ML20216D022
| ML20216D022 | |
| Person / Time | |
|---|---|
| Site: | Prairie Island  |
| Issue date: | 03/06/1998 |
| From: | Sorensen J NORTHERN STATES POWER CO. |
| To: | NRC |
| Shared Package | |
| ML20013F498 | List: |
| References | |
| NUDOCS 9803160261 | |
| Download: ML20216D022 (14) | |
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION NORTHERN STATES POWER COMPANY PRAIRIE ISLAND NUCLEAR GENERATING PLANT DOCKET NO.
50-282 50-306 License Amendment Request dated March 6,1998 Amendment of Technical Specifications to Update the Heatup and Cooldown Rate Curves, Incorporate the Use of a Pressure and Temperature Limits Report, and Change the Pressurizer Power Operated Relief Valves Operability Temperature Northern States Power Company, a Minnesota corporation, with this letter is i
submitting information to support a requested license amendment.
This letter contains no restricted or other defense information.
l NORTHERN STATES POWER COMPANY 1
By u-
[ Joel P Sorensen Plant Manager Prairie Island Nuclear Generating Plant On this the y of tA
/97Iefore me a notary public in and for said county, personally appearelhfobl P Sorensen, Plant Manager, Prairie Island Nuclear Generating Plant, and being first duly sworn acknowledged that he is authorized to execute this document on behalf of the Northern States Power Company, that he knows the contents thereof, and that to the best of his knowledge, information, and belief the statement mad in it ru nd that it is ' ot interposed for delay.
t 1/
l MARCIA K. lac 0fE NOTARYPUBUC4dME3011 i
HENNEPIN COUNTY I
90o3160261 900306 DR ADOCK 050 2
l EXHIBIT A PRAIRIE ISLAND NUCLEAR GENERATING STATION License Amendment Request dated March 6,1998 Evaluation of Proposed Changes to the Technical Specification Appendix A of Operation License DPR-42 and DPR-60 Pursuant to 10 CFR Part 50, Sections 50.59 and 50.90, the holders of Operating Licenses DPR-42 and DPR-60 hereby propose the following changes to the Technical Specifications contained in Appendix A of the Facility Operating Licenses:
BACKGROUND Technical specifications currently include limiting conditions for operation (LCOs) that establish pressure temperature (P/T) limits and low temperature overpressure protection (LTOP)' system limits for the reactor coolant system (RCS). Generic letter 96-03 *, provides licensees the option to relocate these limitations into a licensee-controlled " Pressure Temperature Limits Report (PTLR) provided that the limiting curves and setpoints are developed using an NRC approved methodology. This change will allow the licensee to maintain these curves and setpoints more efficiently and at a lower cost. Periodic adjustments to the curves and setpoints based on neutron irradiation of the reactor vessel or changes in instrument uncertainty will be made under the conditions of 10CFR50.59, and the updated PTLR will be submitted to the NRC upon issuance. Changes to the curves, setpoints or parameters in the PTLR resulting from new or additional analysis of either beltline or weld material properties will be submitted to the NRC prior to issuance of an updated PTLR, As part of the reactor vessel material surveillance program, a reactor vessel material surveillance capsule was removed for testing from Unit 1 during the January 1996 refuel outage and from Unit 2 during the May 1995 refuel outage. The results of this testing provided the basis for calculation of new P/T limit curves and OPPS setpoints. These revised curves and i
' A terminology difference exists between the documents from different organizations.
The NRC Generic Letter 96-03 refers to a Low Temperature Overpressure Protection System (LTOP),
which is equivalent to both the Prairie Island Over Pressure Protection System (OPPS) and the Westinghouse Cold Overpressure Mitigation System (COMS).
- GL 96-03," Relocation of the Pressure Temperature Limit Curves and Low Overpressure Protection System Limits", January 31,1996 l
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Exhibit A l
Page 2 l
setpoints were determined using a NRC approved methodology and incorporated into the d
attached PTLR. The current OPPS setpoints and technical specification P/T limits are only analyzed up to a reactor vessel fluence limit of 20 EFPY, which Unit 1 is projected to reach by May 1,1998.
