70% level 80% level
- 8. Pressurizer NED-I-EIC-0056 92% level not used in safety Water Level High analysis NOTE 1: Setpoint - A delta P corresponding to 40% of full steam flow between 0% ar d 20%
load and then a delta P increasing linearly to a delta P corresponding to 110% of full steam flow at full load.
NOTE 2: Safety Analysis Limit - A delta P corresponding to 56% of full steam flow between 0%
and 20% load and then a delta P increasing linearly to a delta P corresponding to 120% of full steam flow at full load. Since the span of the steam flow channel at 100% load is 120% steam flow the Safety Analysis Limit is limited to 120% steam flow
1 ATTACHMENT ZION MINIMUM TEMPERATURE FOR CRITICALITY 2/10/93 Page 2 Evaluation of the Channel Statistical Allowances and setpoint margins.
1.
Power Range Neutron Flux, Low Per reference 2, the revised PMA term for operation at a minimum critical temperature of 530 degrees F is less than the currently evaluated PMA term.
Therefore the existing setpoint is conservative with respect to the Zio 2.
Power Range Neutron Flux, P 8 Permissive Per reference 2, the revised Power Range Neutron Flux PMA term for minimum critical temperature of 530 degrees F is less than the currently PMA term. Therefore the existing permissive setpoint is conservative w to the Zion Tech Specs.
3.
Intermediate Range Neutron Flux Per reference 2, the revised Intermediate Range PMA term for operation at a minimum critical temperature of 530 degrees F is the same type, magnitud direction as the revised Power Range PMA term and consequently less tha currently evaluated PMA term. Therefore the existing setpoint is conservative w respect to the Zion Tech Specs.
4.
Source Range Neutron Flux For reference 2, the revised Source Range PMA term due to operation at a riinimum critical temperature of 530 degrees F increases fromM ipan. The combination of this term, the control rod movement PMA term and the detector burn up PMA term are still bounded by the 10% PMA term used in the current channel accuracy calculation. Therefore the existing setpoint is conservative with respect to the Zion Tech Specs.
5.
Steam Flow /Feedwater Flow Mismatch Per reference 2, the smallest conservative error for steam flow is larger than th largest non-conservative error for feed flow. Therefore the revised Steam Flow /Feedwater Flow PMA term for operation at a minimum critical temperatu 530 degrees F is less than the currently evaluated PMA term. Therefore the existing setpoint is conservative with respect to the Zion Tech Specs.
i 1
ATTACHMENT 2/10/93 ZION MINIMUM TEMPERATURE FOR CRITICALITY Page 3 I
6.
Steam Flow in Two Steamlines High Per reference 2, the indicated steam flow is greater than the actual steam the revised Steam Flow PMA term for operation at a minimum critical tempe of 530 degrees F is less than the currently evaluated PMA term. Therefore the existing setpoint is conservative with respect to the Zion Tech Specs.
7.
Steam Generator Narrow Range Water Level, Low Low The random PMA term has been replaced by four individual PMA errors treat biases: Process Pressure, Reference Leg Temperature, Fluid Velocity Effects and Downcomer Subcooling.
Per reference 2, the revised PMA terms are:
Process Pressure
= +0.9% span Reference Lag Temperature
= + 0.9% span Fluid Velocity Effect
= -0.1 % span Downcomer Subcooling
= + 0.6 % span t
The Fluid Velocity Effect is in the conservative direction for this setpoint and can ignored.
The revised channel statistical error (CSA) for this setpoint is determined by:
CSA =
IPEA8 + (SCA + SMTE + SD)2 + SPE2 + STE8 + (RCA + RMT RCSA + RD): + RTE'l* + BIAS, + BIAS,o + BIASeve + BIASos Using the values in NED-I-EIC-0055 and the bias terms above as applied to a decreasing trip:
CSA = 10' + (0.50 + 0.44 + 1.10)2 + 1.302 + 0.842 + (0.20 + 0.31 +
0.25):
8
+ 0.25 )" + 0.9 '+ 0.9 + 0 + 0.6 = 5.08% span Using a Total Allowance of 10% span, the margin remains positive.
Margin = TA - CSA = 10% 5.08% = +4.9% span The existing setpoint is conservative with respect to the Zion Tech Specs.
I e
e
ATTACHMENT 2/10/93 ZION MINIMUM TEMPERATURE FOR CRITICALITY Page 4 8.
