ML20217A242
| ML20217A242 | |
| Person / Time | |
|---|---|
| Site: | South Texas  |
| Issue date: | 09/30/1999 |
| From: | HOUSTON LIGHTING & POWER CO. |
| To: | |
| Shared Package | |
| ML20217A238 | List: |
| References | |
| NUDOCS 9910080122 | |
| Download: ML20217A242 (15) | |
Text
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NOC. AE-000651 Page 3 of 6 ATTACHMENT 2 TECHNICAL SPECIFICATION CHANGES 3
1 9910000122 990930 PDR ADOCK 05000498 l
P PDR
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' NOC-AE-000651 Page 4 of 6 BACKDROUND Please note that this supplement does not provide a detailed discussion of the entire request.
Reviewers should consult the original submittals (References 1 and 2) for greater detail.
' After STP Nuclear Operating Company (STPNOC) submitted a request (Reference 1) to update Reactor Coolant System (RCS) flow values in conjunction with replacement of Model E steam generators with Model A94 steam generators, we submitted another request (References 2,3 and 4). This request implemented Nuclear Regulatory Commission (NRC) Generic Letter (GL) 88-
.16,." Removal ofCycle-specific Parameter Limits from Technical Specifications." The NRC has approved this request (Reference 5). It relocated to the Core Operating Limits Report (COLR) all of the operating parameters that were the subject of cur first submittal. Thus, there is no longer a need for the NRC to review changes to these values. However, the approved amendment also replaced one of the parameters in TS 3/4.2.5,"DNB PARAMETERS," with a closely related parameter. It replaced " Reactor Coolant System Flow" with the non-cycle-specific parameter," Thermal Design Reactor Coolant System Flow (TDF)."
Safety analyses in the submittals (References 1 and 2) discuss TDF and why it is preferred in this application. The Determinations ofNo Significant Hazards Consideration in these submittals also remain valid.
To minimize additional review by the NRC that may be caused by this supplement, supporting text of the pending submittal is not changed. The NRC has already reviewed this material, it is still valid for the one remaining parameter and thus, we do not repeat it in this supplement.
' DESCRIPTION OF PROPOSED CHANCES r
' A TDF value of 370,000 gpm was provided in TS 3/4.2.5,"DNB PARAMETERS," for use with the Model E steam generators. It is now necessary to add a corresponding TDF value of 392,000 gpm for the new Model A94 steam generators.
' STPNOC has also decided to not change the value for steam generator high-high level fo-the Model A94. Please ignore references to a change in high-high steam generator level.
. To improve the quality of the submittal and assist review, STPNOC is adding an attachment that contains the final version of the TS to show the intended outcome of this request.
In summary, this supplement makes the following changes to the submittal (Reference 1):
' 1. Removes from consideration those relocated parameters that no longer need review,
- 2. Removes from the submittal those TS pages associated with the relocated parameters,
- 3. ' Requests to add a TDF value in TS 3/4.2.5, "DNB PARAMETERS," for the Model A94,
- 4. Removes references to a changed value for Steam' Generator High-High Level, and
- 5. Provides new TS pages as they should appear with the requested changes incorporated.
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Specific Changes:
. Remove all marked-up TS pages no longer affected by this request, l
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NOC-AE-000651 Page 5 of 6 1
'e' On Page 3/4 2-11, make existing RCS TDF rate in LCO 3.2.5.c applicable to Model E steam
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generators and add a TDF value of 392,000 gpm applicable to Model A94 steam generators.
On Pages B 3/4 2-5 and B 3/4 2-6, make existing flow requirement applicable to Model E steam generators and add a new flow requirement applicable to A94 steam generators.
' Add attachment 6," Technical Specification Reconstituted Pages," the new TS and Bases j
pages with proposed changes incorporated.
i SAFETY EVALUATION This is a summary and is not intended to replace the safety evaluations contained in the original requests (References 1 and 2).
Relocation of Reactor Coolant System (RCS) related TS limits to the COLR pmvides various benefits and is supported by existing NRC-approved methodologies, such as the Westinghouse reload methodology (WCAP-9272-P-A). This WCAP examines on a cycle-specific basis each of the DNB limits of Reactor Coolant System T.vs, Reactor Coolant System total flow rate, pressurizer pressure, OTDT and OPDT setpoints, and the supporting bases for the setpoints. The NRC has approved (Reference 5) relocating these values to the COLR.
