ML20138M057
| ML20138M057 | |
| Person / Time | |
|---|---|
| Site: | Point Beach  |
| Issue date: | 02/20/1997 |
| From: | NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20138M052 | List: |
| References | |
| NUDOCS 9702250246 | |
| Download: ML20138M057 (6) | |
Text
.
l p %q p
t UNITED STATES l
g j
NUCLEAR REGULATORY COMMISSION l
WASHINGTON, D.C. 3086&0001 o%...../
l SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION l
l RELATED TO AMENDMENT NOR 172 AND 176 TO FACILITY OPERATING LICENSE NOS. DPR-24 AND DPR-27 WISCONSIN ELECTRIC POWER COMPANY POINT BEACH NUCLEAR PLANT. UNIT NOS. I AND 2 DOCKET NOS. 50-266 AND 50-301
1.0 INTRODUCTION
By letter dated September 19, 1996, as supplemented November 18, 1996, revised on January 13 and supplemented on January 27, 1997, the Wisconsin Electric Power Company (the licensee) regaested amendments to the Technical Specificaticas (TS) appended to Facility Operating License Nos. DPR-24 and DPR-27 for the Point Beach Nuclear Plant, Unit Nos. I and 2.
The proposed amendments would redesignate the overpressure mitigation system to the low temperatcre overpressure protection (LTOP) sutam in TS 15.3.15, Table 15.4.1-1, TS 15.6.9, and Table of Contents entry for TS 15.3.15.
l l
Additiofially, the proposed amendments would revise TS 15.3.1, " Reactor Coolant l
System," and TS 15.3.15, " Low Temperature Overpressure Protection System," to specify that below a reactor coolant system (RCS) temperature of M5 F the pressurizer power-operated relief valves (PORVs) and the LTOP sy:. tem are operable, modi?y TS 15.3.15 requirement to limit operation of the high pressure safety injection pump at an RCS cold leg temperature of 5; 275 'F to whenever the LTOP system is required to be operable, and revise the PORV l
setpoint from 425 psig to 440 psig.
2.0
. EVALUATION I
2.1 LTOP Setooint The purpose of the LTOP system is to control the RCS pressure at low temperature so that the integrity of the reactor coolant pressure boundary is not con: promised by violating the pressure temperature (P-T) limits of 10 CFR Part 50, Appendix G, which is based on Appendix G.Section XI of the American Society of Mechanical Engineers Boiler and Pressure Vusel Code (ASME Code).
Currently, Point Beach Nuclear Pl.nt (PBNP) IS state that the PORV lift setting be less than or equa'i to 425 psig. The licensee is proposing to increase the PORV lift setting to less thar, or equal 440 psig. On January 27, 1997, '% Commission exempted PBNP from tht eequirements 10 CFR 50.60, thereby, permitting the use of ASME Code Cate N-514, " Low mperature Overpressare Protectior.," for use at PBNP to establish the lift setpoint of the PORV for overpressure protection during iow temperature conditions. As i
delineated in the Coda Case, the LTOP system shall limit the maximum pressure 9702250246 970220 PDR ADOCK 05000266 P
l l,
l-",
in the vessel to 110% (1.1) of the pressure determined to satisfy Appendix l
G,Section XI of the ASME Code.
The licensee has performed an evaluation that shows the PORV setpoint of less than or equal to 440 psig is sufficient to ensure that the peak pressure for the most limiting reactor vessel weld in both units is less than 1.1 times the ASME 3ection III, Appendix G, for both the limiting mass addition transient and the limiting energy addition transient assuming the RCS is water solid.
