ML20011E055
| ML20011E055 | |
| Person / Time | |
|---|---|
| Site: | Salem  |
| Issue date: | 01/29/1990 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20011E054 | List: |
| References | |
| NUDOCS 9002070192 | |
| Download: ML20011E055 (5) | |
Text
.
.,'ih onog(
- 8 UNITED STATES NUCLEAR REGULATORY COMMISSION o
4'
(
WASHING TON, D, C. 20t 66 o
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR SUPPORTING AMENDMENT NOS.108 AND 86 TO FACILITY OPERA LICENSE NOS. DPR-70 AND DPR-75 PUBLIC SERVICE ELECTRIC & GAS COMPANY PHILADELPHIA ELECTRIC COMPANY
'DELMARVA POWER AND LIGHT COMPANY ATLANTIC CITY ELECTRIC COMPANY SALEM GENERATING STATION. UNIT NOS. 1 AND 2 DOCKET HOS. 50-272 AND 50-311 1.0.lHTRODUCTION
~
In response to Generic Letter 88-11. "NRC Position on Radiation Embrittlement of Reactor Vessel Materials and Its Effect on Plant Operations," the Public Service Electric and' Gas Company requested permission to revise the
-pressure / temperature (P/T) limits-intheSalemNuclear.GeneratingPlant Technical Specifications. The request was documented in a letter from the Llicensee dated December 28, 1988 and. supplemented in: letters dated July 31',. 1 1989, October 18, 1989 and December 19, 1989. The supplemental letters datedt July 31, 1989 and October 18,1989, provided additional technical information and revised technical-specification pages based on the results of the technical
-review.. The supplemental ~1etter dated December 19, 1989 provided corrected technical specification pages only and did not change the-scope of the amendment-request or affect the staff's original no significant hazards determination.
)
The purpose of the revision is to change the effectiveness of-the P/T limits to f
15 effective. full power years (EFPY) for Salem Unit 1 and 10 EFPY for Unit 2.
.The proposed P/T limits were developed based on Regulatory Guide (RG) 1.99, Revisinn 2.- The proposed revision provides up-to-date P/T limits-for the-operation of the reactor coolant system during heatup, cooldown, criticality,
,and hydrotest.
To evaluate the P/T limits, the staff uses the following NRC regulations and guidance: : Appendices G and H of 10 CFR Part 50; the ASTM Standards and the ASME Code, which are referenced in Appendices G and H; 10 CFR 50.36(c)(2); RG 1.99, Rev. 2; Standard Review Plant (SRP) Section 5.3.2; and Generic Letter 88-11.
Each licensee authorized to operate a nuclear power reactor is required by I
10 CFR 50.36 to provide Technical Specifications for the operation of the plant.
In particular, 10 CFR 50.36(c)(2) requires that limiting conditions of operation be included in the Technical Specifications. The P/T limits are among the limit-l ing conditions of operation in the Technical Specifications for all commercial i
. nuclear plants in the U.S.
Appendices G and H of 10 CFR Part 50 describe specific 9002070192 900129 0
PDR A D O C K 0 5 0 0 0 2 7:2 P
PDC h
j
r I
2 T
requirements for fracture toughness and reactor vessel material surveillance that must be considered in setting P/T limits. An acceptable method for constructing the P/T limits is described in SRP Section 5.3.2.
Appendix G of 10 CFR Part 50 specifies fracture toughness and testing requirements for reactor vessel materials in accordance with the ASME Code and, in particular, that the beltline materials in the surveillance capsules be tested in accordance with Appendix H of 10 CFR Part 50 ~ Appendix H, in turn, refers to ASTM Standards.
These tests define the extent of vessel embrittlement at the time of capsule with-drawal in terms of the increase in reference temperature. Appendix G also requires the licensee to predict the effects of neutron irradiation on vessel embrittlement by calculating the adjusted reference temperature (ART) and Charpy upper shelf energy (USE). Generic Letter 88-11 requested that licensees and permittees use the methods in Regulatory Guide 1.99, Revision 2, to predict the effect-of neutron irradiation on reactor vessel materials. This guide defines the ART as the' sum of unirrediated reference temperature, the increase in reference temperature resulting from neutron irradiation, and a margin to account for uncertainties in the prediction method.
