ML20217M690
| ML20217M690 | |
| Person / Time | |
|---|---|
| Site: | Prairie Island  |
| Issue date: | 04/29/1998 |
| From: | NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20217M635 | List: |
| References | |
| NUDOCS 9805050084 | |
| Download: ML20217M690 (8) | |
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4 UNITED STATES l
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j NUCLEAR REGULATORY COMMISSION j*
2 WASHINGTON, D.C. 'anaan nnni i
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION REVIEW OF PRESSURE TEMPERATURE LIMITS REPORT AND METHODOLOGY FOR THE RELOCATION OF THE REACTOR COOLANT SYSTEM PRESSURE TEMPERATURE LIMIT CURVES AND LOW TEMPERATURE OVERPRESSURE PROTECTION SYSTEM LIMITS NORTHERN STATES POWER COMPANY j
PRAIRIE ISLAND NUCLEAR GENERATING PLANT. UNITS 1 AND 2 DOCKET NOS. 50-282 AND 50-306 i
1.0 INTRODUCTION
By letter dated March 6,1998 (Reference 3), and supplemented by letters dated March 6,1998 (Reference 4 ), March 30,1998 (Reference 5), March 31,1998 (Reference 6), and April 13,1998 (Reference 7), Northem States Power Company (NSP) requested changes to the technical specifications (TS) for the Prairie Island Nuclear Generating Plant, Units 1 and 2. The requested changes included (1) revising the reactor coolant system (RCS) pressure temperaNre (P/T) limit curves and low temperature overpressure protection (LTOP) system limits, (2) relocating the P/T limit curves and LTOP system limits from the TS to a licensee-controlled document identified as a Pressure Temperature Limits Report (PTLR), and (3) changing the affected limiting conditions for operation and bases accordingly. The P/T limit curves and LTOP system setpoints were developed, in part, using the staff-approved methodology documented in WCAP-14040-NP-A, l
Revision 2 (Reference 1). These changes are made in accordance with Generic Letter 96-03, l
" Relocation of the Pressure Temperature Limit Curves and ow Temperature overpressure l
Protection System Limits," dated January 31,1996 (Reference 2). Generic Letter 96-03 provides licensees the option to relocate the P/T limit curves and the LTOP system setpoints to a licensee-controlled PTLR provided that the limiting curves and setpoints are developed using an t
L NRC-approved methodology. The licensee proposes to extend the period of applicability of the P/T limit curves and LTOP system setpoints to 35 effective full power years (EFPYs) of reactor operation.
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2.0 BACKGROUND
2.1 Neutron Fluence The fluence evaluation which is the basis for the proposed revised P/T curves is documented in WCAP-14779, Revision 2, and WCAP-14613, Revision 2, for Units 1 and 2, respectively. The evaluation of the pressurized thermal shock is docuraented in WCAP-14781, Revision 3, and 9805050084 980429
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PDR ADOCK 05000282 P
PDR j
WCAP-14638, Revision 2, for Units 1 and 2, respectively. The surveillance capsule reports in WCAP-14779 and WCAP-14613 document the evaluation of capsules S and P for Units 1 and 2, respectively. In addition they document the reevaluation of previously removed capsules (V,P,R) and (V.T, R) for Units I and 2, respectively. The analyses were performed using the BUGLE-93 cross sections in the DOT computer program. The BUGLE-93 cross sections are based on the ENDF/B-VI cross section file which is the staff-recommended file. In addition the analyses utilized the Pa and Se approximations for angular elastic scattering and spacial quadrature, respectively. These approximations are recommended by the staff.
2.2 Pressure Temoerature Limits The methodologies for assessing P/T limits and reactor pressure vessel (RPV) surveillance programs are discussed, in part, in the following documents: (1) 10 CFR Part 50, " Appendix G -
Fracture Toughness Requirements"; (2) 10 CFR Part 50, " Appendix H - Reactor Vessel Material Surveillance Program Requirements"; (3) 10 CFR 50.60 " Acceptance Criteria for Fracture l
Prevention Measures for Lightwater Nuclear Power Reactors for Normal Operation"; (4) 10 CFR 50.61 " Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events"; and (5) Regulatory Guide (RG) 1.99, Revision 2 " Radiation Embrittlement of Reactor Vessel Materials."
NSP has applied the methodology of WCAP-14040-NP-A, Revision 2, as the general methodology for generating the P/T limit curves for heatup, cooldown, and hydrostatic testing conditions of the PI-1 and PI-2 (Prairie Island) reactor coolant pressure boundaries (RCPBs).