The setpoint calculation includes temperature and pressure instrument uncertainties, in addftion to a pressure correction to account for the i
elevation static pressure difference and the dynamic pressure difference, created from the flu of both reactor coolant pumps and both ruidual heat removal (RHR) pumps, Detween the pressure sensor and the limiting weld. The licensee is using the same methodology to verify that the current PORV setpoint is adequate that was used to verify that the previous setpoints were adequate. The previous setpoints were approved by the staff and documented in I
a safety evaluation issued on May 20, 1980.
i The most limiting pressure transient for Point Beach is the mass addition 2
transient which assumes that a high head safety injection pump is started i
while the RCS is water solid and the temperature is at the coldest allowable j
temperature.
The Appendix G iimits are most limiting at the coldest possible temperature and the use of this point will provide limiting results. The results indicate that the calculated peak pressure will not exceed the limits.
The energy addition transient is analyzed at four different temperatures and assumes that the first reactor coolant purg inadvertently starts with the steam generator temperature 50 *F higher thri the RCS temperature. The start of a reactor coolant pump causes the water-solid RCS to heat up and results in i
an increase in RCS p*ressure. This transient is evaluated based on initial RCS temperatures of 100 F,140 *F,180 *F, and 250 *F with bias correction for two reactor coolant pumps operating and two RHR pumps operating, which is conservative because the analysis assumes only one reactor coolant pump is started. The range of temperatures chosen provides enough information to show that the.results will'be representative. The results indicate that the PORV setpoint of 440 psi is acceptable and prevents the RCS pressures from exceeding the limits.
The staff review therefore concludes that the PORV setpoint is acceptable.
2.2 LTOP Enacle Temnerature Code Case N-514.1dicates that the LTOP system is required to be operable at a water temperature corresponding to a metal temperature of RT 50 F at the beltline location that is controlling in the Appendix G limiIy +lculations.
ca Based on the limiting RT F for Unit 1 (See section 2.4) the metal temperature would be 312N *of 262.4The LTOP enable temperature of 355 *F provid F.
l a 93 'F margin between the water temperature and the limiting RT, (355 'F minus 262 *F = 93 *F).
ThemarginincludesuncertaintiesresultTngfrom instrument error and the difference between the metal temperature and fluid temperature at the maximum heatup rate (100 'F/hr).
The staff finds that the licensee's proposed TS LTOP enable temperature of 355 *F is acceptable.
I r
l
! l 2.3 LTOP TS Chanaes The licensee is proposing changes to TS 15.3.15 to:
retitle the overpressure mitigating system to the LTOP system, I
restate the TS's objective as restricting RCS operation when the RCS is operated at low temperatures rather than without a pressure-absorbing volume in the pressurizer, replace RCS temperature limits for LTOP and PORV limiting conditions for operation from "less than the minimum pressurhation temperature for the inservice pressure test as specified in Fige 15.3. 1-1" to less than 355 *F, revise the restriction on having no more than one high pressure safety l
injection pump operable from "the temperature of one or both RCS cold legs it less than or equal to 275 F" to when "LTOP is required to be enabled," and revise the TS basis to reflect the above changes and correct a typographical error " insure" to " ensure."
These proposed changes modify the condition for overpressure protection during low temperature operation by ensuring temperature requirements for LTOP system l
operability are consistent with the LTOP event analysis and Appendix G.
Based on a review of the applicability, objective, specifications, and the basis for the affected TS, the staff fir,ds these changes acceptable.
The licensee is proposing to change to TS 15.3.1.A.5 to replace the requirement to maintain RCS temperature greater than "the minimum pressurization temperature for the inservice pressure test as defined in Figure 15.3.1-1" to 355 F.
This change will ensure consistency with the operability requirements for the TS 15.3.15 as described above and are acceptable to the staff.
The licensee is proposing to change the system designation from the overtemperature mitigating system to LTOP system in Table 15.4.1-1 and TS i
15.6.9.2.
These changes are administrative and are acceptable to the staff.
2.4 Reactor Vessel Material Data The staff reviewed the applicability of reactor vessel material data included in calculations supporting the proposed change in PORV setpoint for LTOP system operability. An adjusted reference temperature (ART) was calculated i
{
based on the methods in Regulatory Guide (RG) 1.99, " Radiation Embrittlement of Reactor Vessel Materials," Revision 2.
