ML20214C923

Safety Evaluation Supporting Amends 80 & 73 to Licenses DPR-42 & DPR-60,respectively

ML20214C923
Person / Time
Site:	Prairie Island
Issue date:	11/14/1986
From:	Office of Nuclear Reactor Regulation
To:
Shared Package
ML20214C887	List: ML20214C884 ML20214C917 ML20214C923
References
TAC-62137, TAC-62138, NUDOCS 8611210285
Download: ML20214C923 (13)

Text

. .

dyro

~g UNITED STATES 8 t NUCLEAR REGULATORY COMMISSION

& :E, WASHINGTON, D. C. 20555

\;4.....l SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMEN 0 MENT NOS. 80 AND 73 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 INTRODUCTION By letter dated July 15, 1986, Northern States Power Company (NSP), the licensee, requested amendments to Facility Operating License Nos. DPR-42 and DPR-60 for the Prairie Island Nuclear Generating Plant, Unit Nos. 1 and 2 (PINGP). The amendments would change the technical specifications by revising (1) the existing reactor coolant system heatup and cooldown curves that would increase the effective limitation from 10 to 15 effective full power years, (2) the Radiation Environmental Monitoring Program, (3) design features appearing in Section 5 of the technical specifications, (4) Security Plan Implementing Procedures, and (5) administrative changes appearing in Section 6 of the technical specifications (TS). Specifically, the amendment changes would affect Sections TS 3.1, TS 4.10-1, TS 5.1, TS 5.2, TS 5.3, TS 5.4, TS 5.5, TS 5.6, TS 6.2, TS 6.5, TS 6.6, TS 6.7 TS 6.8, Tables TS 3.1-1, TS 3.1.2, TS 4.10-1 and TS 4.10-2, and Figures TS 3.1-1 thru 5. These changes consider the increased neutron fluence up to 15 effective full power years for the limiting material of the reactor pressure vessel in regard to the heatup and cooldown curves and other changes involving error corrections, clarifications, minor administrative changes and changes to attain consistency throughout different sections of the TS.

DISCUSSION AND EVALUATION A. Heatup and Cooldown Curves Discussion The licensee requested changes to the pressure temperature limits described in Figures TS 3.1-1 and TS 3.1-2 and deletion of Figures TS 3.1-3 and TS 3.1-4 and Tables TS 3.1-1 and TS 3.1-2 from the technical specifications.

The changes in the heatup and cooldown limits are based on the test results from the Prairie Island surveillance program, contained in References 1 through 5. References 1 through 5 were submitted for staff review in letters from the licensee dated September 17, 1977, October 12, 1982, April 25, 1986, December 6, 1978 and May 13, 1981, respectively.

8611210285 861114 2 ADOCK 0500 gDR

l 1

I i

i Pressure-temperature limits must be calculated in accordance with the requirements of Appendix G, 10 CFR Part 50, which became effective on July 26, 1983. Pressure-temperature limits that are calculated in accordance with the requirements of Appendix G,10 CFR Part 50 are dependent upon the initial reference temperature (RTNDT) f r the  ;