Steam Generator Narrow Range Water Level, Low The random PMA term has been replaced by four individual PMA errors treated as biases: Process Pressure, Reference Leg Temperature. Fluid Velocity Effects and Downcomer Subcooling.
Por reference 2, the revised PMA terms are:
Process Pressure
= -0.1 % span Reference Lag Temperature
= + 0.9% span Fluic' /elocity Effect
= -0.1 % span Downcomer Subcooling
= + 0.6% span The Process Pressure and Fluid Velocity Effects are in the conservative direction for this setpoint and can be ignored.
The revised channel statistical error (CSA) for this setpoint is determined by:
CSA = [ PEA 2 + (SCA + SMTE + SD)8 + SPE2 + STE +'(RCA + RMTE +
2 RCSA + RD)8 + RTE ]" + BIASn, + BIASc + BIASm + BIASos 2
l Using the values in NED-I-EIC-0055 and the bias terms above as applied to a decreasing trip:
CSA =
102 + (0.50 + 0.44 + 1.10)' + 1.308 + 0.842 + (0.20 + 0.31 + 0 +
0.25)2 + 0.2515 + 0 + 0.9 + 0 + 0.6 = 4.18% span 2
Using a Total Allowance of 25% span, the margin remains positive.
i Margin = TA - CSA = 25% - 4.18 % = + 20.8% span The existing setpoint is conservative with respect to the Zion Tech Specs.
9.
Steam Generator Narrow Range Water Level, High-High The random PMA term has been replaced by four individual PMA errors treated as biases: Process Pressure, Reference Leg Temperature, Fluid Velocity Effects and Downcomer Subcooling.
Per reference 2, the revised PMA terms are:
Process Pressure
= 2.2% span Reference Lag Temperature
= -0.4% span Fluid Velocity Effect
= -0.1 % span Downcomer Subcooling
= + 0.5% span i
I
ATTACHMENT 2/10/93 ZION MINIMUM TEMPERATURE FOR CRITICALITY Page 5 The Downcomer Subcooling effect is in the conservative direction for this setpoint and can be ignored.
The revised channel statistical error (CSA) for this setpoint is determined by:
CSA = (PEA + (SCA + SMTE + SD)2 + SPE2 + STE2 + (RCA + RMTE +
2 RCSA + RD) + RTE 21" + BIAS, + BIASu + BIASm + BIASo, Using the values in NED-1-EIC-0055 and the bias terms above as applied to an increasing trip:
CSA =
(08 + (0.50 + 0.44 + 1.10)2 1.308 + 0.842 + (0.20 + 0.31 + 0 +
+
0.25)2 + 0.2515 + 2.2 + 0.4 + 0.1 + 0 = 5.38% span 2
Using a Total Allowance of 10% span, the margin remains positive.
Margin = 10% - 5.38% = +4.6% span The existing setpoint is conservative with respect to the Zion Tech Specs.
(
B.
Pressurizer Water Level High The random PMA term has been replaced by three individual PMA errors: Process Pressure, Reference Leg Temperature and Subcooling.
Per reference 2, the revised PMA terms are:
Process Pressure 1.2% span
=
Reference Leg Temperature
= -0.7% span Subcooling
+ 7.2% span
=
The Process Pressure effect due to temperature is in the conservative direction for this setpoint and can be ignored. The Process Pressure error shown is due to pressure changes and assumes that the operator does not significantly adjust the reference pressure in the high direction.
The Subcooling effect is in the conservative direction for this setpoint and can be ignored.
The revised channel statistical error (CSA) for this setpoint is determined by:
CSA =
[PMA,,,2 + PEA 8 + (SCA + SMTE + SD)2 + SPE' + STE2 + (RCA +
RMTE + RCSA + RD)2 + RTE 1* + BIASc + BIAS.
8
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ATTACHMENT ZION MINIMUM TEMPERATURE FOR CRITICAUT 2/10/93 Page 6 '
Using the values in NED-I EIC-0055 and the terms above a trip:
e to an increasing CSA =
11.22 + O2 + (0.50 + 0.61 + 2.20)2 + 4.722 +
+ 0 + 0.25)2 + 0.25 2}' + 0.7 + 0 = 6 83 %
1.502 + (0.20 + 0.31 span Using a Total Allowance of 8% span, the margin remains positiv Margin = TA - CSA = 8 % - 6.83% = + 1.1 % span The existing setpoint is conservative with respect to the Zion T pecs.