Results of the Unit 1 and 2 reactor vessel capsule withdrawal and specimen tests were
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previously submitted to the NRC for evaluation. These tests have been reanalyzed 8 and are 5
the basis for the proposed revisions to the P/T limit curves and OPPS setpoints. The limits 7
calculated from the material test data for Unit 1 bound the calculated limits from Unit 2.8 To simplify operations, these bounding limits are incorporated into the PTLR as a single set of P/T limit curves and OPPS setpoints applicable to both units.
The determination of the OPPS setpoints' used ASME Code Case N-514. This permits the OPPS pressure relief setpoint to be established such that the maximum pressure at the reactor vessel material's most limiting location is limited to 110% of the pressure determined to satisfy ASME Section XI Appendix G Article G-2215 limits. This ASME code case has not yet been approved for generic use by the NRC and an exemption request has been submitted to receive approval for its application in this case."
A change to the temperature at which the Pressurizer Power Operated Relief Valves (PORVs) must be operable, is being submitted to make it consistent with NUREG 1431," Standard Technical Specifications for Westinghouse Plants." The change to the temperature for PORV operability from 310 F to 350 F is consistent with a value that was previously contained in the PI Technical Specifications. Previously the temperature was 350 F and was changed to 310 F to be coincident with the OPPS enable temperature for operational simplicity. This was approved by the NRC staff in License Amendment 91/84 dated October 27,1989.
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- WCAP-14040-NP-A Rev 2," Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves", January 1996.
- Exhibit D of this submittal.
8 WCAP-14613 Rev 2," Analysis of Capsule P from the Northern States Power Company Prairie Island Unit 2 Reactor Vessel Radiation Surveillance Program", February 1998.
- WCAP-14779 Rev 2," Analysis of Capsule S from the Northern States Power Company Prairie Island Unit i Reactor Vessel Radiation Surveillance Program", February 1998
' WCAP-14780 Rev 3," Prairie Island Unit 1 Heatup and Cooldown Limit Curves for Normal Operation",
February 1998.
- WCAP-14637 Rev 2, " Prairie Island Unit 2 Heatup and Cooldown Limit Curves for Normal Operation, February 1998.
- Westinghouse Letter NSP-98-0120 Rev 2," Prairie Island Units 1 and 2 COMS Setpoint Analysis",
February 1998.
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"
- Request for Exemption from the Requirements of 10CFR50.60", letter from Joel P. Sorensen to the NRC, January 15,1998, as supperceded March 6,1998
I Exhibit A Paga 3 PROPOSED CHANGES AND REASONS FOR CHANGES:
i TS 1.0 Definitions The definition contained in Generic Letter 96-03 for the term Pressure and Temperature Limits Report (PTLR) is added with two exceptions. The PTLR will not be unit specific for Prairie Island, and the Prairie Island specific terminology, Over Pressure Protection System (OPPS),
will be used in place of the Generic Letter terminology, Low Temperature Over-pressure Protection System (LTOP).
This change will provide for a common understanding of the contents and use of the PTLR.
TS 3.1.A.1.c(4)
Reactor Coolant Loops and Coolant Circulation TS 3.1.A.2.c(2)
Pressurizer Power Operated Relief Valves TS 3.3.A.3 Safety injection and Residual Heat Removal Systems Table TS.4.1-1c Miscellaneous Instrumentatlcn Surveillance Gequirements The temperature limit,"310 F*", has been replaced with the wording, "the Over Pressure Protection System Enable Temperature specified in the PTLR."
The OPPS enable temperature value and the associated reactor fluence limit have been relocated to the Pressure and Temperature Limits Report.