The ART is defined as the sum of initial nil-ductility transition reference temperature (RT 1) of the material, the increase in the RT, caused by neutron irrad,iation, and a margintoaccountforuncertaintYesinthepredictionmethod.
The increase in RT, is calculated from the product of a chemistry factor (CF) and a fluence i
factor. The chemistry factor is calculated depending upon the amount of
I
, copper and nickel in the vessel material as specified in Table 1 of RG 1.99, which was listed as 0 'F rather than 10 *F.The staff found a typographical error for the initial U
-Rev. 2.
The licensee's calculation use the correct value.
i The most limiting weld for Unit 1 is the intermediate-to-lower shell i
circumferential weld, SA-1101. The most limiting material for Unit 2 is the intermediate-to-lower shell cirec:;4aential weld, SA-1484. The licensee's calculation used the Ig4T locations for determining the ART.
For weld SA-Il01 an ART equal to 262.4 F was determined based on a copper content of 0.26 weight percent (wt.%), nickel content of 0.60 wt.%, CF of 180 'F, initigl RT,
j of 10 *F, a margin of 56 'F, and a neutron fluence at 1/4T of 1.39 x 10 s
neutrons per square centimeter (n/ca').
For SA-1484 an ART equal to 249.7 'F was determined based on a copper content of 0.24 wt.%, nickel content of 0.60 i
wt.%, CF of 173 'F, initial g,/cm,.of -5 'F, a margin of 66 'F, and a neutron fluence at'l/4T of 1.39 x 10 n
The licensee calculated the limiting *F for-the limiting material in the temperature for the closure flange region using a reference temperature of 60 closure flange region that is highly stressed by bolt preload. The licensee determined that the minimum allowable temperature of 77.8 'F (60 'F +
instrument uncertainty of 17.8 F) is required prior to pressurizing the j
reactor vessel.
The staff verified that the copper and nickel content and initial RT with the NRC reactor vessel material database as reported by the lic., agreed j
ensee in response to Generic Letter 92-01, " Reactor Vessel Structural Integrity." The staff also verified that the licensee evaluated the most limiting material locations for both units. The staff reviewed the material properties used in the calculation of the ART values for the limiting materials using RG 1.99, Revision 2, methodology. The staff reviewed the licensee's analysis of i
minimum temperature at the closure head flange performed in accordance with 10 CFR Part 50, Appendix G, Section IV.A.2.
The staff verified that the l
properties used for the most limiting materials and conditions were used to i
verify the new calculated LTOP setpoint. The staff concluded that appropriate i
material data was used in the licensee's calculations to verify meeting the j
applicable requirements of 10 CFR Part 50, Appendix G, based on the use of ASME Code Case N-514.
2.5 Instrument Error The proposed TS change required determination of new LTOP setpoints. The new setpoints include pressure instrument error, associated with existing instrumentation, a bias value associated with the elevation difference of the wide range pressure transmitters and the mid-plane of the reactor vessel, and a bias value for flow effects for the operation of both reactor coolant pumps and two RHR pumps. The total pressure instrument location bias values for each unit based on one reactor coolant pump operating and with two reactor coolant pumps running are:
-41.3 psig and -70.3 psig for Unit 1, and -44.6 psig and -74.4 psig for Unit 2.
Instrument uncertainties for pressure and
.)
l l
I i temperature of 13 psi and 17.8 'F are included in the licensee's
{
calculation, j
The licensee inclusion of instrument uncertainties was found acceptable by the staff.
3.0 EXIGENT CIRCUMSTANCES
The Commission's regulations, 10 CFR 50.91, contain provisions for issuance of amendments where the Commission finds that exigent circumstances exist, in that a licentee and the Commission must act quickly and that time does not j
permit the Commission to publish a Federal Reaiste.t notice allowing 30 days for prior public comment. The exigency exists in that the proposed amendments l
are needed prior to the start of reactor vessel head tensioning and time does l
not permit the Commission to publish a notice allowing 30 days for prior l.
public comment.
