ML20117J085
| ML20117J085 | |
| Person / Time | |
|---|---|
| Site: | Prairie Island  |
| Issue date: | 05/21/1996 |
| From: | NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20117J082 | List: |
| References | |
| NUDOCS 9605300150 | |
| Download: ML20117J085 (5) | |
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'i UNITED STATES j
NUCLEAR REGULATORY COMMISSION 2
WASHINGTON, D.C. 20056 0001
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NOS.123 AND 116 TO FACILITY OPERATING LICENSE NOS. DPR-42 AND DPR-60 NORTHERN STATES POWER COMPANY PRAIRIE ISLAND NUCLEAR GENERATING PLANT. UNIT NOS. 1 AND 2 DOCKET NOS. 50-282 AND 50-306
1.0 INTRODUCTION
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By application dated May 4,1995, as supplemented November 27, 1995, and March 1, 1996, Northern States Power Company, the licensee for the Prairie Island Nuclear Generating Plant (PI), requested a Technical Specification (TS) change that vould raise the as-found tolerance of the pressurizer and main steam safety valves (MSSVs) from i I percent to i 3 percent. The TS change i
requested does not affect the required as-left setpoint tolerance after testing (1 1 percent). The increase in the acceptable safety valve setpoint tolerances affects TS 3.4.a.l.a.
In addition, changes to the safety limit curves were proposed to accommodate this and other changes. Since changes.
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were requested in Section 2 of the TS, the licensee also proposed changes to the section format to conform to the standard TS and to correct typographical errors.
The November 27, 1995, and March 1,1996, letters provided clarifying information in response to NRC staff questions. This information was within the scope of the original application and did not change the staff's initial proposed no significant hazards consideration determination.
2.0 BACKGROUND
The pressurizer safety valves provide, in conjunction with the Reactor 2
Protection System, overpressure protection for the Reactor Coolant System (RCS). The safety valves are designed to prevent the system pressure from exceeding the system safety limit. At PI there are two pressurizer safety valves set at 2485 psig. There are a total of 10 MSSVs on the 2 main steam headers which provide overpressure protection for the main steam system.
PI safety valves have been found outside the i 1 percent tolerance following operation and the licensee performed an evaluation to support changing the acceptable as-found setpoint to i 3 percent. Other licensees have encountered similar difficulties due to setpoint drift.
The safety limit curves of Figure TS.2.1-1 define the regions of acceptable operation with respect to average temperatures, power, and pressurizer pressure. These boundaries of acceptable operations are limited by the thermal overpower limit (fuel melting), thermal overtemperature limit (cladding damage based on departure from nucleate boiling (DNB) 9605300150 960521 PDR ADOCK 05000282 P
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considerations), and the locus of points where the steam generator safety valves open. These limits are used to set certain reactor trip setpoints.
3.0 EVALUATION Safety Limits Curve and TS Section 2 The proposed changes in this area would adjust for the impact of changing the lift setting tolerances for the pressurizer safety valves and MSSVs.
The proposed changes would also remove the curve showing the locus of points at which the MSSVs open and lower the DNB limit part of the curves.
1 As previously stated, TS Figure 2.1-1 defines the regions of acceptable i
operation with respect to average RCS temperature, reactor thermal power, and i
RCS pressure for which the minimum DNB ratio is not less than the safety analysis limit (1.30 for Exxon fuel and 1.17 for Westinghouse fuel).
The i
figure also shows the locus of points at which steam generator safety valves
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open. The licensee has requested removal of the curve showing the locus of points at which the safety valves open.
Since this locus curve is not a safety limit but, rather, represents a limiting condition governed by plant equipment which prevents plant operating conditions from approaching the safety limits, the staff concurs with its removal from Figure 2.1-1.
Figure 4
B.2.1-1 has instead been incorporated into TS Bases 2.1 along with the AT trips and the locus of points where the steam generator safety valves open to demonstrate that the AT trips and the steam generator safety valves do protect the reactor from exceeding the safety limits.