TS 3.1.A.2.c(2)
Pressurizer Power Operated Relief Valves TS 3.1.A.2.c(3)
Pressurizer Power Operated Relief Valves TS 3.3.A.4 Safety injection and Residual Heat Removal Systems The temperature limit, "200'F", has been replaced with the wording, "the temperature specified in the PTLR for disabling both safety injection pumps."
The mass addition transient calculation in OPPS Setpoint Analysis used 200 F for an analytical limit and assumed that only three charging pumps and no SI pumps would be available to inject into the RCS when below 200 F. This analyticallimit has been relocated to the Pressure and Temperature Limits Report.
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Exhibit A Paga 4 TS 3.1.A.2.c Pressurizer Power Operated Relief Valves
[ Affected subparagraphs: (1), (1)(a), (1)(b), and (1)(b)3 ]
The temperature dependency for PORV LCOs has been changed from "310 F*" to "350 F".
The reference to the reactor fiuence limit associated with the OPPS Enable Temperature has been removed. The 350 F temperature value is not related to the brittle fracture limitations on the reactor vessel material.
I This temperature is used to establish the requirements for PORV operability and LCOs that are consistent with plant operations for temperatures greater than those required for low temperature over pressure protection. In Amendment 91/84 this temperature was changed from 350 F to be coincident with the OPPS Enable Temperature of 310 F for operational flexibility. Changing this temperature back to 350 F will be consistent with the temperature in NUREG 1431, Standard Technical Specifications, Westinghouse Plants. This provides for a 40 F temperature band to permit fining up the PORVs from operation with the OPPS enabled (RCS < 310 F) to operation in the normal pressure relief mode (RCS > 350 F).
TS 3.1.B.1.a Pressure / Temperature Limits Changed wording to require that the RCS temperature and pressure limits and heatup and cooldown rates shall be maintained within the limits specified in the PTLR. Deleted the specific wording identifying maximum heatup and cooldown rates.
The pressure and temperature limits and heatup and cooldown rates are determined using the methodology of WCAP-14040-NP-A. These limits are contained in the PTLR.
Figure TS.3.1-1 Unit 1 and Unit 2 Reactor Coolant System Heat Up Limitations Figure TS.3.12 Unit 1 and Unit 2 Reactor Coolant System Cooldown Limitations These figures have been deleted.
Revised figures are provided in the PTLR.
TS 3.3.A.1.c Safety injection and Residual Heat Removal Systems l
Reference to the Over Pressure Protection System Enable Temperature of *310 F*" has been removed. Reference to TS 3.1.A.d(2) has been removed.
TS 3.1.A.d(2) does not address itself to safety injection pump control switch position so inclusion of a reference to it is confusing and superfluous. References to TS 3.3.A.3 and 1
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Exhibit A Pago5 3.3.A.4 are left to stand alone. TS 3.3.A.4 does not apply to RCS temperatures that are less than the Over Pressure Protection System Enable Temperature of"310 F*" but applies to RCS I
temperatures that are less than the temperature specified in the PTLR for disabling both safety injection pumps. TS 3.3.A.4 contains a cross reference to 3.1.A.d(2)
TS 3.3.A.5 Safety injection and Residual Heat Removal Systems Added the requirements that the reactor coolant system accumulators are isolated from the RCS whenever RCS temperature is less than the OPPS enable temperature. This specification will not apply whenever the accumulator is depressurized or the reacotr vessel head is removed.
This provides assurance that a condition assumed in the mass injection transient analysis will not be violated.
TS 6.7.A.7 Reactor Coolant System Pressure and Temperature Limits Report Added the requirements for the PTLR and reporting in accordance with the guidance of Generic Letter 96-03.
This ensures that the methodology used for pressure and temperature limits and OPPS setpoints are those previously approved by the NRC. Provides a minimum submittal criteria and the minimum requirements for information content.
I Basis 3.1.A Reactor Coolant System Operational Components Basis 3.1.B Reactor Coolant System Pressure / Temperature Limits Basis 3.3 Engineered Safety Features The bases for Specifications 3.1.A 3.1.B and 3.3 are revised in accordance with the changes made in the specifications as stated above. The changes to the bases are shown in Exhibit B.
