ML20082K376
| ML20082K376 | |
| Person / Time | |
|---|---|
| Site: | Diablo Canyon  |
| Issue date: | 04/13/1995 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20082K374 | List: |
| References | |
| NUDOCS 9504190257 | |
| Download: ML20082K376 (6) | |
Text
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b ar UNITED STATES p
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NUCLEAR REGULATORY COMMISSION t
WASHINGTON D.C. 20666-0001
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO.100 - TO FACILITY OPERATING LICENSE NO. DPR-80 AND AMENDMENT NO. 99 TO FACILITY OPERATING LICENSE NO. DPR-82 PACIFIC GAS AND ELECTRIC COMPANY DIABLO CANYON NUCLEAR POWER PLANT. UNIT NOS. 1 AND 2 DOCKET NOS. 50-275 AND 50-323
1.0 INTRODUCTION
By letter of August 17, 1994, Pacific Gas and Electric Company (or the licensee) submitted a request for changes to the Technical Specifications (TS). The proposed amendments would revise the combined TS for the Diablo Canyon Nuclear Power Plant (DCPP), Unit Nos. I and 2 to extend the current 8 effective full-power years (EFPY) up to 12 EFPY for the reactor vessel, to specify a new low-temperature overpressure protection (LTOP) system actuation pressure setpoint and a new LTOP system enable temperature. The associated Bases would also be appropriately revised. The amendments revise the TS to change TS 3/4.4.9.1, " Reactor Coolant System - Pressure / Temperature Limits,"
Figures 3.4-2, " Reactor Coolant System Heatup Limitations - Applicable Up to 8 EFPY," and 3.4-3, " Reactor Coolant System Cooldown Limitations - Applicable Up to 8 EFPY," to extend the applicability up to 12 effective full-power \\ years (EFPYs). TS 3/4/4/9/3, " Overpressure Protection Systems," is revised to specify a new low-temperature overprotection (LTOP) system actuation pressure setpoint. Additionally, TS 3/4.1.2.2, " Flow Paths - Operating;" TS 3/4.1.2.4,
" Charging Pumps - Operating;" TS 3/4.4.1.3, " Hot Shutdown;" TS 3/4.4.1.4.1,
" Cold Shutdown - Loops Filled;" TS 3/4.4.9.3, " Overpressure Protection Systems;" and TS 3/4.5.3, "ECCS Subsystems - T Less than 350 Degrees F,"
are revised to specify a new LTOP system enablEtemperature.
In order to assess the extension of EFPY, the staff evaluates the pressure-temperature (P-T) limits for the facility based on the following NRC regulations and guidance: Appendix G to 10 CFR Part 50; Generic Letters (GLs) 88-11, "NRC Position on Radiation Embrittlement of Reactor Vessel Materials and its Impact on Plant Operations," and GL 92-01, " Reactor Vessel Structural Integrity"; Regulatory Guide (RG) 1.99, " Radiation Embrittlement of Reactor Vessel Materials," Revision 2; and Standard Review Plan (SRP) Section 5.3.2,
" Pressure-Temperature Limits." Appendix G to 10 CFR Part 50 requires that P-T limits for the reactor vessel must be at least as conservative as those obtained by Appendix G to Section III of the American Society of Mechanical Engineers (ASME) Code. GL 88-11 requires that licensees use the methods in RG 1.99, Revision 2, to predict the effect of neutron irradiation by calculating the adjusted reference temperature (ART) of reactor vessel materials. The ART is defined as the sum of initial nil-ductility transition reference temperature (RTut) of the material, the increase in RT, caused by g
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neutron irradiation, and a margin to account for uncertainties in the prediction method. The increase in RT is calculated from the product of a chemistry factor and a fluence factor.gYhe chemistry factor is dependent upon l
the amount of copper and nickel in the vessel beltline materials. GL 92-01 requires licensees to submit reactor vessel materials data, which the staff will use in the review of the P-T limits submittals. SRP Section 5.3.2 provides guidance on calculation of the P-T limits using linear elastic fracture mechanics methodology specified in Appendix G to Section III of the ASME Code.