Methodologies for calculation of the relocated values confonn to Topical Report WCAP-14483-A, " Generic Methodology for Expanded Core Operating Limits Report," approved January 19, i
1999. TS require STPNOC to provide copies of the COLR to the NRC each time it is changed.
. Relocated TS values remain consistent with the STP Updated Final Safety Analysis Report (UFSAR) and continue to provide their safety function through the involved TS. Actions required to be taken when parameters exceed limits remain in the TS.
The relocation of cycle-specific RCS flow related values to the COLR allows STP to manage available margins to increase cycle operating margins and optimize core reload design. It also provides improved correlation between TS, safety analyses, and control and protection systems settings for each cycle. Finally, relocation of these values to the COLR conserves resources by reducing the number oflicense amendment submittals and associated NRC reviews.
Consistent with WCAP-14483-P-A, STPNOC changed the Reactor Coolant System Flow value in the Technical Specifications to Thermal Design Reactor Coolant System Flow, as approved by the Nuclear Regulatory Commission in Amendments 97 and 84 on September 29,1998. This request designates the value currently shown as applicable to Model E steam generators, and adds a new value for Model A94 steam generators.
RCS Minimum Measured Flow (MMF) rate increases from 392,300 gpm for Model E steam generators to 403,000 gpm with a best-estimate core bypass flow of 6.5 % for Modd A94 steam generators. RCS Thermal Design Flow (TDF) rate increases from 370,000 gpm for Model E steam generators to 392,000 gpm with a design core bypass flow of 8.5 % for Model A94 steam generators. Decreased flow resistance in Model A94 steam generators results m increased RCS flow. Increasing RCS flow bypassing the fuel rod region of the reactor core negates undesired effects or the core from increased loop flow. This is known as " core bypass flow." The increase in core bypass flow is created by removing thimble-plugging-devices and as a consequence of a b
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I NOC-AE-000651 Attchment 2 Page 6 of 6
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'rhodifiEation that converted reactor vessel upper head temperature to "T-cold." Although total E RCS flow has increased with _new Model A94 steam generators, the modifications that increase core bypass flow allows coolant flow in the fuel rod region to remain essentially unchanged.
Therefore, even though there will be an increase in overall flow after STP installs Model A94
- steam generators, there will be no significant change in flow through the fuel rod region of the reactor core. Thus, no significant affect on core thermal and hydraulic dynamic design factors will result.
REFERENCES:
1.
. NOC-AE-000080, Letter from T. H. Cloninger to NRC Document Control Desk dated May
- 7,' 1998," Proposed Amendment to Technical Specifications to Reflect Replacement Steam '
Generator Reactor Coolant Flow."
2.
NOC-AE-000471, Letter from T. H. Cloninger to NRC Document Control Desk dated June 7,1999," Proposed License Amendment for Relocation of Cycle-Specific Operating
' Parameters fmm the Technical Specifications to the Core Operating Limits Report."
3.'
- NOC-AE-000569, Letter from T. H. Cloninger to NRC Document Control Desk dated June 24,1999 " Supplement to the Proposed License Amendment for Relocation of Cycle-l Specific Operating Parameters from the Technical Speci_fications to the Core Operating Limits Report."
~ 4.
NOC-AE-000601, Letter from J. J. Sheppard to NRC Document Contml Desk dated
- August 24,1999," Supplement 2 to the Proposed License Amendment for Relocation of Cycle-Specific Operating Parameters from the Technical Specifications to the Core Operating Limits Report."
5.-
NRC Letter dated September 2,1999, from Thomas W. Alexion to William T. Cottle,
" SOUTH TEXAS PROJECT, UNITS 1 AND 2-ISSUANCE OF AMENDMENTS RE:
I RELOCATION OF CYCLE-SPECIFIC PARAMETERS FROM THE TECHNICAL SPECIFICATIONS TO THE CORE OPERATING LIMITS REPORT (TAC NOS.-
MA5693 AND MA5694)
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j NOC-AE-000651 Page 1 of 7 ATTACHMENT 3 TECHNICAL SPECIFICATION MARKED-UP PAGES REMOVED BY SUPPLEMENT
NOC-AE-000651 Attaciunent 3 Page 2 of 7 A
JECHNICAL SPECIFICATIONS MARKED-UP PAGES TO BE REMOVED Please remove the below listed Technical Specification marked-up pages from the STPNOC submittal, NOC-AE-0080, dated May 7,1998.
