ML20063H190
| ML20063H190 | |
| Person / Time | |
|---|---|
| Site: | Wolf Creek  |
| Issue date: | 02/07/1994 |
| From: | Hagan R WOLF CREEK NUCLEAR OPERATING CORP. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML20063H193 | List: |
| References | |
| NA-94-0018, NA-94-18, NUDOCS 9402170150 | |
| Download: ML20063H190 (19) | |
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I W@ NUCLEAR LF CREEK OPERATING CO Robert C. Hagan Vice President Nuclear Assurance February 7, 1994 NA 94-0018 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Mail Station F1-137 Washington, D. C. 20555
Reference:
Subject:
Docket No. 50-482: Proposed Revision to Technical Specifications to Revise Power nerate Parameters Gentlemen Reft te 1 proposed revisions to Facility Operating License No. NPF-42 for Wolf Creek Jenerating Station -(WCGS) and various technical specification -sections to allow plant operation at 3565 MWt (versus the original license limitation of 3411 MWt) and to implement aThot reduction program. This request was approved by the NRC and issued as Amendment No. 69 to the WCGS Operating License and Technical Specifications (Reference 2) . .This amendment included'a Thot Reduction. limit of 5 degrees F, and was implemented in December of 1993.
This letter transmits an application for amendment to the WCGS Operating License to change the 5 degree F Thot reduction limitation to allow operation with.a Thot
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reduction of between 0 to 5 degrees F. This ' license . amendment ' request proposes revisions to rerate parameters in various technical specification sections needed to support the requested Thot limit reduction. During power ascension to 3565 MWt, WCGS was unable to achieve 3565 MWt before the turbine control valves reached a valve wide open condition. The proposed decrease of the 5 degree F Thot reduction limit is designed to increase power output to the .100 percent power level while still maintaining some degree of temperature reduction for steam generator life extension. WC 7C requests that this application for amendment be reviewed in an expediticas manner, since this ac;. ion is needed to allow' WCGS to -
increase power output to the maximum allowed by the operating license. WCNOC has determined that this is the only feasible method to increase power output until design changes can be implemented during the nea, refueling outage.
Attachment I provides a safety evaluation including a' description of the proposed change. Attachment II provides. a no significant ~ hazards consideration determination and Attachment III provides an environmental impact determination.
The specific changes to the technical specifications proposed by this request are provided in Attachment IV.
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^ ' B "" # 8urungt n. KS 66839 / Phone: (316) 364-8831 9402170150 940207 D
PCR ADOCK 05000482 An Equal Opporturuty Employer M/F/HC/ VET P PDR f ') '
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"*NA 94-0018
- Page 2 of 2 In accordance with 10 CFR 50.91, a copy of this application, with attachments, is being provided to the designated Kansas State Official. This proposed revision to the WCGS technical specifications wil] be fully implemented within 60 days of-formal Nuclear Regulatory Commission approval.
If you have any questions concerning this matter, .please. contact me a t --
(316) 364-8831, extension 4553, or Mr. Kevin J. Moles, at extension 4565.
Very truly yours, f
/
7 Ro ert C. Hagan Vice President Nuclear Assurance .
RCH/jra Attachments I - Safety Evaluation II - No Significant Hazards Consideration Determination III - Environmental Impact Determination IV - Proposed Technical Specification Change cc: G. W. Allen (KDHE), w/a L. J. Callan (NRC), w/a G. A. Pick (NRC), w/a W. D. Reckley (NRC), w/a L. A. Yandell (hTC) , w/a
STATE OF KANSAS )
) SS COUNTY OF COFFEY )
Robert C. Hagan, of lawful age, being first duly sworn upon oath says that he is Vice President Nuclear Assurance of Wolf Creek Nuclear Operating Corporation; that he has read the foregoing document and knows the content thereof; that he has executed that same for and on behalf of said Corporation with full power and authority to do so; and that the facts therein stated are true and correct to the best of his knowledge, information and belief.