I To ensure the adequacy of the P-T limits, the staff also evaluates the adequacy of the fluence value used by the licensee. Because neutron irradiation of the reactor vessel affects the LTOP system settings, the staff reviews the LTOP setpoints for conformance to SRP 5.2.2, " Overpressure Protection."
2.0 EVALUATION 1
The licensee has reevaluated the fluence for 12 EFPYs and found it to be lower than that assumed in the original estimate for 8 EFPYs. The original fluence value for 8 EFPYs was estimated using the out-in-in loading scheme (new fuel is added to the outermost core region and moved inward for two successive cycles before removal). This design results in high neutron leakage from the vessel. However, since cycle 1, the licensee has employed low leakage core loading patterns which have reduced the reactor vessel peak flux by about 40 t
percent. Thus, the licensee reestimated the neutron sources to reflect a realistic fluence value. Also, two surveillance capsules from each unit have been removed and evaluated, and the actual surveillance results confirmed the calculated projections, i
By letter dated September 15, 1993, the licensee submitted WCAP-13750,
" Analysis of Capsule Y from the Pacific Gas and Electric Company Diablo Canyon Unit 1 Reactor Vessel Radiation Surveillance Program," which provided the evaluation of surveillance capsule Y.
This capsule was removed from the Unit I reactor vessel on September 30, 1992. WCAP-13750 includes a detailed account of the revised estimation for 12 EFPY and updates the estimate of fluence and the associated P-T limits based on capsule S which was removed from Unit I at the end of cycle 1 (August 1986). The calculations were carried out using the DOT two-dimensional discrete ordinates transport code described in WANL-PR(LL)-034, " Nuclear Rocket Shielding Methods, Modification, Updating and Input Data Preparation," Volume 5, "Two Dimensional Discrete Ordinates Transport Technique," dated August 1970. As discussed in WCAP-13750, the approximations used included a P scattering cross section and 3
angular quadrature. The neutron sources accounted for the effects of the a S, ion yields due to plutonium buildup from increased burnup in the 4
fiss peripheral assemblies. However, the cross sections used were based on the SAILOR cross section library which utilized the ENDF/B-IV data. These data underestimate the value of the fluence particularly through iron; the Diablo Canyon plants have iron thermal shields. The WCAP-13750 evaluation identified this underestimation and compensated with a bias factor of 1.16.
To estimate the bias, WCAP-13750 utilized data from cavity dosimetry from cycles 2, 3, and
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The staff concludes that an increase of 16 percent appropriately compensates for the underestimation. Based on its review of the licensee's reevaluation, the staff finds the projected fluence values up to 12 j
EFPY to be acceptable.
For the Units 1 and 2 reactor vessels, the licensee determined that lower shell longitudinal weld 3-442C is the limiting material at the 1/4T location i
and intermediate shell plate B5454-2 at the 3/4T location (T - the thickness i
of the reactor vessel beltline). The licensee calculated an ART of 161*F at j
the 1/4T location and 131*F at the 3/4T location. The calculation variables are shown in the attached table.
j The staff verified the copper and nickel contents and initial RT, with respect to the NRC reactor vessel material database from the licensee's
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response to GL 92-01. The staff used the material properties to perform an j
independent calculation of the ART values for the limiting materials using RG 1.99, Revision 2.
Based on the its calculation, the staff verified that j
the licensee's calculated ARTS are acceptable.
j Substituting the ARTS from the attached table into equations in SRP Section 5.3.2, the staff verified that the proposed P-T limits for heatup, cooldown, i
criticality, and inservice hydrostatic test satisfy the requirements in Paragraphs IV.A.2 and IV.A.3 of Appendix G of 10 CFR Part 50.
In addition to beltline materials, Appendix G of 10 CFR Part 50 also imposes a i
minimum temperature at the closure head flange based on the reference j
temperature for the flange material.
Section IV.A.2 of Appendix G states that when the pressure exceeds 20 percent of the preservice system hydrostatic test j
pressure, the temperature of the closure flange regions highly stressed by the i
bolt preload must exceed the reference temperature of the material in those j
regions by at least 120*F for normal operation and by 90*F for hydrostatic i
pressure tests and leak tests.