limiting materials in the beltline and closure flange regions of the l reactor vessel and the increase in RT resulting from neutron irradiationdamagetothelimitingbekk$inematerial. The Prairie Island reactor vessels were procured to ASME Code requirements, which did specify fracture toughness testing to determine the initial RT for each vessel material. ThelicenseeindicatesthattheinitialkY forthelimitingmaterialsintheclosureflangeandbeltlineregionshf the Prairie Island vessels were estimated using the method recommended by the staff in Branch Technical Position MTEB 5-2, " Fracture Toughness Requirements," and from surveillance program data. These methods result in an initial RT f r the limiting beltline base metal and weld metal of 14*F and 0 F,NDTrespectively, and an initial RT f r the limiting closure flange material of -4*F NDT The increase in RT NDT resulting from neutron irradiation damage was estimated by the licensee using the empirical relationship documented in Regulatory Guide 1.99, Rev. 1, April 1977, "Ef.fects of Residual Elements on Predicted Radiation Damage to Reactor Vessel Materials." This method of predicting neutron irradiation damage is dependent upon the predicted amount of neutron fluence and the amounts of residual elements (copper and phosphorus) in the beltline material. The neutron fluence used to predict neutron irradiation damage is based on the calculated neutron fluence at the vessel location with peak flux. These neutron fluence predictions were verified by measurements from passive neutron flux monitors and by analysis, which was made with the DOT Code, a two-dimensional discrete ordinates code. Inputs into the analysis included 47 neutron energy groups, P3 expansion of the scattering cross section, and power distributions representative of time-averaged conditions derived from statistical studies of long-term operation of Westinghouse 2-loop plants. The cross sections used in the analysis were obtained from the SAILOR cross section library. Using this method of analysis, the measured average fast (E > 1.0 MeV) neutron flux derived fggm the five threshold reaction dosimeters in Capsule U in 1.49 x 10 q The design basis calculated fast neutronfluxis1.51x10{gm-seg. q/cm -sec. Since the calculated flux is greater than the measured flux from the capsule dosimetry, neutron fluence calculations using the calculated flux should conservatively predict neutron fluence. An evaluation of the licensee's neutron fluence estimates and protection against pressurizer thermal shock (PTS) ever.ts is elaborated in Enclosure 1, which was discussed in Reference 6.

The prediction curves in Regulatory Guide 1.99, Rev. I are dependent upon the amounts of residual elements in the beltline material. The licensee's letter dated January 10, 1986, identified the residual elements in the core region welds and forgings.

Exhibit E, in the January 10, 1986 letter from the licensee, identified the weld metal with the greatest amount of copper. This weld metal was located in the intermediate to lower shell weld in the reactor vessel of Unit 2. However, in a telecon on September 8, 1986 the licensee and Westinghouse informed the staff that weld metal with the greatest amounts of copper in the intermediate to the lower weld of Unit 2 was only used in the root passes of the weld joint. This weld joint's geometry was described as a double U with the root located half way between the inside and outside surfaces. The thru-wall dimension of the root is approximately 1/2 inch. Since the minimum wall thickness of the beltline weld is 6.7 inches, the weld metal in the root is only a small part of the overall weld and will not significantly affect the weld's fracture toughness. In addition, the location of the root at the mid-wall thickness reduces the effect of neutron irradiation damage to the material. Hence, weld metal with the greatest amount of copper will not be the limiting material and will not affect the proposed pressure-temperature limits. Based on the above discussion, the limiting beltline material for both units is weld metal used to fabricate the intermediate to lower shell weld in Unit 1. This weld metal was used in the Unit I surveillance program. The licensee indicated that information discussed in the telecon will be included in a future revisica to the USAR.

Tables 1 and 2 compare the ART measured from surveillance material inUnits1and2,respectivelyf[othevaluespredictedusing Regulatory Guide 1.99, Rev. 1. Except for HSST Plate 02, the values predicted by the regulatory guide exceed the values measured from the surveillance material. HSST Plate 02 is an experimental heat of material, which was not removed from material used to fabricate the Prairie Island reactor vessels. All other materials identified in Tables 1 and 2 are representative of the materials used to fabricate the beltlines of the Prairie Island reactor vessels. Since the values predicted for ART using Regulatory Guide 1.99, Rev. 1, exceedthevaluesmeasurkhTfrom surveillance materials that represent the Prairie Island reactor vessel beltlines, the prediction method in Regulatory Guide 1.99, Rev. 1, should conservatively predict the ART NDT f r the Prairie Island reactor vessel beltline materials.

The information contained in TS Tables TS 3.1-1 and TS 3.1-2 and Figures TS 3.1-3 and TS 3.1-4 are either reported in the USAR, in regulatory guides issued by the NRC, or in surveillance reports docketed by the licensee. In addition, they are not used during plant operation. Hence, they may be deleted from the plant's TS.

Evaluation The staff has used the nethod of calculating pressure-temperature limits in USNRC Standard Review Plan 5.3.1, NUREG-0800, Rev. 1, July 1981 to evaluate the proposed pressure-temperature limits. The amount of neutron irradiation damage was calculated using design basis calculated neutron fluences and the Regulatory Guide 1.99, Rev. 1, prediction curves. Our

. . j l

1 conclusion is that the proposed pressure-temperature limits meet the safety margins of Appendix G, 10 CFR Part 50 for 15 EFPY and the changes to TS Section 3.1(B) end bases are acceptable.