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',,p g gC Nuclear Assurance SUBSGIBED and sworn to before me this 7 day of diupb , 1994, m
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. Notary Public Q, Expiration Date 5 //1/95
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. Attachment I to NA 94-0018 Page 1 of 11 ATTACHMENT I SAFETY EVALUATION i
. Attachment I to NA 94-0018 Page 2 of 11 Safety Evaluation Proposed change This license amendment request proposes to revise the technical specifications to allow plant operation at a core power of 3565 megawatts thermal (MWt) and a hot leg temperature (Thot) between 613.2 *F and 618.2
- F .. This hot leg tempervture represents an increase in Thot from the current value of 613.2 *F. The vessel average temperature (Tavg) will also correspondingly increase from the current value of 581.2 *F up to the proposed maximum value of 586.5 *F. This proposed license amendment rem.ves the limitation that Wolf Creek Generating Station (WCGS) operate with the 5 *F Thot reduction limit requested in our January 5, 1993 submittal (Letter NA 93-0001, from R. C. Hagan, WCNOC, to the USNRC) .
Background / System Dageription The WCGS power rerate program was performed in order to incrGase the reactor thermal power by 4.5%, from 3411 MWt to 3565 MWt. Along with the power rerate, plant operation changes included a 5 "F Thot reduction to provide for steam generator life extension. Plant system changes (i.e., setpoint sce. ling and Tavg program changes) for power rerate were performed at 95% of 3411 MWt. However, during the power ascension to 3565 MWt, WCGS was unable to achieve 35G5 MWt before-the turbine control valves reached a valves wide open condition.
The proposed decrease of the 5'F Thot reduction is designed to increase power output to the 100 percent power level while still maintaining some degree of temperature reduction for steam generator life extension. WCNOC plans to implement the proposed change by incrementally increasing Thot from its current value (613.2 "F) up to the proposed maximum limit (618.2 'F) until 100 percent power is reached. The following evaluation provides justification for Cycle 7 operation at 3565 MWt with a Thot reduction of between 0 and 5 *F.
The range of Thot values presented in this document are justified based on the original safety analyses performed at the lower temperatures for Cycle 7, combined with the additional safety analyses performed at the higher temperatures for. Cycle
- 7. WCNOC does not intend for this submittal to provide justification for general operation with a range of allowable temperatures in Cycles 8 and beyond.
The technical specification parameter changes necessary for increasing Reactor Coolant System (RCS) temperature are provided as Attachment IV and are summarized in Table 1 below. This table provides an explanation for each change as well as the current and revised values for each parameter.
. . Attachment I to NA 94-0018 Page 3 of 11 Table 1 Summary of Technical Specifications Changes Technical Specification Page Description Of Change Reason For Change Table 2.2-1 2-4 Overtemperature AT Total OTAT setpoint optimized with Item 7 Allowance (TA) reduced from 7.9 assumed full power AT and to 7.0, Z increased from 4.61 to revised f1(AI) .
4.86, and Sensor Error (S) reduced from 2.57 to 1.67 Table 2.2-1 2-8 Overtemperature AT T' increased T' is increased to reflect the Note 1 from 581.2 *F to 586.5 "F new average RCS temperature.
Table 2.2-1 2-8 Overtemperature AT f1(AI) f1(AI) positive breakpoint Note 1 positive breakpoint reduced from reduced based on the new
+7 % AI to +5 % AI maneuvering analysis for the increased Tavg-Table 2.2-1 2-8 Overtemperature AT f1(AI) f1 ( AI) positive slope increased Note 1 positive slope increased from based on the new maneuvering 1.384 % AT / % AI to 1.56 % AT / analysis for the increased
% AI Tavg-Table 2.2-1 2-8 Overtemperature AT Allowable OTAT setpoint optimized with Note 2 Value (AV) reduced from 2.5 % AT assumed full pNer AT and Span to 1.8 % AT Span revised f1(AI).
Table 2.2-1 2-4 Overpower AT Total Allowance OPAT setpoint optimized with Item 8 (TA) reduced from 5.0 to 4.6, Z assumed full power AT.
reduced from 2.15 to 2.02, and Sensor Error (S) reduced from 0.15 to 0.14 Table 2.2-1 2-10 overpower AT T increased from T is increased to reflect the Note 3 581.2 *F to 586.5 *F new average RCS temperature.