Pages:
2-1 2-2
. 2-2a (new figure 2.1-2) i 2-4 B 2-1 l
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NOC-AE-000651 Page1of5 ATTACHMENT 4 1
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TECHNICAL SPECIFICATION MARKED-UP PAGES j
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NOC-AE-000651 if Page 2 0f 5 PROPOSED CHANGES TO TECHNICAL SPECIFICATIONS The below listed Technical Specification pages and Bases pages are annotated to identify
-Technical Specification changes proposed as a result of differences in RCS flow created by replacement of Westinghouse Model E steam generators with Westinghouse Model A94 steam generators. The proposed changes are exhibited in this attachment.
c+H+- *e9 is to be removed and gderlined text is to be added.
Pages:
3/4 2-11 B 3/4 2-5
' B 3/4 2-6 f
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POWER DISTRIBUTION LIMITS -
i 3/4.2.5 DNB PARAMETERS i
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~' LIMITING CONDITION FOR OPERATION 3.2.5 The following DNB-related parameters shall be maintained within the limits following:
.j a.
Reactor Coolant System Tm, s; the limit as specified in the Core Operating Limits Report.
b.
Pressurizer Pressure, > the limit as specified in the Core Operating 1.imits Report.
a.-
Thermal Design Reactor Coolant System Flotu ??0.a^^ ;;. for:
'1.
Model E Steam Generators.
2 370.000splD i
2.
Model A94 Steam Generators._2 392.000 com 1
APPLICABILITY:' MODE 1.
ACTION:
With any of the above parameters exceeding its limit, restore the parameter to within its limit within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or reduce THERMAL POWER to less than 5% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.-
SURVEILLANCE REQUIREMENTS 4.2.5.1-Each of the parameters shown above shall be verified to be' within its limits at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. The provisions of Specification 4.0.4 are not applicable for verification that RCS flow is within its limit.
4.2.5.2.
The RCS flow rate indicators shall be subjected to a channel calibration at least once per 18 months.
NOTE SR 4.2.5.3 is required at beginning-of-cycle with reactor power 290% RTP.
4.2.5.3 The RCS total flow rate shall be determined by precision heat balance or elbow tap AP.
- measurements at least once per 18 months. The provisions of Specification 4.0.4 are not applicable.
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. SOUTH TEXAS-UNITS 1 & 2 3/4 2-11 Unit 1 - Amendment No. 446 Unit 2-Amendment No 403 c
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POWER DISTRIBUTION LIMITS NO CHANGES THIS PAGE BASES HEAT FLUX HOT CHANNEL FACTOR and NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR (continued)
When an Fo measurement is taken, an allowance for both experimental error and manufacturing tolerance must be made. An allowance of 5% is appropriate for a full-core map taken with the incore Detector Flux Mapping System, and a 3% allowance is appropriate for manufacturing tolerance.
The Radial Peaking Factor, F,y(Z), is measured periodically to provide assurance that the Hot Channel Factor, Fo(Z), remains within its limit. The F,y limit for RATED THERMAL POWER (F,y") as provided in the Core Operating Limits Report (COLR) per Specification 6.9.1.6 was determined from expected power control maneuvers over the full range of bumup conditions in the core.
3/4.2.4 QUADRANT POWER TILT RATIO The QUADRANT POWER TILT RATIO limit assures that the radial power distribution satisfies the design values used in the pcwer capability analysis. Radial power distribution measurements are made during STARTUP testing and perlodically during power operation.
The limit of 1.02, at which corrective action is required, provides DNB and linear heat generation rate protection with x-y plane power tilts. A limit of 1.02 was selected to provide an allowance for the i
uncertainty associated with the indicated power tilt.
The 2-hour time allowance for operation with a tilt condition greater than 1.02 is provided to allow identification and correction of a dropped or misaligned control rod. in the event such action does not correct the tilt, the margin for uncertainty of Fo is reinstated by reducing the maximum allowed power by 3% for each percent of tilt in excess of 1.
For purposes of monitoring QUADRANT POWER TILT RATIO when one excore detector is inoperarable, the moveable incore detectors are used to confirm that the normalized symmetric power distribution is consistent with the QUADRANT POWER TILT RATIO. The incore detector monitoring is
'done with a full incore flux map or two sets of four symmetric thimbles. The two sets of four symmetric thimbles is a unique set of eight detector locations. These locations are C-8, E-5, E-11, H-3, H-13, L-5, L-11, N-8.
' 3/4.2.5 DNB PARAMETERS In MODE 1, the limits on pressurizer pressure, RCS coolant average temperature, and RCS flow rate must be maintained during steady state operation in order to ensure DNBR criteria will be met in the event of an unplanned loss of forced coolant flow or other DNB limiting transient. In all other MODES, the power level is low enough that the DNB is not a concem.