Westinghouse Electric Corporation (WEC) submitted this methodology to the staff as its basis for developing the cold overpressure mitigating system setpoints and RCPB heatup and cooldown limit curves for WEC-designed nuclear reactors.
Pursuant to 10 CFR Part 50, Appendix G, the P/T limits and minimum temperatures established for RPVs must meet the requirements for these parameters as set forth in Table 1 of the rule. In addition to the minimum requirements, the PTT limit curves are required to be at least as conservative as those that would be obtained by following the methods of analysis and the safety margins found in the 1989 Edition of the American Society of Mechanical Engineers (ASME) l Code,Section XI, Appendix G.
j 2.3 Low Temoerature Overoressure Protection System l
The licensee designated the LTOP syst9m as the Over Pressure Protection System (OPPS).
The OPPS mitigates overpressure transients at low temperatures so that the integrity of the I
RCPB is not compromised by violating the 10 CFR Part 50, Appendix G, P/T limits under steady-state operating conditions. Prairie Island Units 1 and 2 OPPS uses the pressurizer power-operated relief valves (PORVs) or an RCS vent with the reactor depressurized to i
accomplish this function. The system is manually enabled by operators and uses a single I
setpoint as the lift pressure for the PORVs. The design basis of Prairie Island Units 1 and 2 OPPS considers both mass-addition and heat-addition transients. The mass-addition analyses in the supporting PTLR account for the injection from up to three charging pumps to the RCS in the full range of P/T conditions starting from 68 *F.
For an RCS temperature greater than or equal to 200 'F, an inadvertent injection from one safety injection pump and a maximum of three charging pumps are assumed. The heat-addition analyses account for heat input from the secondary side of the steam generators into the RCS upon starting a single reactor coolant pump (RCP) when the RCS temperature is as much as 50 'F lower than the steam generator
3 secondary side temperature. The proposed TS provided restriction in plant operation within the configuration assumed in the analysis for OPPS design.
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The Prairie Island Units 1 and 2 proposed design of OPPS including the determination of its enable temperature and the PORV actuation setpoint was established using the staff-approved l_
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. methodology d' ocumented in WCAP-14040-NP-A. Also, the licensee has applied for an j'
exemption from certain requirements of 10 CFR Part 50, Appendix G, and adopted a provision in ASME Code Case N-514 that permits a 10% relaxation of the P/T limits in its design of OPPS.
I 3.0 EVALUATIONS
.3.1 Neutron Fluence The surveillance capsule reports document the measured and calculated values of four capsules J
l for each plant. The measured to calculated (M/C) ratios of the pressure vessel fluence values are consistent, and their deviation from unity is reasonable. Likewise, the different dosimeter response M/C ratios are consistent and close to unity. Overall, the measured values are slightly higher than the corresponding calculated values. The licensee adopted the calculated values (for L
each unit) for the estimation of the P/T curves. This is conservative and, therefore, it is acceptable.
The Unit 1 estimated pressure vessel inside diameter fluence value for 35 EFPYs is 3.95x10
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' n/cm. The corresponding value for Unit 2 is 4.18x10 n/cm. These values were taken into j
8 account in the P/T curves. These values were derived using staff-recommended cross sections and approximations; therefore, we find them acceptable.
3.2 Pressure Temoerature (Pm Limits 3.2.1 Revised P/T Limit Heatuo and Cooldown Curves for the PI-1 and PI-2 RPVs The staff performed an independent analysis using the methods described in 10 CFR Part 50, Appendix G and in Standard Review Plan (SRP) 5.3.2, " Pressure Temperature Limits," in order to determine whether NSP's methods for determining the minimum allowable RCS pressures and temperatures during heatup, cooldown, and hydrostatic testing conditions were conservative relative to the stafs analysis. For the stafs evaluations of the beltline materials, the staff applied the methodology found in 10 CFR Part 50, Appendix G. The stars analysis methods were consistent with those applied by NSP with the following exceptions:
The stars pressure stress equation was based on a simple hoop stress equation.
1 The sta#s method for evaluating the thermal gradient across the RPV wall was based on
=
a simple steady-state thermal gradient.
The sta#s method for determining the stress intensities due to thermal stresses was based on Figure 4-5 of the Welding Research Council (WRC) Bulletin 175.
The staff followed the criteria of the revised rule 10 CFR 50.61 and RG 1.99, Revision 2, as its basis for calculating the end of life (EOL) 1/4t and 3/4t RT,e1 values, and tne RTpts values for the beltline materials in the PI-1 and PI-2 RPVs.