Table 2.2-1 2-10 Overpower AT Allowable Value OPAT setpoint optimized with Note 4 (AV) reduced from 2.8 % AT Span assumed full power AT.
to 2.6 % AT Span Table 3.2-1 3/4 Indicated RCS T avg increased Increased to reflect the new 2-16 from 585 *F to 590.5 *F average RCS temperature.
3/4.2.5 B 3/4 " analytical limits of 587.7 0F" Increased to reflect the new Bases 2-3 changed to " analytical limits of average RCS temperature. (Tavg 593.0 *F" Limit = Full Power Tavg + 6.5 "F) 3/4.2.5 B 3/4 " indicated T avg value of 585 *F" (T avg Limit = Full Power Tavg +
Bases 2-3 changed to " indicated T avg value 4
- F) ; Tavg Limit < SA T avg -
of 590.5 *F" Channel Error; (593.0 *F -
1.66 *F)
. . Attachment I to NA 94-0018 Page,4 of 11 Safety Evaluation Analyses and evaluations were performed to assess the impact of 3565 MWt power operation with a Thot reduction of between O' and 5 'F , It was concluded that this change could be implemented with minimal impact to the Updated Safety Analysis Report (USAR) analyses and to the technical specifications. This safety evaluation supports tne temperature change through evaluations that include the following:
A. Evaluation of the impact of the proposed changes on the core thermal limits, including revision of the OPAT and OTAT safety analysis limit trip setpoints for protection of the core limit lines.
B. Evaluation of the impact of the proposed changes on the core thermal-hydraulic analyses.
C. Evaluation of the impact of the proposed changes on the core design, including the core maneuvering analysis.
D. Evaluation of the impact of the proposed changes on all LOCA and non-LOCA events, including re-analysis of the Loss of Flow event.
E. Evaluation of the proposed changes on the Containment analyses.
A discussion of each of these items is provided below. Table 2 presents the revised operating parameters proposed for the remainder of Cycle 7. Other items not specifically cited in this safety evaluation have been reviewed and have been found to be bounded by the evaluations performed for de January 5, 1993 submittal (Letter NA 93-0001, from R. C. Hagan, WCNOC to the USNRC). These values are compared in Table 2 to the current range of temperatures analyzed for the power rerate.
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. Attachment I to NA 94-0018 Page 5 of 11 Table 2 Proposed Rerate Operating Parameters Current Proposed Upper Lower Condition Condition Bound Bound 5 *F Thot 0 *F Thot 1.8 *F Thot 15
- F Thot Parameter Reduction Reduction Ingrgggg Reduction NSSS Power, MWt 3579 3579 3579 3579 Reactor Power, MWt 3565 3565 3565 3565 Thermal Design Flow Per loop, gpm 93600 93600 93600 93600 Total flow, gpm 374400 374400 374400 374400 Reactor Flow, 140.9 139.8 139.4 142.9 Total, (Mlbm/hr)
Reactor Coolant Press, psia 2250 2250 2250 2250 Core Bypass, t 8.4 8.4 8.4 8.4 Fuel Design 17x17 17x17. 17x17 17x17 V5H w/IFMs V5H w/IFMs V5H w/IFMs V5H w/IFMs Reactor Coolant Temnerature. *F Core Outlet 618.4 623.3 625.0 608.5 Vessel Outlet 613.2 618.2 620.0 603.2 Core Average 585.7 591.1 593.0 575.1 Vessel Average 581.2 586.5 588.4 570.7 Vessel / Core Inlet 549.3 554.8 556.8 538.2 Steam Generator Outlet 549.0 554.5 556.6 538.0' Vessel AT (Thot-Teold) 63.9 63.4 63.2 65.0 Steam Generator Steam Temperature, F 530.6 536.3 538.4 519.4 Steam Pressure, psia 890 934 950 807 Steam Flow, total, Mlbm/hr 15.88 15.91 15.92 15.83 Feedwater Temp, F 446 446 446 446 Zero Load Temp, F 557 557 557 557 SG Tube Plugging, % 10 10 10 10 i
l Attachment I to NA-XXXX Page 6 of 11 A. Core Thermal Limit Protection Protection of the core thermal limits is graphically illustrated in WCGS USAR Figure 15.0-1, and is provided in Figure 1 below. Figure 1 shows that OTAT
' and OPAT protection system setpoints and the steam generator safety valves prevent system violation of the Vessel Exit Boiling Limits, Departure from Nucleate Boiling Ratio (DNBR) correlation quality limits, and DNB limits for a range of pressurizer pressures defined by the Low and High Pressurizer Pressure trip setpoints.