Exhibit A Paga 6 SAFETY EVALUATION The current Prairie Island Technical Specifications include P/T limits and heatup and cooldown rates for operation of the RCS to prevent brittle failure of the reactor vessel. These limitations are derived analytically based on limiting materials in the reactor vessel beltline region and are modified periodically based on a decrease in the reactor vessel material toughness due to embrittlement caused by neutron irradiation. These periodic modifications are required at the end of the reactor vessel fluence period or when new reactor vessel material surveillance program data becomes available.
As required by 10CFR50 Appendix G, the operating P/T limits are calculated and form the basis of operations to ensure that the fracture toughness requirements for the Reactor Coolant Pressure Boundary are maintained. In addition,10CFR50 Appendix H requires that irradiated j
capsules of reactor vessel beltline material specimens be periodically withdrawn from inside the i
reactor vessel and tested to provide measured data on the effects of radiation fluence. The specimen data is applied, in part, to establish new P/T limits necessary to compensate for the increase in the nil-ductility transition reference temperature, RTwor. This ensures that the reactor vessel is operated at conditions to preclude brittle fracture.
Appendix H of 10CFR50 and ASTM 185 specify the material property tests that must be carried out on the reactor vessel material surveillance capsule specimens. The results of these tests are used to determine the brittle fracture P/T limit curves for steady state, heatup and cooldown. Heat injection and mass injection events are modeled using a system transient analysis computer code to identify OPPS setpoints that will prevent the RCS pressure from exceeding the steady state P/T limits during these events. The analysis methodology to determine the operating P/T limit curves and the OPPS setpoints is described in WCAP-14040-NP-A. The NRC has reviewed ti,a methodology of WCAP-14040-NP-A against the requirements of 10CFR50 Appendix G,10CFR50 Appendix H, Regulatory Guide 1.99 Rev 2, ASME Code Section XI Appendix G, and the Standard Review Plan section 5.3.2 and issued a Safety Evaluation Report (SER) that identifies this methodology as acceptable for application to the determination of P/T limits and OPPS setpoints.
The determination of Prairie Island specific P/T limit curves and OPPS setpoints to a reactor vessel fluence limit of 35 EFPY involved the following notable conservatisms and items:
- 1. The approved methodology of WCAP-14040-NP-A was used in conjunction with the calculation relaxation permitted by ASME Code Case N-514. (This code case has not yet been approved for generic use by the NRC and an exemption request has been submitted to permit its application.) Although Prairie Island could avoid the use of ASME Code Case N-514 by reducing the OPPS pressure relief setpoint, use of the 500 psig setpoint is
Exhibit A l
Pags 7 beneficialin providing added safety to the operation of the nuclear plant, in excess of the slight reduction in the margin relative to the limits set by Appendix G. This is based on the following reasons:
- a. Reducing the OPPS pressure setpoint below 500 psig would increase the probability that the RCP's #1 seal will fail as a result of OPPS operation. Tripping an RCP would l
reduce the equipment available to provide for decay heat removal.
The PORV takes several seconds to stroke open or closed, so the pressure relief from l
OPPS operation is not instantaneous. Because of this, the RCS pressure response to each mass injection transient and heat injection transient has an overshoot and an undershoot. With an OPPS setpoint of 500 psig neither the heat injection transients' nor the mass injection transients' resultant pressure overshoot will violate the "ASME l
Appendix G with ASME Code Case N-514" brittle fracture limits curve, and the challenge from the resultant pressure undershoot to the RCP #1 seal required minimum differential pressure will be minimized.
- b. Reducing the OPPS pressure setpoint below 500 psig would reduce the available pressure operating band and thereby increase the probability that OPPS will be inadvertently initiated by pressure fluctuations induced by typical system evolutions (e.g.
stading a Reactor Coolant Pump, putting the Residual Heat Removal system into service, shifting operating charging pumps, etc.).