The locus of points past the first knee in Figure 2.1-1 for all pressures represents the thermal-hydraulic conditions above which the hot channel has a DNBR less than the limit. The licensee has proposed to lower this portion of the curves because the bypass flow fraction was increased to 6 percent when the fuel thimble plugs were removed, the Fh limit was increased to 1.75 with the loss-of-coolant accident (LOCA) analysis performed for Unit 2 Cycle 16, and a rod bow penalty of 2.6 percent was added to the DNB calculation.
The staff finds this proposed change acceptable since it was evaluated using approved DNB methodology and provides appropriate safety margins in the curves of Figure 2.1-1.
The licensee has also proposed changi,ng the x-axis of Figure 2.1-1 from "%
Rated Core Power" to " delta-T(T 4,)
F".
The staff finds this proposed s
change acceptable for the following reasons.
The full power AT is different at different temperatures and pressures because water properties are j
nonlinear. This makes it difficult to plot the curves at each pressure using the same scale for the percent power axis.
In addition, the AT trip setpoints which the reactor protection system actually calculates is based on the AT, not the percent power. Therefore, the proposed x-axis is a more appropriate variable.
l An additional proposed change would remove the curve at 1685 psig and add curves at 1785 and 1885 psig to Figure 2.1-1.
Since 1685 psig is a much lower pressure than would ever be achieved in plant operation due to the TS limits of the low pressurizer pressure trip setpoint, its elimination is acceptable.
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P The addition of the curves at 1785 and 1885 psig ensures that the entire range of pressures allowed by TS 2.3.A.2.c is bounded. This change is, therefore, j
acceptable, j
The remaining proposed changes to TS Section 2 combine the specifications of Sections 2.1 and 2.2 into one section, correct typographical errors, and s-delete the Safety Limit Violation specification of 6.4 in the Administrative Controls Section of the TS and incorporate it into TS 2.2 in confermance with the Standard Technical Specifications. The Table of Contents has also been revised to reflect the above changes.
These proposed changes are administrative and only serve to correct typographical errors and relocate certain specifications to make the TS more clear. The staff has reviewed these administrative changes and finds them acceptable.
The proposed changes to Bases Section 2 support the changes made to TS Section 2 and are, therefore, acceptable.
3.2 Safety Valve Setooint Tolerance The pressurizer safety valves (PSVs), in conjunction with other safety systems, provide overpressure protection for the RCS. The valves are designed to prevent the RCS from exceeding 110 percent of the design pressure, which is the safety 1.imit, for the most limiting overpressure transient.
For Prairie Island,110 percent of the design pressure is 2735 psig.
Prairie Island has two PSVs which are set at 2485 psig.
As a result of increasing the acceptable tolerance of the pressurizer safety valves to +3 percent, the RCS may reach a higher peak pressure than the i
previously calculated value. The licensee has performed calculations to show j
that the increased pressure is acceptable.
The most severe overpressure transient for the RCS at Prairie Island is a loss of external load or a turbine trip. -The licensee recalculated the maximum RCS pressure for the turbine trip using the following conservative assumptions:
1.
No credit is taken for the anticipatory turbine trip. The reactor trip was modeled' to occur later in the transient as a result of high RCS pressure.
2.
No credit is taken for turbine bypass to the condenser.
3.
No credit is taken for the pressurizer sprays to reduce pressure.
4.
No credit is taken for the pressurizer power-operated relief valves (PORVs).
5.
No credit is taken for 'Se main steam PORVs.
6.
The transient occurs at 30 psi higher than normal operating RCS pressure and at 102 percent of rated power due to instrument
'mcertainty.
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! l The pressurizer and MSSVs all were modeled to open at 3 percent higher than their setpoint with a 3 percent accumulation.
The results show that the peak RCS pressure for this limiting transient is 2560 psig which is lower than the i
maximum allowable peak pressure of 2735 psig and is therefore acceptable.
With regard to the -3 percent tolerance, the plant has been analyzed assuming a pressurizer PORV opens. The PORV setpoint is below the -3 percent tolerance of the PSVs. Therefore, for conditions under which the PORV is modeled to open below the lower tolerance (-3 percent), no further analysis is required.