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80 ,
OPAT Trip Line
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s- s-N N \
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70 N N N
\ s N N s
\ \ \ g
\ s p g
N N 01 o s s A*3 N As s O s s N
- 60 s s s Da N s
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s s t N N o s s N s
O N s N s
s s N s
' s 50 %
s s N N N \
\ \ \
N '
OTAT Trip Line s s
SGSV Line N 570 580 590 600 610 620 Tavg. (Deg-F)
FIGURE 1 The vessel exit boiling lines are determined from an enthalpy balance across the vessel. Since the core average temperature is not used in the determination of the Vessel Exit Boiling Limits, the proposed increase in Tavg does not affect the Vessel Exit Boiling Limits generated for Cycle 7 .
operation.
The core DNB limit lines shcw the limits for which there is L. 95% probability at the 95% confidence level that DNB will not occur during normal operation, )
operational transients, or 4s a result of conditions arising from any fault of l moder3N frequency. The _. :neration of the DNB limit lines is straightforward; i I
at es; combination of pressure and power level, an iteration on inlet temperature is made lo determine the inlet temperature that yields the applicable cycle critical heat flux thermal design limit (See Section B).
Since the core average temperature is not used in the determination of the DNB limit lines, the proposed increase in T avg does not affect the DNB limit lines originally generated for Cycle 7 rerate operation. '
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! Attachment I to NA 94-0018
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I Although the proposed increased core average temperature (Ta vg) does not j affect the core thermal limit lines previously derived for Cycle 7 operation, ;
the relationships between the OTAT and OPAT protection functions and the core !
3 limits when compared on a AT versus Ta vg coordinate system do change as a result of the increased T avg. Also, limitations on reactor power and temperature are imposed by the steam generator safety valves (SGSV). As primary side temperature increases, heat transfer across the steam generator tubes will eventually result in secondary side conditions that lift the SGSVs.
The SGSV line in the core thermal limits shows the limits at which operation is possible without resulting in the actuation of the SGSVs. The SGSV line is computed from the log mean temperature difference equation for heat transfer across a shell and tube heat exchanger. This calculation is dependent upon the vessel outlet temperature.
Protection of the core thermal limits has been confirmed for operation at the revised rerate conditions with vessel T avg 5 586.5 *F. This protection is provided by the revised OTAT and OPAT trip setpoints and the revised Axial Flux Difference versus Power limits.
B. Thermal-Hydraulic Analysis A complete description of the thermal-hydraulic methods used by WCNOC for Cycle 7 DNB evaluations is provided in WCNOC's January 5, 1993 submittal for the proposed Cycle 7 rerate of WCGS (Letter NA 93-0001, from R. C. Hagan, WCNOC to the USNRC).
No changes to the thermal-hydraulic methods and core thermal limits presented in the January 5, 1993 submittal are required due to the revised operating temperatures proposed in this application for amendment. The limiting ANS Condition II transient with respect to DNB, the Loss of Flow event, has been re-evaluated to ensure that minimum DNBR requirements are maintained.
C. Core Design The operational and transient limits /setpoints applicable for the remainder of Cycle 7 operation at 3565 MWt with increased T avg ($586. 5 "F) have been examined. Specifically, the allowable Axial Flux Difference (AFD) versus technical specification power limits and the OTAT/OPAT trip limits were examined.
The AFD limits are set based upon che results of a maneuvering analysis as described in NSAG-007, Rev. O, " Reload Safety Evaluation Methodology for the Wolf Creek Generating Station," W. S. Kennamore, et.al., January, 1992. The maneuvering analysis was re-performed for the later portion of Cycle 7 operation with an increased T avg. As a result of the calculation, the allowable f(AI) limits at 100% power were revised to more restrictive limits than are currently allowed.