The values presented in the COLR are indicated values and include measurement uncertainties. The value for pressurizer pressure is averaged using plant computer /QDPS readings from a minimum of at least 3 channels. The value for RCS coolant average temperature is averaged using control board readings from a minimum of at least 3 channels. The value for RCS flow rate is the average from a minimum of at least 2 flow transmitters per RCS loop using plant computer /QDPS points.
SOUTH TEXAS - UNITS 1 & 2 B 3/4 2-5 Unit 1 - Amendment No.W,115 Unit 2 - Amendment No. 47,103
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POWER DISTRIBUTION LIMITS BASES 3/4.2.5 DNB PARAMETERS (continued)
The value for thermal design RCS flow rate presented in Technical Specification 3.2.5 is the l
thermal design reactor coolant system flow rate used in the analysis approved by the Nuclear Regulatory Commission in Amendments 97 and 84 on September 29,1998. This flow rate is an analytical limit censistent with 10% tube plugging of the steam generator tubes and Departure from Nucleate Boiling requirements. Thermal design RCS flow rate with Model E Steam Generators is 370.000 gom.,_pnd with Model A94 Steam Generators it is 392.000 com. These values include 2.8%
measurement uncertainty.
The RCS flow measurement uncertainty of 2.8% bounds the precision heat balance and the
- elbow tap Ap measurement methods. The elbow tap Ap measurement uncertainty presumes that elbow tap Ap measurements are obtained from either QDPS or the plant process computer. Based on instrument uncertainty assumptions, RCS flow measurements using either the precision heat balance
- or the elbow tap Ap measurement methods are to be performed at greater than or equal to 90% RTP at the beginning of a new fuel cycle. The elbow tap Ap RCS flow measurement methodology is described
~ in ST-HL-AE-5707, " Proposed Amendment to Technical Specification Table 2.2-1 and 3/4.2.5 for Reactor Coolant System Flow Monitoring -- Revised," dated August 6,1997, and in ST-HL-AE-5752,
." Amended Response to Request for Additional Information on the Proposed Elbow Tap Technical
. Specification Change (Table 2.2-1 and Section 3/4.2.5)," dated September 18,1997.
The 12-hour periodic surveillance of these parameters through instrument readout is sL3klent to ensure that the parameters are restored within their limits following load changes and other expected transient operation.
i SOUTH TEXAS - UNITS 1 & 2 8 3/4 2-6 Unit 1 - Amendment No..$.15 Unit 2 - Amendment No.443 x
NOC-AE-000651 Page 1 of 4 ATTACHMENT 5 TECHNICAL SPECIFICATION RECONSTITUTED PAGES
[ Note: These pages represent the Technical Specification with amendments incorporated, and are provided for the reviewer's convenience.]
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o PO'NER DISTRIBUTION LIMITS 3/4.2.5 DNB PARAMETERS LIMITING CONDITION FOR OPERATION 3.2.5 The following DNB-related parameters shall be maintained within the limits following:
a.
Reactor Coolant System T.,, s; the limit as specified in the Core Operating Limits Report b.
Pressurizer Pressure, > the limit as specified in the Core Operating Limits Report c.
Thermal Design Reactor Coolant System Flow for:
1.
Model E Steam Generators, 2 370,000 gpm 2.
Model A94 Steam Generators, 2 392,000 gpm APPLICABILITY: MODE 1.
3 ACTION:
With any of the above parameters exceeding its limit, restore the parameter to within its limit within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or reduce THERMAL POWER to less than 5% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
SU_RVEILLANCE REQUlREMENTS _.__ __ _ _ __
4.2.5.1 Each of the parameters shown above shall be verified to be within its limits at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. The provisions of Specification 4.0.4 are not applicable for verification that RCS flow is within its limit.
4.2.5.2 The RCS flow rate indicators shall be subjected to a channel calibration at least once per 18 months.
NOTE SR 4.2.5.3 is required at beginning-of-cycle with reactor power 290% RTP.
4.2.5.3-The RCS total flow rate shall be determined by precision heat balance or elbow tap AP measurements at least once per 18 months. The provisions of Specification 4.0.4 are not applicable.
SOUTH TEXAS-UNITS 1 & 2 3/4 2-11 Unit 1 - Amendment No445 Unit 2 - Amendment nom 3
c POWER DISTRIBUTION LIMITS BASES
. HEAT FLUX HOT CHANNEL FACTOR and NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR j
-(continued) nT When an Fo measurement is taken, an allowance for both experimental error and manufacturing tolerance must be made. An allowance of 5% is appropriate for a full-core map taken with the incore Detector Flux Mapping System, and a 3% allowance is appropriate for manufacturing tolerance.