1 B. Radiation Environmental Monitoring Program 4

Discussion The licensee proposes the following changes to the Radiation Environmental Monitoring Program portion of the technical specifications. '

i

1. In the specification, TS 4.10.A4, Table TS 3.12-2 is referred to 4

regarding radionuclides. The Table number (3.12-2) is incorrect.

The correct Table number should be 4.10-3.

2. The specification Table TS 4.10-1 (Page 4 of 4) requires one sample of corn is to be obtained from a farm having highest D/Q and another sample 10-20 miles away. The corn sampling would be changed by the proposed amendment to allow corn sampling taken from any field that is irrigated by water into which liquid plant wastes have been discharged. The existing specifications assumed agricultural crops of the area would take river water for irrigating the fields in the area. However, over time a uniform process of deep well irrigation has been developed and i the use of river water irrigation has been discontinued.

1 Therefore, corn sampling collection under the existing TS methods is no longer a valid monitoring method to determine the

, impact of liquid releases.

3. The notes on TS Table TS 4.10-2 would be clarified by the i

proposed change by specifying which isotopes apply in considering the lower limit of detection (footnote e) and clarifying note (d) specifying that lower limits of detection of Iodine-131 only applies when I-131 analysis is specified.

i

4. The airborne radioiodine and particulates sample locations have been clarified by eliminating a potential inconsistency that could arise between the sample locations specified in the TS (Table TS 4.10-1, Page 1 of 4) and locations specified in ODCM, Figurc 5.1-1.

Evaluation All changes proposed for Section 4 of the TS dealing with the Radiation Environmental Monitoring Program have been reviewed by the staff. This review disclosed that the proposed changes are merely a L correction in the case of Table TS 4.10-3, an error in corn sampling requirements, and a clarification for the lower limit of detection of radioiodine and sampling location for monitoring airborne radioiodine and particulates. These changes in no way alter the intent of the TS requirements related to the environmental monitoring program nor do they reduce the level of plant safety or alter the protective level of the environmental conditions in the vicinity of the plant

. site. Therefore, the proposed changes are judged consistent with regulatory requirements and, therefore, are acceptable.

I 4

.- ~

C. Design Features-Section 5 Discussion and Evaluation Section 5 of the TS deals with design features of the plant. The ifcensee's proposed changes reflect changes associated with the implementation of 10 CFR Part 50, Appendix I and methods currently used to process radioactive waste. Other proposed changes are administrative in nature, designed to complement the upgrading of the TS.

Section 5.0 of the TS provides a general description of the plant and its systems and does not impose requirements or limiting conditions of operations.

All of these changes were reviewed by the staff. The staff finds that the proposed changes enhance the TS by providing the description of plant systems as they actually exist in the plant and upgrade references to current standards. These proposed changes will not remove or relax any existing requirements that appear in other parts of the TS and, therefore, will have no affect on the probability or consequences of accidents previously considered. In addition, the proposed changes will not remove or relax any existing requirements needed to provide reasonable assurance that the health and safety of the public will be compromised in any way. On this basis, the staff finds the proposed changes acceptable.

D. Security Plan Implementing Procedures Discussion and Evaluation Section 6.5 of the TS is concerned with the plant operating procedures that shall be reviewed by the Operations Committee and approved by a member of plant management. The proposed change would clarify that the Operations Committee would not be required to review nonsafety-related Security Plan implementing procedures and this review function would be exclusively performed by the guards. A similar request dealing with the clarification of the reviews of the nonsafety-related Security Plan implementing procedures was found acceptable by the staff for the Monticello Nuclear Generating Plant (Docket No. 50-263) as noted by Amendment No. 25 dated August 15, 1984. It was never the intent of the staff to have the Operations Committee review nonsafety related Security Plan implementing procedures. Therefore, such an interpretation of this administrative requirement is considered to be in error. The staff has reviewed the licensee's proposed change and finds that it would eliminate this potential misinterpretation in the administrative requirements of the TS. On this basis, we find the change acceptable.

l

E. Administrative-Section 6 Discussion and Evaluation The licensee's proposed changes involve establishing a title to the address for submitting the monthly operating report and deleting the record keeping requirements for the environmental qualification records.