The OTAT and OPAT trip limits are established to protect Centerline Fuel Melting and Reactor Protection System DNB limits. These limits are established based upon the results of the maneuvering analysis. This analysis was performed for the later portion of Cycle 7 with an increased Tavg. The resultant OTAT positive f(AI) breakpoint setting produces a more restrictive operating band than is currently allowed. The OPAT limit remains unchanged.
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Attachment I to NA 94-0018 Page 8 of 11 The cycle specific core kinetics parameters checked in the Reload Safety Analysis Checklist were evaluated for the remainder of Cycle 7 operation with
- an increased T avg of up to 586.5 "F. Key safety parameter 9 including the moderator temperature coefficient, shutdown margin, Doppler caefficients and end of life trip reactivity were calculated and found to be bounded by the assumed values used in safety analyses.
D. LOCA and Non-LOCA Analyses Non-LOCA Analyses A review of the USAR Chapter 15 non-LOCA accidents was performed to determine the transients affected by the increase in vessel average temperature from 581.2 'F up to 586.5 'F. It was determined that only the Complete Loss of Forced Reactor Coolant Flow transient required re-analysis due to the change in temperature. The remaining non-LOCA analyses have been analyzed over a spectrum of vessel average temperatures ranging from 570.7 'F (15 'F Thot reduction) to 588.4 'F (1.8 'F Thot increase) (see Table 2) . These analyses have been evaluated following WCGS Statistical Core Design Methodology (TR 0025, WO1, " Core Thermal Hydraulic Analysis Methodology for the Wolf Creek Generating Station," W. S. Kennamore, et.al., July, 1990) for DNB concerns utilizing the WRB-2 correlation. The results show in all cases that the minimum DNBR remains above the safety analysis limit.
For transients which fall outside of the Statistical Core Design Analysis scope, the statepoints were evaluated using the W-3 correlation (TR-90-0025).
Events in this category are analyzed below 70% power and have been analyzed following the appropriate core average temperature program corresponding to the previously mentioned temperature range. The minimum DNBR for the events evaluated with the W-3 correlation have also been shown to remain above the safety analysis linit.
The Chapter 15 accidents in which RCS overpressurization is a concern, '
including Complete Loss of Forced Reactor Coolant Flow, have been analyzed assuming a vessel average temperature of 588.4 'F following WCGS Transient Analysis methodology (NSAG-006, Rev. O, " Transient Analysis Methodology for the Wolf Creek Generating Station," W. D. Wagner, et. al., January, 1991).
The results of overpressurization analyses show that the maximum RCS pressure remains below the safety analysis limit in all cases.
Complete Loss of Forced Reactor Coolant Flow The RCS response to a Complete Loss of Forced Reactor Coolant Flow results in conditions adverse to heat transfer from the fuel to the coolant. The t
immediate response to a loss of flow is a rapid increase in coolant temperature caused by the reduction in heat transfer to the secondary side through the steam generators. The increase in coolant temperature causes an increase in RCS pressure, This pressure increase is a slight benefit to heat transfer between the fuel and coolant, however it is not significant in comparison to the degradation in heat transfer caused by the reduction in coolant flow and the increase in coolant temperature.
l l The Complete Loss of Forced Reactor Coolant Flow transient is classified as an l ANS Condition III event, an infrequent fault. This event is analyzed to verify that the accident will not result in RCS overpressurization and to show that fuel damage will not occur due to the accident.
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Attachment I to NA 94-0018
' page 9 of 11
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' The methodology followed to justify operation with a vessel average- l temperature of up to 586.5 'F is identical to that presented in the j application for amendment associated with the WCGS Cycle 7 submitted on October 28, 1992 (Letter NA 92-0073, from R. C. Hagan, WCNOC to the USNRC) .
The analysis was performed assuming rerate conditions with a +6.0 moderator ]
temperature coefficient,- 10% steam generator tube plugging and a vessel average temperature of 586.5 'F. The initial reactor power, pressure and ,
temperature were assumed at nominal values, and RCS flow was assumed at i minimum measured flow to meet Statistical Core Design analysis requirements.