. 3.2.2 Assessment of the RT Ts Values for the PI-1 and PI-2 Beltline Materials The staff performed an independent assessment of the RTp7s values for the beltline materials in j
the Prairie Island reactor vessels. The staff determined that the licensee's calculations of the RTprs values were in agreement with those that would be generated if the methods of the revised rule 10 CFR 50.61 or RG 1.99, Revision 2, were applied. For purposes of calculating the limiting RTp7 value, the staff verified that the licensee correctly calculated the limiting projected RTp7s
. values for the PI-1 and PI-2 vessels to be 162 'F and 143 'F, which are the values calculated for the PI-1 nozzle to intermediate shell forging circumferential weld and for the Pl-2 upper forging to intermediate forging seam weld W2, respectively. These values are significantly less than the screening criterion of 300 *F as stated in the revised rule 10 CFR 50.61, and indicate that the Prairie Island vessels will continue to satisfy the requirements of the rule throughout the projected lives of the plants.
3.2.3 Assessment of the EOL 1/4t and 3/4t RTyor Values and the Pronosed Heatuo. Cooldown.
and Hydrostatic Testina Curves for PI-1 and Pl-2 The staff also performed an independent assessment of the EOL 1/4t and 3/4t RTuo7 values for the beltline materials in the PI-1 and PI-2 reactor vessels, and of the proposed P/T limit curves for the PI-1 and PI-2 reactor vessels during heatup, cooldown, and hydrostatic testing conditions.
The assessment of each unit is discussed below.
For the PI-1 reactor vessel, the licensee determined that the most limiting material at the 1/4t and 3/4t locations is the nozzle to intermediate shell circumferential weld. This weld was fabricated using weld wire heat 2269. The licensee calculated an RTuo7 value of 154 'F at the 1/4t location and 136 *F at the 3/4t location at 35 EFPYs. - The neutron fluence used in the RTuor calculation was 1.47 X 10 n/cm at the 1/4t location and 0.66 X 10 n/cm at the 3/4t location. The initial 2
2 RTuor value for the limiting weld was 0 *F. The margin term used in the calculation for the limiting weld was 66 'F for both the 1/4t and 3/41 locations. This number is consistent with the number that is generated when a generic mean value is used to establish the unirradiated RTuor for a beltline weld.
The staff performed an independent calculation of the RTuo7 values for the limiting material using the methodology in RG 1.99, Revision 2. Based on these calculations, the staff verified that the licensee's limiting material for the Pl-1 reactor vessel is the nozzle to intermediate shell circumferential weld that was fabricated using weld wire heat 2269. The staff's calculated RTuor value for the limiting material agreed with the licensee's calculated RTuo7 value at 35 EFPYs.
Substituting the RTuo7 values for the PI-1 limiting weld into the equations in SRP 5.3.2, the staff
. verified that the proposed P/T limits satisfy the requirements in paragraph IV.A.2 of Appe,c. dix G of 10 CFR Part 50.
. For the PI-2 reactor vessel, the licensee determined that the most limiting material at the 1/4t
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and 3/4t locations is the upper to intermediate shell weld seam. This weld was fabricated using weld wire heat 1752. The licensee calculated an RTuoy value of 134 'F at the 1/4t location and 116 *F at the 3/4t location at 35 EFPYs. The neutron fluence used in the RTuo, calculation was 1.59 X 10 n/cm' at the 1/4t location and 0.71 X 10 n/cm at the 3/4t location. The initial RTuoy 2
value. for the limiting weld was -13 'F. The margin term used in the calculation for the limiting weld was 56 'F for both the 1/4t and 3/4t locations. This number is consistent with the number that is generated when a generic mean value is used to establish the unirradiated RTuoy for a beltline weld.
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The staff performed an independent calculation of the RTuoy values for the limiting material using the methodology in RG 1.99, Revision 2. Based on these calculations, the staff verified that the licensee's limiting material for the PI-2 reactor vessel is the upper to intermediate shell weld seam that was fabricated using weld wire heat 1752. The staffs calculated RTuo1 value for the limiting material agreed with the licensee's calculated RTuoy value at 35 EFPYs. Substituting the RTuo, values for the Pl-2 limiting weld into the equations in SRP 5.3.2, the staff verified that the proposed P/T limits satisfy the requirements in paragraph IV.A.2 of Appendix G of 10 CFR Part 50.
In addition to beltline materials,' Appendix G of 10 CFR Part 50 also imposes a minimum temperature at the closure head flange based on the reference temperature for the flange material.Section IV.A.2 of Appendix G states that when the pressure exceeds 20% of the preservice system hydrostatic test pressure, the temperature of the closure flange regions highly
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stressed by the bolt preload must exceed the reference temperature of the material in those regions by at least 120 'F for normal operation and uy 90 'F f - hydrostatic pressure tests and leak tests. The RTuor values for the limiting flange materials in W PI-1 and PI-2 reactor vessels are -4 'F and -22 'F, respectively. The staff has determined that the proposed P/T limits satisfy the requirement for the closure flange region during normal operation and hydrostatic pressure tests and leak tests.