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- c. In the development of the Appendix G P/T curves there are numerous conservatisms:
l (1) A factor of safety of 2.0 on the primary membrane (pressure) stresses.
(2) A margin factor shift of two standard deviations applied to RTwo7 (3) All dynamic crack initiation and crack arrest toughness experimental data is bounded i
by the use of the ASME section ill and XI Appendix G reference stress intensity curves (Kir). The use of the reference stress intensity curve bounds the crack l
initiation fracture toughness (Kic) properties by a factor of 1.2 to 2.5, depending on l
vessel temperature and RTwor.
- 2. The historical value of 310'F for the OPPS Enable Temperature was used in place of the lower calculated value of 243 F, which results from the summation of the analytical limit of 225 F and the indicating instrument channel uncertainty of 18 F. Analysis has shown that an OPPS Enable Temperature of 310 F with the constant value OPPS pressure relief setpoint of 500 psig will protect the 10CFR50 Appendix G brittle fracture P/T limits as
l Exhibit A I
Pago 8 modified by the ASME Code Case N-514 from the minimum unvented RCS temperature through rated operating temperature.
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- 3. One set of P/T Limit curves is used for both reactor units. This set of P/T Limit curves is based on the most limiting material at either unit at both the 1/4 and 3/4 reactor vessel thickness.
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- 4. A conservative neutron fluence was used in the determination of the adjusted reference temperatures (ARTS), such that, where the calculated fluence was greater than the best estimate fluence, the calculated fluence was applied. This approach increased the ART values for Unit 1 by approximately 0.3%. This is an additional conservatism that is in excess of that mandated by current regulations and does not constitute a future methodological commitment.
- 5. In the determination of the adjusted reference temperatures (ARTS), a chemistry factor that is conservative in excess of that mandated by current regulations was applied. Regulations l
provide for one of two different margin terms to be added in the calculation of the ARTS, depending on surveillance data credibility. The larger margin term is reserved for use with a l
generic chemistry factor, which is applied when measured surveillance data is either l
unavailable or deemed to be "non-credible". If surveillance data is deemed to be credible, then the ART is calculated with a chemistry factor based on the surveillance data and a l-smaller margin term. Where surveillance data was available, Prairie Island calculated ARTS by applying the large margin term and the largest chemistry factor term produced by either the "non-credible" surveillance data or the Regulatory Guide 1.99 Rev. 2 generic chemistry factor table. This approach increased the ART value of the Unit 1 Intermediate to Lower Shell Forging Circumferential (W-3)" weld by approximately 11%. This is an additional conservatism that is beyond current approved methodology and does not constitute a future methodological commitment.
- 6. Although the Unit 1 and Unit 2 Nozzle to Intermediate Shell Forging Circumferential (W-2) welds were not within the scope of the reactor vessel material surveillance programs according to the standards applicable at the time the program was established, they are now being treated as within scope. Because a measured value for the initial unirradiated RTuor is not available for the Unit 1 weld material, a generic value for the initial RTwor has been applied in accordance with SRP S.3.2. This generic value is conservative relative to the actual value that would be expected to be obtained by a measurement. Further additional margin was applied to the calculation of the adjusted reference temperature (ART). Application of this generic value and additional margin has resulted in this Unit 1 weld being the most limiting material.
" The W-2 weld is the most limiting material because generic material properties are being applied in the calculation of its adjusted reference temperature (ART).
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Exhibit A Pcg3 9 i
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- 7. Instrument uncertainties are not included in the P/T curves or the Boltup Temperature identified in the PTLR. The instrument uncertainties for these parameters will be included in the operating procedures where these are utilized. The RCS Minimum Temperature When Not Vented identified in the PTLR does include instrument uncertainties. In the Prairie Island specific calculations the Boltup Temperature value is distinct from the RCS Minimum Temperature When Not Vented temperature value, which is the lower bound of the mass injection analysis upon which the OPPS setpoint is based.