A conservatively large absolute value of the Doppler-only power coefficient
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was assumed, which, in combination with the moderator temperature coefficient assumption, results in the maximum power being reached early in the transient.
Reactor protection is assumed to be provided by an RCP power supply undervoltage or underfrequency signal only.
The results of the Complete Loss of Forced Reactor Coolant Flow analysis are comparable to the results presented in the application for amendment associated with the WCGS Cycle 7. Minimum DNBR remains above the safety ,
analysis limit.
The Complete Loss of Forced Reactor Coolant Flow overpressurization analysis was performed previously assuming a vessel average temperature of 588.4 'F and includes the appropriate steady state errors consistent with full power j operation applied in a direction to create the minimum margin to system ;
overpressurization. The analysis shows that the maximum RCS pressure' attained- j remains well below the limit of 110% of RCS design pressure. ,
This analysis demonstrates that, for a Complete Loss of Forced Reactor Coolant .
I Flow, both the maximum system pressure and minimum DNBR' remain within their respective safety analysis limits.
LOCA and LOCA Related Evaluations l A LOCA is the result of a pipe rupture in the RCS pressure boundary. For the analyses reported here, a small break is defined as a rupture ' of the RCS piping with a cross-sectional area less than 1.0 square foot, in which the normally operating charging system flow is not sufficient to sustain pressurizer level and pressure. This event is considered an ANS Condition III event, which is a fault which may occur very infrequently during the life of a plant. A major (large) break is defined as a rupture with a total cross-sectional area equal to or greater than 1.0 square foot. This event is l considered an ANS Condition IV event, a limiting fault, in that it is not l expected to occur during the life of the WCGS, but is postulated as a conservative design basis accident.
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! Small Break LOCA The small break LOCA (SBLOCA) was analyzed to demonstrate conformance with 10 CFR 50.46 requirements. The SBLOCA analysis was re-performed for the rerate effort as described in Section 6.1 of the application for amendment associated with the WCGS Cycle 7 submitted on October 28, 1992 (Letter NA 92-0073, R. C.
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Attachment I to NA 94-0018 Page 10 of 11 The analysis presented in the Cycle 7 report was performed for two temperature conditions: a 1.8 *F Thot increase (620.0 *F) and a 15 *F Thot reduction (603.2
- F) . Therefore, the proposed operating condition with a Thot reduction j between 0 and 5 *F is bounded by the existing analyses. Increasing the vessel '
average temperature from 581.2 *F up to 586.5 *F does not create an unanalyzed )
condition or violate any acceptance criteria.
l Large Break LOCA The large break LOCA was analyzed to demonstrate conformance with 10 CFR 50.46 ,
I requirements. The LOCA analysis was re-performed for the rerate effort as described in Section 6.2 of the application for amendment associated with the WCGS Cycle 7 submitted on October 28, 1992 (Letter MA 92-0073, R. C. Hagan, WCNOC to the USNRC).
The analysis presented in the Cycle 7 report demonstrates that initiating the LOCA with a 15 *FThot reduction (603.2 *F) results in a more severe transient than when analyzed with a 1.8 *F Thot increase (620.0 *F) due to the density of the water spilling from the break. Therefore, increasing the vessel average temperature decreases the severity of the accident including mass and energy releases. The proposed operating condition with a Thot reduction of between 0*- 5* F is bounded by the existing analyses. Increasing the vessel average temperature from 581.2 *F up to 586.5 *F does not create an unanalyzed condition or violate any acceptance criteria.
Discussion of the impact of the proposed temperature reduction on the LOCA mass and energy release is provided in Section E, Containment Analysis.
LOCA Related Evaluations LOCA hydraulic forcing functions, hot leg switchover to prevent boron precipitation, and the post-LOCA long term cooling subcriticality requirement evaluations performed for the power rerate project were reviewed to evaluate the impact for operation without the 5 *F Thot reduction. Review of these events shows that they are not dependent on RCS temperature and not affected by the proposed change. Therefore, the LOCA related evaluations performed for power rerate remain valid for operation with a Thot reduction between 0 -588 F.