The proposed changes would impact TS Sections TS 6.7.A.3 and TS 6.8. The change involving the address for submitting the monthly report is editorial in nature and has no affect on the plant safety level. TS 6.8 dealing with environmental qualification of safety-related equipment specified the schedular requirements and the interim guidelines of NUREG-0588 which have become obsolete and does not serve a useful purpose with the ' <suance of the rule in 10 CFR 50.49. The recent change to 10 CFR Part 50 to include section 50.49 specifies the requirements for environmental qualification of electrical equipment including the record keeping which the licensee must comply. Thus, the deletion of TS 6.8 would eliminate this duplication and potential confusing requirement concerning environmental qualification of electrical equipment. In addition, the deletion of this requirement from the TS will in no way result in a reduction in the plant margin of safety. On the above basis, we find the proposed change deleting section 6.8 and 6.6.B.11 acceptable.

ENVIRONMENTAL CONSIDERATION 4

These amendrents involve a change in the installation or use 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 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 published a proposed finding that these amendments involve no significant hazards consideration and there has been no public comment on such finding.

Accordingly, these amendments meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9).

These amendments also involve changes in recordkeeping, reporting or administrative procedures or requirements. Accordingly, with respect to these items, the amendments meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(10). 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.

, - - - - - , , - - - g -

- - - - - - - - - , _..._..,.,,n .,-m. , - , - , - - - - , . - . , - , , . - , , ,.n.,. .,.n. ., -. -.

,,,,,-,,,,..,.--n,n--n-,,.r-

. ~

CONCLUSION We have 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, and (2) such activities will be conducted in compliance with the Commission's regulations, and the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

Principal Contributors:

B. Elliot C. Willis L. Lois D. Dilanni Date: November 14, 1986

References:

(1) Westinghouse Report WCAP 8916, " Analysis of Capsule V From Northern States Power Company Prairie Island Unit No. 1 Reactor Vessel Radiation Surveillance Program", Davidson et al, August 1977.

(2) Westinghouse Report WCAP 10102, " Analysis of Capsule P From Northern States Power Company Prairie Island Unit 1 Reactor Vessel Radiation Surveillance Program", Yanichko et al, May 1982.

(3) Westinghouse Report WCAP 11006, " Analysis of Capsule R From Northern States Power Company P,rairie Island Unit 1 Reactor Vessel Radiation Surveillance Program", Boggs et al, February 1986.

(4) Westinghouse Report WCAP 9212. " Analysis of Capsule V From Northern States Power Company Prairie Island Unit No. 2 Reactor Vessel Radiation Surveillance Program", Davidson et al, November 1977.

(5) Westinghouse Report WCAP 9877, " Analysis of Capsule T From Northern States Power Company Prairie Island, Unit No. 2 Reactor Vessel Radiation Surveillance Program", Yanichko et al, March 1981.

(6) Memorandum from C. Berlinger to D. Dilanni dated August 28, 1986.

. ~

TABLE I Comparison of Increase in Reference Temperature (ARTNDT)

Measured from Unit 1 Surveillance Material to the ART Predicted byRegulatoryGUkbe1.99,Rev.1 RT PredYNedby RT Measured Regulatory Fro $Dkurveillance Guide 1.99 Capsule / Surveillance Neutron Fluence Material Rev. 1 Material (X1018q/cm2) (op) (op)

Capsule R(*)

a Weld Metal 4.03 117 271 Forging C (Tang. orient.) 4.03 87 131 (Axial orient.) 4.03 80 131 HSST Plate 02 4.03 186 241 Capsule P(b)

Weld Metal 1.25 42 151 Forging C (Tang. orient.) 1.25 20 73 (Axial orient.) 1.25 37 73 HSST Plate 02 1.25 156 134 Capsule V(c)