3.3 Low Pressure Overoressure Protection System The proposed Limiting Conditions for Operation in TS 3.1.A.2.c require that an OPPS be enabled l
with two operable PORVs when the RCS temperature is below the OPPS enable temperature.
Also, (1) when the RCS temperature is above the temperature in which the safety injection pumps are not disabled, one PORV may be inoperable for 7 days. If these conditions cannot be met, the RCS must be depressurized and vented through at least a 3-square inch vent within the next 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. In the case where both PORVs become inoperable, the RCS must be depressurized and vented through at least a 3-square inch vent within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, and (2) when the RCS is below the temperature for disabling both safety injection pumps, one PORV may be inoperable for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. If these conditions cannot be met, the RCS must be depressurized and vented through at least a 3-square inch vent within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. In the case where both PORVs become inoperable, the RCS must be depressurized and vented through at least a 3 square inch vent within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. The astpoints related to the design of OPPS and applicable to both Prairie Island Units 1 and 2 are listed in the licensee's PTLR. The staff evaluation of these setpoints is presented below.
3.3.1 Enable Temperature The OPPS enable temperature is the temperature below which the OPPS is required to be operable. The licensee has established an OPPS enable temperature using the methodology presented in WCAP-14040-NP-A with the provision permitted by ASME Code Case N-514. This Code Case requires the OPPS to be effective at an RCS temperature less than 200 'F or at an RCS temperature corresponding to a reactor vessel metal temperature less than RTuor + M 7 at the beltline location (1/4t). Therefore, the licensee proposed to calculate the enable temperature as RTuo7 + 50 'F + temperature difference between RCS and metal + Instrument Uncertainties. Using the above equation with limiting material adjusted reference temperature for Unit 1 as input, the calculated minimum enable temperature applicable for both Prairie Island Units 1 and 2 is 243 'F. The licensee proposed an enable temperature of 310 'F that includes an additional margin of 67 'F.
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The staff finds that this proposed OPPS enable temperature is conservative with respect to the I
enable temperature allowed by ASME Code Case N-514 and therefore is acceptable.
I 3.3.2 Disablina Safety Iniection Pumo(s) and Isolatina Accumulators in the licensee's analysis for the design of OPPS, it is assumed that wher, (CS temperature is below 200 'F, the OPPS will provide adequate protection for a mass ad-
,n from a maximum of three charging pumps. When the RCS is between 200 'F and the calcL,ated OPPS enable temperature, the OPPS will provide protection for a mass addition from one safety injection pump plus three charging pumps. To support these analysis assumptions, the licensee proposed requirements in TS 3.3.A and the PTLR to disable one safety injection pump when the RCS is above 218 *F and to disable two safety injection pumps when the RCS temperature is below 218 'F. The setpoint of 218 'F includes an instrument uncertainty of 18 'F.
- Also, TS 3.3.A requires that both accumulators be isolated when the RCS temperature is below the OPPS enable temperature.
The licensee in its letter dated March 31,1998, indicated tnat the safety injection pumps are rendered incapable of injecting into the RCS by employing at least two independent means to prevent a pump start such that a single action will result in an injection into RCS. This may be accomplished through the pump control switch being placed in pullout with a block device installed over the control switch that would prevent an unplanned start. This method of disabling the safety injection pump has been stated in the TS Bases 3.3. We find that the licensee-proposed method of disabling the safety injection pump is acceptable.
1 3.3.3 PORV Actuation Setooint OPPS is designed to mitigate overpressure transients at low temperatures to prevent violating 10 CFR Part 50, Appendix G, P/T limits. Additionally, since overpressure events most likely occur during isothermal conditions in the RCS, the NRC has accepted the use of the steady-state Appendix G limits for the design of the OPPS. The OPPS actuation setpoint is the pressure at which the PORVs will lift, when the OPPS is enabled, to limit the peak RCS pressure during a pressurization transient.
Prairie Island Units 1 and 2 use PORVs to provide pressure relief capacity for the OPPS. The methodology used for determining the PORV actuation setpoint is consistent with the methodology presented in WCAP-14040-NP-A.