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E. Containment Analyses ]
I The containment integrity analysis provides the analytical basis for operation of WCGS. Specifically, the containment integrity analysis demonstrates acceptability of operation by demonstrating that the containment peak pressure resulting from a design basis large break LOCA event or Main Steamline Break (MSLB) event will not exceed the design pressure or temperature. The analysis ensures that the containment heat removal capability is sufficient to remove the maximum possible discharge of mass and energy to containment from the Nuclear Steam Supply System without exceeding design pressure. The analysis is also used as a basis for the containment leak rate test pressure to ensure that dose limits will be met in the event of a release of radioactive material l to containment. I Bounding initial temperatures and pressures for the containment integrity analysis are selected to envelope the limiting conditions for operation.
Thus, the most limiting conditions for operation at full Nuclear Steam Supply System power were conservatively chosen.
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Attachment I to NA 94-0018 j Page 11 of 11 Results of the containment integrity analyses performed previously and j documented in the January 5, 1993 submittal (Letter NA 93-0001, from R. C. I Hagan, WCNOC to the USNRC) remain valid for the revised operating conditions l at 3565 MWt and a 0*- 5* F Thot reduction. These analyses showed that the containment internal structures and attachments are adequate to withstand l dynamic localized pressurization effects associated with a design basis pipe l break accident under rerate conditions.
1 i Additional Considerations 1
Radiological Consequences The radiological consequences in Chapter 15 of the WCGS USAR were re-analyzed
( as part of the power rerate effort. Detailed results of the analyses were provided in Section 5.7 of the applicaticn for amendment associated with the (
WCGS Cycle 7 submitted on October 28, 1992 (Letter NA 92-0073, R. C. Hagan, WCNOC to the USNRC). Reactor core and coolant activities are not affected by the RCS temperature change since the activities are based on cycle length and core burnup. Primary and secondary parameter changes caused by the change in RCS temperature which could affect radiological releases are bounded by the parameters for the 1.8 'F Thot increase and 15 *F Thot reduction analyses as outlined in Table 2. Therefore, radiological consequences for operation with a 0*-5 F Thot reduction are bounded by the power rerate analyses performed previously.
Accidents of a Different Type Cycle 7 rerate operation with a Thot reduction of between 0*-5* F will not create the possibility of any new accidents of a type not previously considered in the USAR. The USAR Chapter 15 accident analyses assume initial conditions resulting in the worst case conditions consistent with the Thot increase or Thot reduction parameters listed in Table 2. Therefore, proposed l operation without the Thot reduction is bounded by previous analyses and does not create a new or unanalyzed condition.
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i Attachment II to NA 94-0018 Page 1 of 3 l
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l ATTACHMENT II NO SIGNIFICANT HAZARDS CONSIDERATION DETERMINATION l
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1 Attachment II to NA 94-0018
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No Significant Hazards Consideration Determination This license amendment request proposes to revise the Technical Specifications l
to allow plant operation at a core power of 3565 megawatts thermal (MWt) and a hot leg temperature (Thot) between 613.2 *F and 618.2 *F. This hot leg temperature represents an increase in Thot from the current value of 613.2 *F.
The vessel average temperature (Tavg) will also increase from the current l value of 581.2 *F up to the proposed maximum value of 586.5 *F. This proposed i license amendment removes the limitation that WCGS operate with the 5 *F Thot I reduction limit requested in our January 5, 1993 submittal (Letter NA 93-0001, l from R. C. Hagan, WCNOC to the USNRC).
The results of the safety evaluation for the proposed temperature revision confirm the acceptability of plant operation at 3565 MWt without the current 5 *F Thot reduction. This justification is based on the application of revised OPAT and OTAT trip setpoints. The setpoint changes are due to the l operating temperature change and a change in the Axial Flux Difference (AFD) l versus Power Limits to account for revised peaking limits resulting from the i re-analysis of the Complete Loss of Forced Reactor Coolant Flow event.
Evaluations also confirm that there is no increase in transient specific calculations of fuel rod failure. Evaluations and analyses support the conclusion that all safety analysis acceptance criteria continue to be met.