Incorporation of instrument uncertainties into P/T limits and OPPS setpoints provides assurance that operating conditions are maintained in strict accordance with analytical assumptions. Setpoint changes to accommodate instrument uncertainty do not constitute a change to plant operation that creates an unanalyzed condition. For example, placing the control switches for both SI pumps in " pull out" at an RCS temperature greater than the analytical temperature limit of 200 F.
The relocation of the pressure and temperature limits, the heatup and cooldown rates, and OPPS setpoints into the PTLR does not eliminate the requirement to operate in accordance i
with the requirements of 10CFR50 Appendix G. The requirement to operate within these restrictions is still identified in Technical Specifications. The methodology used in determining the figures, values, and parameters has been approved by the NRC and changes to these items are now administratively controlled in Technical Specifications.
i The PORVs have two principle roles: OPPS and automatic RCS pressure control. For RCS 1
temperatures below the OPPS enable temperature of 310 F the PORVs are required to protect the RCS from pressure transients at low temperatures that could exceed the limits of 10CFR50 Appendix G. For RCS temperatures above 350 F the PORVs are required to be available to permit automatic control of RCS pressure and prevent challenges to the pressurizer safety valve setpoints. The PORVs do not perform any automatic safety related actions, although PORVs in the AUTO mode will actuate before the pressurizer safety valves to relieve RCS pressure increases. For RCS temperatures below 350 F both the pressure and core energy i
are sufficiently decreased that pressure surges become less significant. For RCS temperatures below 350 F the RHR system is capable of removing the reactor decay heat and thereby controlling RCS pressure and temperature. In the unlikely event that a significant pressure surge were to occur in this temperature range with neither RHR nor the PORVs in service, one pressurizer safety valve would be operable to mitigate potential overpressure transients.
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Based on the above discussion, Northem States Power Company believes there is reasonable assurance that the health and safety of the public will not be adversely affected by these proposed changes.
i I
i Exhibit A L
Pag)10 DETERMINATION OF SIGNIFICANT HAZARDS CONSIDERATIONS
- 1. The proposed amendment will not involve a significant increase in the probability or consecuences of an accident previousiv evaluated.
The proposed change to update the Prairie Island pressure and temperature limits curves l
and the Over Pressure Protection System (OPPS) setpoints for reactor vessel fluence to j
35 EFPY is based upon measurements and calculations that were performed in accordance with an NRC approved methodology for performing reactor vessel fracture analysis to meet 10CFR50 Appendix G and H requirements. These calculations made application of American Society of Mechanical Engineers (ASME) Code Case N-514, " Low Temperature Overpressure Protection", in determining the acceptable OPPS setpoint for -
Prairie Island Units 1 and 2. This permits the OPPS pressure relief setpoint to be established such that the maximum pressure at the reactor vessel material's most limiting j
location is limited to 110% of the pressure determined to satisfy ASME Section XI, l
Appendix G, Article G-2215. As detailed in the exemption request to apply this ASME l
Code Case, the development of the Appendix G pressure / temperature limit curves incorporates numerous conservatisms. For this reason the ASME code committee approved this code case. Application of this code case with the approved methodology l
does not produce a significant increase in the probability or magnitude of brittle fracture of l
the reactor vessel.
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The proposed change to relocate the pressure and temperature limits curves and the Over l
Pressure Protection System (OPPS) setpoints to a Pressure and Temperature Limits I
Report is an administrative change. It does not affect any system which is a contributor to initiating events for previously evaluated anticipated operational occurrences and therefore does not involve any significant increase in the probability or consequence of an accident previously evaluated.
l' The proposed change in PORV operability temperature from 310'F to a new value of 350*F does not affect any system which is a contributor to initiating events for previously evaluated anticipated operational occurrences and therefore does not involve any significant increase in the probability or consequence of an accident previously evaluated.
- 2. The proposed amendment will not create the possibility of a new or different kind of accident from any accident Dreviously analvZed.