Weld Metal .546 25 100 Forging C (Tang. orient.) .546 38 48 (Axial orient.) .546 24 48 HSST Plate 02 .546 110 89 (a) Data from Reference 3 (b) Data from Reference 2 (c) Data from Reference 1

Table II Comparison of Increase in Reference Temperature (ARTNDT)

Measured from Unit 2 Surveillance Material to the ART Predicted byRegulatoryGUkhe1.99,Rev.1 RT PredYk[edby RT Measured Regulatory Fro $Ddurveillance Guide 1.99 Capsule / Surveillance Neutron Fluence Material Rev. 1 Material (X1018q /cm2) (op) (op)

Capsule T(a)

Weld Metal 1.05 60 99 Forging (Tang. orient.) 1.05 55 62 (Axial orient.) 1.05 35 62 HSST Plate 02 1.05 160 123 Capsule V(b)

Weld Metal .586 60 74 Forging (Tang orient.) .586 35 46 (Axial orient.) .586 30 46 HSST Plate 02 .586 125 92 (a) Data from Reference 5 (b) Data from Reference 4 l

l l

l

Enclosure 1 Prairie Island Units 1 and 2, Fast Neutron Fluence Reevaluation for the Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events (10 CFR 50.61)

By letter dated July 15, 1986, Northern States Power Company, (the licensee) for the Prairie Island Units 1 and 2 plants, requested a license amendment for the Pressure-Temperature curves revision as required by 10 CFR Part 50, Appendix G (Reference 1). The Pressure-Temperature curves and the Pressure Vessel fast neutron (E 1 1.0 MeV) fluence were reevaluated for 15 effective full power years. The projected estimate of the revised fluence was found to be 21%

higher than the original estimate. Then, according to 10 CFR 50.61, the licensee submitted the new projected fluence and RT values. In addition, PTS in the PTS evaluations, the staff requested that the licensee submit the RT PTS reevaluations along with the Appendix G submittals (Reference 2).

The following is the 60 effective full power years RT for Prairie Island PTS Units 1 and 2 with the new values of the fluence. The equation specified in 10 CFR 50.61 as applicable for the Prairie Island Units 1 and 2 is:

RT 0 PTS = I+M+(-10+470.Cu+350.Cu.Ni)f .27 where: Unit 1 Unit 2 I = Initial RT NDT

= 0*F O'F M = Uncertainty Margin = 59*F 59'F Cu = w/o Copper in weld W-3 = 0.14 0.19 Ni = w/o Nickel in' weld W-3 = 0.17 0.13 f = peak fluence (E 1 1.0 MeV) on weld W-3 x 10 18 cm2/n = 9.75 9.75 Then: Unit 1 RT 0 PTS = 59 + (-10 + 470 x 0.14 + 350 x 0.14 x 0.17) x 9.75 27

= 59 + 64.1 x 1.85 n 177.6 F i

l A

Unit 2 0

RTPTS = 59 + (-10 + 470 x 0.19 + 350 x 0.19 x 0.13) x 9.75 27

= 59 + 87.95 x 1.85 = 221.7*F The applicable criteria specified in 10 CFR 50.61 is 300 F, therefore, the revised values meet the criteria for 60 effective full power years by a large margin.

References

1. Letter from D. Musolf, Northern States Power Company, to Director, NRR, dated July 15, 1986.

2. Memoranda from C. E. Rossi to D. Dilanni, " Fast Neutron Fluence for the Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events, dated April 28 and May 12, 1986, for Units 2 and 1, respectively.

i l

N -, . . . . . _ - - . _ - - . . , . - - - . - -

Distrbution Copies:

{2DocketFile..,50-282/306 NRC PDR Local PDR PAD #1 r/f PAD #1 p/f TNovak, Actg Div Dir GLear DDiIanni PShuttleworth NThompson, DHFT OGC-Bethesda LHarmon EJordan BGrimes JPartlow EButcher, TSCB TBarnhart (8)

WJones FOB, DPLA ACRS (10)

OPA LFMB ( TAC No. 62137 & 62138) i 1

1

,, . . - - _ _ _ _ , - - , . - - - - - - _ _ - - - - . - - . - . , , - , . . - . . . . - - . - - , . - - - - . . . . . . - - - - - , , _ . - . . . - .