Standard I - Involve a Significant Increase in the Probability of Consequences of an Accident Previously Evaluated 1
l The probability of occurrence and the consequences of an accident evaluated
- previously in the USAR are not increased due to the propo
- ,ed technical
( specification change. Plant operation at 3565 MWt with the revised ,
temperatures does not affect any of the mechanisms postulated in the USAR to cause LOCA or non-LOCA design basis events. Analyses, evaluations and minimum DNBR calculations confirm that the USAR conclusions remain valid for the proposed changes. On these bases it is concluded that the probability and consequences of the accidents previously evaluated in the USAR are not increaed.
Standard II - Create the Possibility of a New or Different Kind of Accident from any Previously Evaluated The proposed technical specification changes do not increase the probability ,
of occurrence of a malfunction of equipment important to safety or increase
- the consequences of a malfunction of equipment evaluated in the USAR. The technical specification changes do not create the possibility of a new or -
different kind of accident from any accident previously evaluated because the change in operating That will not impose a new operating configuration that i
would create a new failure scenario. The proposed changes do not change the l plant configuration in a way that introduces a new potential hazard to the plant and do not involve a significant reduction in the margin of safety. No new failure modes will be created by the proposed changes for any plant !
l equipment. Operation with a 0*-5* F Thot reduction is bounded by the analyses ,
performed previously for the power rerate and approved by the NRC in Amendment l No. 69 to the WCGS Technical Specifications on November 10, 1993, and does not create a new or unanalyzed condition. For these reasons, the possibility of a new accident which is different from any already evaluated in the USAR is not created.
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Attachment II to NA 94-0018 I
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Standard III - Involve a Significant Reduction in the Margin of Safety l
The analyses and evaluations discussed in the safety evaluation demonstrate that all applicable safety analysis acceptance criteria continue to be met for the proposed operating conditions. The change in operating Thot does not involve a significant reduction in a margin of safety because the operating temperature is one of the inherent assumptions that determines the safe operating range defined by the accident analyses, which are in turn protected by the technical specifications. The acceptance criteria for the accident analyses are conservative with respect to the operating conditions defined by ;
the technical specifications. The analyses performed for the power rerate and this proposed change confirm that the accident analyses criteria are met at the revised configuration. Therefore, it is concluded that the proposed change does not involve a reduction in a margin of safety described in the bases to any technical specification. i l
Based on the above discussions, it has been determined that the requested '
technical specification revision does not involve a significant increase in the probability or consequences of an accident or other adverse condition over !
previous evaluations; or create the possibility of a new or different kind of !
accident or condition over previous evaluation; or involve a significant l i reduction in a margin of safety. Therefore, the requested license amendment l does not involve a significant hazards consideration.
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Attachment III to NA 94-0018
- Page 1 of 2 ATTACHMENT III ENVIRONMENTAL IMPACT DETERMINATION 9
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Attachment III to NA 94-0018 Page 2 of 2
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Environmental Impact Determination
! 4 l This amendment request meets the eligibility criteria for categorical l
- exclusion set forth in 10 CFR 51.22 (c) (9) as specified below
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l (i) the amendment involves no significant hazards consideration l l
As demonstrated in Attachment II, the proposed changes do not involve any significant hazards consideration.
(ii) there is no significant change in the types or significant increase in the amounts of any effluents that may be released offsite The proposed changes involve decreasing the 5' Thot reduction associated with the WCGS power rerate to 3565 MWt. The changes will not affect the types of effluents since no new process is involved. Operation with a 0 -5 That reduction is bounded by the analyses performed for the WCGS power rerate, thus decreasing the 5* Thot reduction will not result in a significant change in the previously analyzed amounts of any effluents that may be released offsite.
(iii) there is no significant increase in individual or cumulative occupational radiation exposure Radiological consequences for operation with a 0*- 5' F Thot reduction are bounded by the power rerate analyses performed previously. Therefore, the proposed changes will not result in a significant increase in individual or cumulative occupational radiation exposure.
Based on the above, it is concluded that there will be no impact on the environment resulting from the proposed changes and that the proposed changes meet the criteria specified in 10 CFR 51.22 for a categorical exclusion from the requirements of 10 CFR 51.21 relative to requiring a specific environmental assesament by the Commission.
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