Appendix H of 10 CFR Part 50 requires the licensee to establish a surveillance program to periodically withdraw surveillance capsules from the reactor vessel.
Appendix H refers to the ASTM Standards which, in turn, require that the capsules be installed in the vessel before startup and that they contain test specimens made from plate, weld, and heat-affected-zone (HAZ) materials of the reactor beltline.
2.0 EVALUATION i-A.
Salem Unit 1 P
The staff evaluated the effect of neutron irradiation embrittlement on each belt-line material in the Salem Unit I reactor vessel in accordance with RG 1.99, Rev.
L 2.
The: staff has determined that the limiting material at 1/4T (T = reactor beltline thickness) for 15 EFPY is the lower shell longitudinal seam weld, 3-042C. -This weld uses Linde flux 1092 with heat number 34B009 and has 0.19%
copper, 1.00% nickel, and an unirradiated RT of
- However, a conservative 0.35% copper was used in the M ensee'56'F (Ref. 1).
s calculation of the ART.
The staff also used the 0.35% copper in the calculation. At the 3/4T location, I
the limiting material is the intermediate shell plate, B2402-1. The plate has 0.22% copper, 0.53% nickel, and an unirradiated RT of 45'F.
NDT The licensee removed three surveillance capsules T, Y, and Z from Unit 1 and the L
results of capsule analysis were reported in References 2, 3, and 4 The limiting material, the longitudinal weld, was not included in any of the three capsules but the plate, B2402-1, was included in the T and Z capsules. Therefore, the ART of the weld was calculated based on Section 1.1 of RG 1.99 and the ART of the plate l
was calculated based on Section 2.1.
The staff as well as the licensee calculated l
D f
I j
en ART of 222.5'F for the longitudinal weld at 1/4T for the proposed 15 EFPY P/T limits. For plate B-2402-1, the licensee calculated an ART of 162'F and the staff, 145'F, at the 3/4T location. The difference between the two ARTS is that the licensee used a conservative margin.
Substituting the ART of 222.5'F and 162'F into equations in SRP 5.3.2, the staff verified that the proposed Unit 1 P/T limits for heatup, cooldown, and hydrotest meet the beltline material requirements in Appendix G of 10 CFR Part 50.
{
Because the ART of 222.5*F exceeds the Appendix G allowable of 200'F, Section' IV.B of Appendix G states that the reactor vessel must be designed to permit a i
thermal annealing treatment at a sufficiently high temperature to recover material toughness properties of ferritic materials of the reactor vessel beltline. The Salem Units 1 and 2 reactor vessels are designed to permit a thermal annealing treatment as stated in Section 5.4.3.1 of the Salem Updated Final Safety Analysis Report; therefore, the_ licensee satisfiesSection IV.B of i
Appendix B.
h In' addition to beltline materials, Appendix G of 10 CFR Part 50 also imposes
_P/T limits based on the reference temperature for the reactor vessel closure i
flange raterials.Section IV.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 stressed by the bolt preload must excged the. reference temperature of tge material _in those regions.by at least 120 F fornormaloperationandby90Fforhydrostatigpressuretestsandleaktests.
Based on the flange reference temperature of 28 F, the staff has determined that the proposed P/T limits satisfy Section IV.2 of Appendix G.
Section:1V.B of Appendix G requires that the predicted Charpy USE at end of life
.be above 50-ft-lb. The Plate B2402-1 has the lowest unirradiated USE of 73 ft-lb.
The percentage.of USE reduction due to irradiation was calculated based on sur-veillance capsule results as allowed by RG 1.99. The staff used the irradiated USE~ data of the B2404-1 plate to calculated an end of life USE of 57 ft-lb. This l
value is greater than 50 ft-lb and, therefore, satisfies the Appendix G requirement.
B.
Salem Unit 2
'For Salem Unit 2, the staff has confirmed the licensee's calculation that the limiting material at 1/4T and 3/4T for 10 EFPY is the same longitudinal weld as in Salem Unit 1.
The ART for the weld at 1/4T is 178.6*F and at 3/4T,116.1*F, with the neutron fluence at vessel inside diameter of 4.16E18 n/cm2 The
. licensee has removed capsules T & V from the Unit 2 reactor (Reference 5&6);
however, the surveillance data of the plate and weld specimens were not used because they were not limiting.