- The Radial Peaking Factor, F,y(Z), is measured periodically to provide assurance that the Hot Channel Factor, Fo(Z), remains within its limit. The F,y limit for RA TED THERMAL POWER (F,yRTP)gg provided in the Core Operating Limits Report (COLR) per Specification 6.9.1.6 was determined from expected power control maneuvers over the full range of bumup conditions in the core.
3/4.2.4 QUADRANT POWER TILT RATIO The QUADRANT POWER TILT RATIO limit assures that the radial power distribution satisfies the design values used in the power capability analysis.' Radial power distribution measurements are made during STARTUP testing and periodically during power operation.
The limit of 1.02, at which corrective action is required, provides DNB and linear heat generation rate protection with x-y plane power tilts. A limit of 1.02 was selected to provide an allowance for the uncertainty associated with the indicated power tilt.
The 2-hour time allowance for operation with a tilt condition greater than 1.02 is provided to allow identification and correction'of a dropped or misaligned control rod. In the event such action does not correct the tilt, the margin for uncertainty of Fo is reinstated by reducing the maximum allowed power by 3% for each percent of tilt in excess of 1.
For purposes'of monitoring QUADRANT POWER TILT RATIO when one excore detector is inoperarable, the moveable incore detectors are used to confirm that the normalized symmetric powcr distribution is consistent with the QUADRANT POWER TILT RATIO. The incore detector monitoring is done with a full incore flux map or two sets of four symmetric thimbles. The two sets of four symmetric
. thimbles is a unique set of eight detector locations. These locations are C-8, E-5, E-11, H-3, H-13, L-5, L-11, N-8.
3/4.2.5 DNB PARAMETERS in MODE 1, the limits on pressurizer pressure, RCS coolant average temperature, and RCS flow rate must be maintained during steady state operation in order to ensure DNBR criteria will be met in the cvent of an unplenned loss of forced coolant flow or other DNB limiting transient. In all other MODES,
. the power level is low enough that the DNB is not a concem.
The values presented in the COLR are indicated values and include measurement uncertainties.
The value for pressurizer pressure is averaged using plant computer /QDPS readings from a minimum of at least 3 channels. The value for RCS coolant average temperature is averaged using control board readings from a minimum of at least 3 channels. The value for RCS flow rate is the average from a minimum of at least 2 flow transmitters per RCS toop using plant computer /QDPS points.
SOUTH TEXAS-UNITS 1 & 2 B 3/4 2-5 '
Unit 1 - Amendment No.37,115 Unit 2 - Amendment No,47,103
' POWER DISTRIBUTION LIMITS I
BASES
'3/4.2.5' DNB PARAMETERS (continued).
The value for thermal design RCS flow rate presented in Technical Specification 3.2.5 is the l
~ thermal design reactor coolant system flow rate used in the analysis approved by the Nuclear Regulatory Commission in Amendments 97 and 84 on September 29,1998. This flow rate is an analytical limit consistent with.10% tube plugging of the steam generator tubes and Departure from Nucleate Boiling requirements. Thermal Design RCS flow rate with Model E Steam Generators is 370,000 gpm, and with Model A94 Steam Generators it is 392,000 gpm. These values include 2.8%
measurement uncertainty.
The RCS flow measurement uncertainty of 2.8% bounds the precision heat balance and the elbow tap Ap measurement methods. The elbow tap Ap measurement uncertainty presumes that elbow tap Ap measurements are obtained from either QDPS or the plant process computer. Based on
' instrument uncertainty assumptions, RCS flow measurements using either the precision heat balance or
. the elbow tap Ap measurement methods are to be performed at greater than or equal to 90% RTP at the beginning of a new fuel cycle. The. bow tap Ap RCS flow measurement methodology is described in ST-HL-AE-5707," Proposed Amendment to Technical Specification Table 2.2-1 and 3/4.2.5 for Reactor Coolant System Flow Monitoring - Revised," dated August 6,1997, and in ST-HL-AE-5752, " Amended Response to Request for Additional Information on the Proposed Elbow Tap. Technical Specification -
Change (Table 2.2-1 and Section 3/4.2.5)," dated September 18,1997,
.The 12-hour periodic surveillance of these parameters through instrument readout is sufficient to ensure that the parameters are restorad within their limits following load changes and other expected transient operation.
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l SOUTH TEXAS-UNITS 1 & 2 B 3/4 2-6 Unit 1 - Amendment No. 445 Unit 2-Amendment No 403 l
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