The licensee-proposed PORV actuation setpoint of 500 psig in the PTLR was calculated in accordance with the proposed methodology. The licensee, in its submittal dated March 6,1998, provided a tabulation listing PORV setpoints, transient pressure overshoot, instrumentation uncertainties, pressure difference between the pressure transmitter and the reactor vessel mid-plane with one or two RCPs in operation and corresponding P/T limits under various temperature conditions below the OPPS enable temperature. The data presented in this tabulation confirms that the proposed PORV setpoint of 500 psig will provide adequate protection to the P/T limits established by 10 CFR Part 50, Appendix G, with the provision of ASME Code Case N-514 under steady-state conditions during a design-basis overpressure transient (mass-addition or heat-addition) as described in Section 1.0 of this report. Based on the above discussion, we find the proposed PORV setpoint acceptable.
7-3.3.4 RCS Vent Size The proposed TS 3.1.A.2.c specifies a vent size of 3 square inches as an alternative to an operable OPPS when the RCS is depressurized, The bases for the 3-square inch vent is stated in the TS Bases, page B.3.1.-3. It states that the vent size is based on the 2.956-square inch cross sectional flow area of a pressurizer PORV. Since the vent size is compatible with the PORV size, which is sufficient to mitigate a design-basis overpressure transient, we find it acceptable.
4.0 CONCLUSION
S Based upon the staff evaluations, as discussed in Section 3.0 above, the NRC staff concludes that it is acceptable for NSP to relocate the PR limit curves and LTOP system limits from the Prairie Island Units 1 and 2 TS to a licensee-controlled PTLR. The proposed heatup, cooldown, and hydrostatic testing curves for Pl-1 and PI-2 will expire at 35 EFPYs.
The staff has reviewed the proposed fluence values for Prairie Island Units 1 and 2 for the revision of the PR curves and finds that the proposed values are conservative and, therefore, acceptable. -In addition, the staff has determined that the proposed PU limit curves are acceptable for use and are consistent with the requirements of Appendix G to 10 CFR Part 50, and Appendix G to Section XI of the ASME Code.
However, since WCAP-14040-NP-A does not address the credibility of RPV surveillance material, the licensee, in future PU limit evaluations, shoeld address the credibility of the surveillance material.
The staff also reviewed the licensee's analyses related to the proposed setpoints of the OPPS as discussed in Section 3.0 above. The licensee has considered instrument uncertainties in its setpoint calculation using Instrument Society of America S67.04-1994. The staff finds that the licensee's analyses were performed in a manner consistent with the approved methodology and that the results of the analyses conservatively demonstrated that the P/T limits established by 10 CFR Part 50, Appendix G, with provisions provided by ASME Code Case N-514 will be adequately protected with these setpoints and, therefore, finds NSP's analyses acceptable.
The staff has determined that the proposed PTLR meets the criteria of Generic Letter 96-03, and is acceptable to the staff.
5.0 REFERENCES
1.
WCAP-14040-NP-A, Revision 2, Westinghouse Electric Corporation, " Methodology Used to Develop Cold Overprassure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves," January 15,1996.
1 2.
NRC Generic Letter 96-03, " Relocation of the Pressure Temperature Limit Curves and Low Temperature Overpressure Protection System Limits," January 31,1996.
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8-l 3.
Letter from Joel P. Sorensen, Northern States Power Company, to NRC Document Control l
Desk, " Amendment of Technical Specifications to Update the Heatup and Cooldown Rate Curves, incorporate the Use of a Pressure Temperature Limits Report, and Change the Pressurizer Power Operated Relief Valves Operability Temperature," March 6,1998.
(WCAP-14780 and WCAP-14637 are Attached).
4.
Letter from Joel P. Sorensen, Northem States Power Company, to NRC Document Control Desk, " Revised Prairie Island Units 1 and 2 Reactor Vessel Material Surveillance Reports,"
March 6,1998. (WCAP-14779, Revision 2, WCAP-14781, Revision 3, WCAP-14613, Revision 2, and WCAP-14638, Revision 2 are Attached).
5.
Letter from Joel P. Sorensen, Northern States Power Company, to NRC Document Control Desk, " Response to March 16 and 19,1998 Requests for Additional Information for License Amendment Request dated March 6,1998, " March 30,1998.
- 6.
Letter from Joel P. Sorensen, Northern States Power Company, to NRC Document Control Desk, " Response to March 13,1998 Request for Additional Information for License Amendment Request dated March 6,1998, " March 31,1998.
7.
Letter from Joel P. Sorensen, Northern States Power Company to NRC Document Control Desk, " Supplement to the License Amendment Request dated March 6,1998," April 13, 1998.
Principal Contributors: M. Khanna C.Y. Liang L. Lois M. W. Weston Date: April 29, 1998 l
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