The proposed change to update the Prairie Island pressure and temperature limits curves and the Over Pressure Protection System (OPPS) setpoints for a reactor fluence limit of 35 EFPY does not introduce a new mode of operation or testing, or make physical changes
Exhibit A Pcge 11 to the plant. (The new Technical Specification requirement to isolate the accumulators whenever the RCS temperature is less than the OPPS enable temperature does not introduce a new mode of operation since Unit Shutdown procedures close the accumulator discharge valves and tag out their breakers when RCS pressure falls below 1000 psig.)
The general methods employed to develop this change are well understood and have been previously reviewed and approved. Updating the operating restrictions, OPPS setpoints, and reactor fluence limit for operation do not create a possibility of a new or different kind l
of accident from those previously analyzed.
l The proposed change to relocate the pressure and temperature limits curves and the Over Pressure Protection System (OPPS) setpoints to a Pressure and Temperature Limits l
Report is an administrative change. The proposed change does not alter the design, l
function, or operation of any plant component, therefore a possibility of a new or different l
kind of accident from those previously analyzed has not be created.
The proposed change in the PORV operability temperature from 310 F to a new value of l
350'F does not involve a physical alteration of the plant. Since no new or different type of equipment will be installed, this change does not create the possibility of a new or different i
kind of accident from any accident previously evaluated.
- 3. The orooosed amendment will not involve a sianificant reduction in the maroin of safety.
Although neutron irradiation reduces the material fracture toughness of the reactor vessel, deterministic analyses have demonstrated that proposed P/T curves, OPPS setpoints and reactor vessel fluence limits for operation will preserve the required margin of safety in the RCS boundary during postulated low temperature pressurization events.
The proposed change to use the PTLR is administrative in nature and does not impact the j
operation of the Prairie Island Nuclear Generating Plant in a manner that would result in any significant reduction in any margin of safety.
l The proposed change in the PORV operability temperature from 310 F to a new value of l
350'F does not impact any systems that are relied upon for core cooling or RCS pressure relief at RCS temperatures below 350 F. Setting the PORV operability temperature back to 350* aligns the PORVs with the Pressurizer Safety Valve operability requirement so the PORVs are still available to limit challenges to the Pressurizer Safety Valve settings during conditions of higher RCS pressure and energy when pressure surges become more significant. (In Amendment 91/84 this temperature was changed for operational flexibility from its previous value of 350 F to a new value of 310 F to be coincident with the OPPS Enable Temperature. This change was not done to establish a larger margin of safety.)
For RCS temperatures below 350 F both the pressure and core energy are sufficiently
Exhibit A Pcga 12 decreased that pressure surges become less significant. For RCS temperatures below 350*F the RHR system is capable of removing the reactor decay heat and thereby controlling RCS pressure and temperature. In the unlikely event that a significant pressure surge wo,e to occur in this temperature range with neither RHR nor the PORVs in service, one pressurizer safety valve would be operable to mitigate potential overpressure transients. Thus this change does not involve a significant reduction in the margin of i
safety associated with either the RCS boundary or fuel cladding.
Based on the evaluation described above, and pursuant to 10 CFR Part 50, section 50.91, Northem States Power Company has determined that operating the Prairie Island Nuclear Generating Plant in accordance with the proposed license amendment request does not involve any significant hazards considerations as defined Nuclear Regulatory Commission regulations in 10 CFR. Part 50 section 50.92.
ENVIRONMENTAL ASSESSMENT i
Northem States Power Company has evaluated the proposed change and determined that:
- 1. The change does not involve a significant hazards consideration, l
- 2. The change does not involve a significant change in the types or significant increase in the amounts of any effluents that may be released offsite, and
- 3. The change does not involve a significant increase in individual or cumulative occupational radiation exposure.
i Accordingly, the proposed changes meet the eligibility criterion for categorical exclusion set forth in 10 CFR part 51 section 51.22(c)(9). Therefore, pursuant to 10 CFR Part 51 section 51.22(b), an environmental assessment of the proposed changes is not required.
f.