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l ENCLOSURE 5 ZION UNIT 1 CYCLE 13 SPIL NEUTRONIC EVALUATION FEBRUARY 23,1993 MINIMUM TEMPERATURE FOR CRITICALITY LICENSE AMENDMENT REQUEST 93 09 i
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1 Weatnghouse CommercialNuclear Pona m Bactrtc Carporation FuelDivisica
- "r wsrau en ca February 23, 1993 93CW'-G-0023 Dr. T. Rieck, Manager Commonwealth Edison Company Nuclear Fuel Services Room 900, Edison Building 125 South Clark Street Chicago, ILL 60603 Dear Dr. Rieck CWSODEWEALTE EDISON CC3tPANY SION NUCLEAR PONRK PLANT UNIT 1 CYCLE 13 SPIL MEUTRONIC EVALUATION FOR BEDUCED 2 CRITICAL The neutronic portion of the SPIL for 21C13 was. evaluated for possible impact of a minimum critical.testperature of 530 Degrees Fahrenheit.
No violations were found and, therefore, the E1C13 sPIL previousl Please coratact m or Ed Pulver (y transmitted remains valid.if you have further questions. yse 412) 374-2187) sincerely, Michael Neber Project Engineer Domestic Sales & Customer Projects NJW mid cca D. Ishakaylo E. 5. Young A. Wong J. R. Johnson - 51on Station P. McBale - X Byron
- 8. Sishop - 3 Braidwood J. A. Johnson - 3 Eion t
W. W. Kuk L. K. Keploy D. Beddingfield - X Chicago Sales C. Saaney - M Downers Grove
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ENCLOSURE G ZION UNIT 2 CYCLE 13 SPIL NEUTRONIC EVALUATION JULY 22,1993 MINIMUM TEMPERATURE FOR CRITICALITY i
LICENSE AMENDMENT REQUEST 03-09
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+4-July 22,1993 Z2C13/070 Mr. Michaelj. Weber Westinghouse Electric Corporation Nucieat Fuel Division P.O. Box 355 Pittsburgh, PA 15230
Subject:
Updated Evaluation of the Zion Unit 2 Cycle 13 SPIL Based on Terit of 530T.
Referer:te: 1. " Zion Unit 2 Cycle 13 Redesigned SPIL Transmittal",
J. W. Swogger to T. Kleck, 92CW* G 0111, Dated October 23,1992.
- 2. " Evaluation of Zion Unit 2 Cycle 13 SPIL Based on Tcrit of 5304', R.J. Chin to M.J. Weber, Z2C13/061, Dated February 11,1993.
- 3.
- Revised Steamline Break Transmittal-Zion Unit 2, Cycle 13*,
R.J. Chin to M.J. Weber, Z2Cl 3/069. Dated June 25,1993.
- 4. " Zion Unit 2 Cycle 13 Steamline Break DNBR Analysis *,
M.J. Weber to T. A. Rieck,93CW* C 0084, Dated July 22,1993.
Edison has completed the reevaluation of the neutronic portion of the Z2C13 SPil for the possible impact of using a minimum critical temperature of 5301. This reevaluation was necessitated by the fact that a new Steamline Break statepoint for the minimum critical temperature of 5301 statepoint was generated. This new minimum critical temperature of 5304 statepoint was simultaneously evaluated along with the new statepoint generated for the increased feedwater Isolation time.
The data for the most limiting core configuration and statepoint was transmitted to Westinghouse in Reference 3 for evaluation The results of this evaluation, transmitted to Edison in Reference 4. showed that there was no violation of the Steamline Break accident. This transmittal updated Reference 2 to include the new steamline break evaluation for a minimum critical temperature of 5304. No violations of the neutronic portion of the Z2C13 SPIL were found. Therefore, the redesigned SPIL in Reference 1 remains valid.
Should you have any questions or comments, please contact Brian Manges at (312) 394 3872 or Annie Wong at (312) 394-3843.
Sincerely, M.N O re.a Ronald J. Chin Nuclear Design Supervisor Nuclear Fuel Services RIC:B1.M:cs Z2C13. doc cc:
E. F.Pulver bec:
E.H. Young PND File 10.6 Z2C13 s1 K N. Kovar/J. E. Ballard NFS-CF l
R.J. Niederer
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