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the licensee, the staff has determined that the propos,e#. of 53*F provided by d P-T limits have i
satisfied the requirement for the closure flange region during normal operation, hydrostatic pressure test and leak test.
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Because the LTOP system setpoints are affected by neutron irradiation of the i
reactor vessel, the licensee has assessed these setpoints. The purpose of j
these setpoints is to prevent overpressurization at low temperatures. At temperatures below the LTOP enable temperature, the setpoint of the power-operated relief valves (PORVs) is reset to a lower pressure in order to protect the primary system from an LTOP event. The existing PORV setpoint of 450 psig would normally not need to be changed because the fluence was found acceptable to 12 EFPY. However, conditions were identified in Information Notice 93-58, "Nonconservatism in Low-Temperature Overpressure Protection for Pressurized-Water Reactors," where the LTOP setpoint did not consider (1) the static head correction due to the elevation difference between the LTOP hot leg pressure tap location and the reactor vessel beltline and (2) the pressure drop across the vessel due to operation of the reactor coolant and residual l
heat removal pumps and which may impact the PORV setpoint.
In addition, the j
licensee included an additional 0.25 second response time in the wide-range I
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pressure sensor which resulted from installation of the Eagle-21 reactor
,1 protection system. Finally, the licensee assumed a combined emergency core
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cooling system (ECCS) injection from one centrifugal charging pump ani the a
positive displacement pump which was not considered in the original LTOP 1
design basis transient and which will exceed the original design basis j
transient flow. These additional conservatisms result in the LTOP pressure setpoint being changed from less than or equal to 450 )sig to less than or i
j equal to 435 psig. We find the proposed value, with tie additional i
considerations discussed above, to be acceptable, j
Similarly, the present LTOP enable temperature of 323*F included in the TS would not need to be changed to satisfy Appendix G limits since the heatup and j
cooldown curves are unchanged from 8 to 12 EFPY. However, this LTOP enable j
temperature was determined before issuance of the position in Branch Technical l
Position RSB 5-2, " Overpressure Protection of Pressurized Water Reactors While Operating at Low Temperatures," which defines the enable temperature as the i
water temperature which corresponds to a metal temperature at least equal to
+ 90*F at the beltline region. Utilizing this position, the licensee RT, determined the enable temperature to be 270*F for both units through 12 j
has EFPY of operation. The staff has found that this value satisfies the j
requirements of Branch Technical Position RSB 5-2 and is therefore acceptable.
The licensee has also proposed by this request deleting reference in the Bases of the TS to Table 4.4-5, in which the surveillance capsule withdrawal schedule had formerly been located. This reference was inadvertently not deleted in a previous amendment. This is an editorial change, and the staff j
concurs. The licensee has also proposed to revise the Bases of the TS to be i
consistent with the proposed TS amendments, and we find these Bases revisions j
acceptable.
3.0 STATE CONSULTATION
i In accordance with the Connission's regulations, the California State official l
was notified of the proposed issuance of the amendments. The State official j
had no comments.
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4.0 ENVIRONMENTAL CONSIDERATION
These amendments change a requirement with respect to the installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20 and change surveillance requirements. The NRC staff has determined that the amendments involve no significant increase in the amounts, and no significant change in the types, 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 considera-tion, and there has been no public comment on such finding (59 FR 51622).
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 or environmental assessment need be prepared in connection with the issuance of the amendments.
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5.0 CONCLUSION
The Commission 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 ao.endment will not be inimical to the common defense and security or to the health and safety of the public.
Attachment:
Table Principal Contributors:
C. Fairbanks L. Lois M. Miller Date:
April 13,1995
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Table ADJUSTED REFERENCE TEMPERATURE FOR REACTOR VESSEL LIMITING MATERIALS FOR PRESSURE-TEMPERATURE LIMITS 12 EFFECTIVE FULL POWEP. YEARS DIABLO CANYON UNITS 1 AND 2 Limiting Cu Ni Fluence
RT*
1 Weld 3-442C Plate B5454-2 Units 1 and 2 3-442C (1/4T)
.198 1.00 3.63E18 151.9
-56 65.5 161 B5454-2 (3/4T)
.14
.59 1.30E18 46.8 67 17 131 l
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