Substituting both ARTS into equations in SRP 5.3.2, the staff verified that the proposed Unit 2 P/T limits for heat up, cooldown, and hydrotest meet the beltline material requirements in Appendix G of-10 CFR Part 50.
w
- ~,-
w g
a i
]
t i 1J '
TheUnit2closureflangehasthesameRT@ propose'dP/TlimitssatisfySection of 28 F as that of Unit 1.
Based on the RT the staff determined that t IV.2ofAhn,dixG. The staff also determined that the USEs in beltline 1
materials satisfy the 50 ft-lb requirement. This was based on the lowest initial USE of 69 ft-lb of the plate B4711-3 and the end of life fluence of
~ 1.7E19 n/cm,
j a
C.
CONCLUSION' The staff concludes that the proposed P/T limits for the reactor coolant system.
for heatup, cooldown, leak test, and criticality are valid through 15 EFPY for
- 1 Unit 1 and 10 EFPY for Unit 2 because the limits conform to the requirements of Appendices G and H of 10 CFR Part 50. The licensee's submittal also satisfies Generic-Letter 88-11, because the licensee used the method in RG 1.99, Rev. 2, to calculate the ART. Hence, the proposed P/T limits may be incorporated into the Salem Units 1 and 2 Technical Specifications, j
The limiting material identified in this evaluation is different from the one in the current Salem Updated Final Safety Analysis Report; therefore, the licensee should revise the limiting material and chemistry contents of the material in Chapter 5 of the UFSAR.
Administrative changes (e.g. page. numbers, paragraph numbers, typographical I
errors) were made, with the concurrence of the licensee. to the revised technical-specification pages. These changes brought the revised technical specification pages into agreement with the marked up technical specification pages that were included with the amendment request.
D.
REFERENCES 1.
December 28, 1988, LetterfromS.E.Miltenberger(PSE&G)toNRCDocument Control Desk,
Subject:
Request for Amendment Salem Units 1 and 2.
2.
November 7, 1988, Letter from S. LaBruna (PSE8G) to NRC Document Control Desk,
Subject:
Surveillance Specimen Capsule Test Report, Salem Units 1.
3.
Yanichko S. E., et al, " Analysis of Capsule T from the Public Service Electric and Gas Company Salem Unit 1 Reactor Vessel Radiation-Surveillance Program," Westinghouse, WCAP-9678, February 1980.
4.
Boggs, R. S., et al, " Analysis of Capsule Y from the Public Service Electric ano Gas Company Salem Unit 1 Reactor Vessel Radiation Surveillance Program,"
Westinghouse, WCAP-10694, December 1984.
5.
Boggs, R. S., et al, " Analysis of Capsule T from the Public Service Electric and Gas Company Salem Unit 2 Reactor Vessel Radiation Surveillance Program,"
Westinghouse WCAP-10492, March 1984.
g, y
o c
4.
f 1 6.
.Yanichko, S. E., et al, " Analysis of Capsule U from the Public Service Electric and Gas Company. Salem Unit 2 Reactor Vessel Radiation Surveillance Program," Westinghouse, WCAP-11554 September 1987.
1
3.0 ENVIRONMENTAL CONSIDERATION
j These amendments involve a change to a requirement with respect to the i
installation or use of a facility component located within the restricted y
area as defined in 10 CFR Part 20 and changes to the surveillance 1
requirements..The 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' consideration and there has been no public comment on such finding. Accordingly, the amendments 9
meet the eligibility criteria for categorical exclusion set forth in 10 CFR51.22(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.
4.0 CONCLUSION
i The Commission made a proposed determination that the amendments involve no significant hazards consideration which was published in the Federal Register (54 FR 51258) on December 13, 1989. No public. comments were a.
received. The State of New Jersey notified the NRC that they did not have any comments by letter dated May 1,1989, The staff has concluded, based on the' considerations discussed above, that*
(1) there is reasonable assurance that the health and safety of the ublic will not be endangered.by operation in the proposed manner,-
and 2)suchactivitieswillbeconductedincompliancewiththe Commission's regulations and the issuance of the amendments will not be inimical to the common defense and security nor to the health and safety of the public.
Principal Contributor:
J. Tsao Dated: January 29, 1990
- -