Additionally, the reactor trip setpoint is below the -3 percent tolerance of the PSVs to preclude the valve openir,g prior to the reactor trip on high RCS 1
pressure. No LOCAs were reanalyzed because RCS pressure continually drops and the PSVs will not be challenged following a LOCA.
The primary purpose of the MSSVs is to provide overpressure protection for the secondary system. The MSSVs also provide protection against RCS overpressurization by providing a means to remove energy from the RCS.
The MSSVs are required to limit the. secondary system pressure to 110 percent of J
the design pressure. The design pressure for the Prairie Island secondary system is 1085 psig and the safety limit, 110 percent of the design pressure, is 1195 psig. Additionally, the MSSV design capacity is to relieve 110 3
percent of rated steam flow at 110 percent steam generator design pressure.
The modification does not physically affect the valves or the relieving capacity.
Prairie Island has five MSSVs on each steam line with setpoints of 1077 psig, 1093 psig, 1110 psig, 1120 psig, and 1131 psig.
t As a result of increasing the acceptable tolerance of the MSSVs to +3 percent the secondary system may reach a higher pressure than the previously analyzed value.
The licensee has performed calculations to show that the increased pressure is acceptable. The most severe overpressure transient for the secondary system at Prairie Island is also loss of external load or a turbine trip. The licensee recalculated the maximum pressure for the turbine trip using the same conservative assumptions as in the case for the peak RCS analysis with the exception of the RCS pressurizer pressure controller which is modeled to continue to operate and slow RCS pressure increase and delay the reactor trip. This maximizes the secondary system pressure and is acceptable.
The pressurizer and MSSVs all were acdeled to open at 3 percent higher than their setpoint with a 3 percent accumulation. The results show that the peak main steam pressure for this limiting transient is 1153 psig which is lower than the 1195 psig safety limit and therefore acceptable.
The licensee reanalyzed the small break LOCA (SBLOCA) because the secondary system is relied on to remove some of the decay heat from the primary RCS i
following the SBLOCA. The results for the SBLOCA reanalysis show that the emergency core cooling system acceptance criteria continue to be met and the results are acceptable.
For the large break LOCAs no analysis was required because the secondary system is not relied on to remove any decay heat from the primary RCS. Additionally, the licensee analyzed the effect of the modification on the auxiliary feedwater flow and determined that safety valves will maintain steam generator pressure low enough to assure minimum required auxiliary feedwater flow. The staff finds these analyses acceptable.
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The reanalyses necessitated changes to TS Table 4.1-2A, Minimum Frequencies I
for Equipment Tests. The licensee proposed changes to items 3 and 4 of TS Table 4.1-2A to reflect the changes in pressurizer and MSSV setpoints.
In addition, TS Table 4.1-2A has been reformatted and the footnote for item 11 has been relocated to become part of the Table. These changes were reviewed by the staff and are acceptable.
4.0 STATE CONSULTATION
In accordance with the Comission's regulations, the Minnesota State official was notified of the proposed issuance of the amendments. The State official had no comments.
l 5.0 ENVIROl94 ENTAL CONSIDERATION The amendments change requirements with respect to installation or use of a I
facility component located within the restricted area as defined in 10 CFR Part 20. The NRC staff has determined that the amendments involve no l
significant increase in the amounts, and no significant change in the types, l
of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure.
The Commission has previously issued a proposed finding that the amendments involve no significant hazards consideration and there has been no 4
public comment on such finding (60 FR 47621). Accordingly, the amendments meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9).
Pursuant to 10 CFR 51.22(b), no environmental impact statement i
or environmental assessment need be prepared in connection with the issuance I
of the amendments.
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6.0 CONCLUSION
The Comission has concluded, based on the considerations discussed above, that:
(1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendments will not be inimical to the common defense and security or to the health and safety of the public.
Principal Contributors:
L. Kopp C. Jackson Date:
May 21, 1996 i
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