The licensee revised its original application on January 13, 1997, in anticipation of receiving an exemption from the requirements of 10 CFR 50.60, which the Commission granted on January 27, 1997. After submitting the revised application, the licensee determined that additional calculations were requireo, which the licensee promptly submitted on January 27, 1997. At the time of the licensee's January 27, 1997, submittal, reactor vessel head tensioning was scheduled for February 10, 1997, and startup of Point Beach Unit 2 was scheduled for February 25, 1997.
The-staff has determined that the licensee used its best efforts to make a timely application.
Accordingly, the Commission has determined that exigent circumstances exist pursuant to 10 CFR 50.91(a)(6), the submittal of information was timely and could not have been avoided, and that the licensee did not create the j
exigency.
l l
4.0 FINAL NO SIGNIFICANT HAZARDS CONSIDERATION
S DETERMINATION The Commission's regulations in 10 CFR 50.92(c) state that the Commission may make a final determination that a license amendment involves no significant hazerds consideration if operation of the facility in accordance with the proposed amendment would not (1) involve a significant increase in the probability or consequences of an accident previously evaluated, (2) create the possibility of a new or different kind of accident from any accident previously evaluated, or (3) result in a significant reduction in the margin of safety.
The NRC staff has made a final determination that no significant hazards consideration is involved for the proposed amendments and that the amendments should be issued as allowed by the criteria contained in 10 CFR 50.91. The NRC staff's final determination is presented below.
(1)
The proposed changes would not involve a significant increase in the i
probability or consequences of an accident previously evaluated.
i The proposed changes will explicitly define the temperature at which LTOP is required to be enabled, raise the temperature at which one high i
pressure safety injection pump is required to be rendered inoperable, i
and increase the setpoint of the PORVs. The changes do not affect any i
accident analyses since the LTOP is required only when RCS temperatures 4
1
. - ~
.. - ~,
E.
1 l
- are low.
LTOP is not required during power operation. The consequences I
or probability of a previously evaluated accident will, therefore,-not J
j-significantly be increased.
(2)
The proposed changes would not create the possibility of a new or 4
different kind of accident from any accident previously evaluated.
The proposed changes will still meet the requirements for fracture i
toughness requirements required by 10 CFR 50.60 as modified by the use of ASME Code Case N-514 which was developed by the ASME as an alternative to describe requirements in 10 CFR Part 50, Appendix G..
)
Therefore, a new or different kind of accident is not created.
l 1
l (3)
The proposed changes would not result in a significant reduction in the i
margin of safety.
j i
i The proposed changes increase the range of the temperature region where i
the LTOP system is needed, while increasing the allowed setpoint i
pressure by only 3.5 percent. Therefore, these changes do not involve a j
- significant reduction in a margin of safety.
j f
5.0 STATE CONSULTATION
In accordance with the Commission's regulations, the Wisconsin State official was notified of the proposed issuance of the amendments. The State official had no comments.-
6.0 ENVIRONMENTAL.CONSIDERATIDH j
These amendments change a requirement with respect to the installation or use i
of a facility component located within the restricted area as defined in 10
)
CFR Part 20. The staff has determined that the amendments involve no significant increase in the amounts, and no si5nificant change in the types, of any effluent that may be released offsite, and that there is no significant
-increase in individual or cumulative occupational radiation exposure. The Commission made a final finding that the amendments involve no significant hazards consideration. Accordingly, these 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 or environmental assessment need be prepared in connection with the issuance of these amendments.
7.0 CONCLUSION
The staff has concluded, based on the considerations discussed above, that (1) there is reasonable assuran::e 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 I
defense and security or to the health and safety of the public.
Principal Contributors:
C. Jackson S. Sheng Date:
February'20, 1997
~
__