ML20151J577
| ML20151J577 | |
| Person / Time | |
|---|---|
| Site: | Prairie Island  |
| Issue date: | 07/29/1997 |
| From: | Richard Anderson NORTHERN STATES POWER CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| GL-97-01, GL-97-1, NUDOCS 9708050182 | |
| Download: ML20151J577 (10) | |
Text
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Northern States Power Company Prairie Island Nuclear Generating Plant 1717 Wakonade Dr. East Welch. Minnesota 55089 July 29,1997 Generic Letter 97-01 U S Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555 PRAIRIE ISLAND NUCLEAR GENERATING PLANT Docket Nos. 50-282 License Nos. DPR-42 50-306 DPR-60 120 Day Response to Generic Letter 97-01 Degradation of Control Rod Drive Mechanism Nozzle and other Vessel Closure Head Penetrations Generic Letter 97-01 was issued to obtain plant specific information to describe their program for ensuring the timely inspection of PWR control rod drive mechanism (CRDM) and other vessel closure head penetrations. The NRC staff requested submittal of our response to the Generic Letter within 120 days.
Our resoonse to the 120 day Generic Letter request is provided in Attachment 2 of this submittal. In the response we have made two new Nuclear Regulatory Commission commitments, indicated as the two italicized paragraphs in response to information request section 1.4.
Please contact Jack Leveille (612-388-1121, Ext. 4662) if you have any questions related to this letter.
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'R O Anderson j
Director
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Nucle?r Energy Engineering I
Copies and attachments: (next page) l'lllIl'lllh"llll'{l!l{IlllllJIIllll l
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i-USNRC NORTHERN STATES POWER COMPANY 1
July 29,1997 Page 2 1
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c: Regional Administrator - Region 111, NRC Senior Resident inspector, NRC NRR Project Manager, NRC J E Silberg Attachments:
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- 1. Affidavit F
- 2. Generic Letter 97-01 120 day response
- 3. Prairie Island Units 1 & 2 Input Values For Probabilistic Analysis
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s 97-01,2 DOC
UNITED STATES NUCLEAR REGULATORY COMMISSION NORTHERN STATES POWER COMPANY PRAIRIE ISLAND NUCLEAR GENERATING PLANT DOCKET NO. 50-282 50-306 GENERIC LETTER 97-01, Degradation of Control Rod Drive Mechanism
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Nozzle and other Vessel Closure Head Penetrations - 10 CFR 50.54(f)
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Northern States Power Company, a Minnesota corporation, with this letter is submitting information requested by NRC Generic Letter 97-01.
This letter contains no restricted or other defense information.
1 NORTHERN STATES POWER COMPANY
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iRogpf0 Aiiderson Director Nuclear Energy Engineering 1
County, p rsonally appe(rpd Rop %
On this ay of r O Anderson, Director, Nuclear Energy Engineering; and being first before me a notary public in and for said duly sworn acknowledge &that he is authorized to execute this document on behalf of Northern States Power Company, that he knows the contents Jh,reof, and that to the best of his knowledge, information, t
and b f the statements m de in i are tru 4 it is not int 'rposed for delay.
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GENERIC LETTER 97-01 120 DAY RESPONSE l
Generic Letter 97-01 (GL) Degradation of Control Rod Drive Mechanism Nozzle and Other Vessel Closure Head Penetrations, was issued to request licensees to describe their program for insuring the timely inspection of PWR control rod drive mechanism (CRDM) and other closure head penetrations. This response provides Prairie Island information relative to the information requested by the GL.
Prior to issuance of the GL, Prairie Island has worked with the Westinghouse Owners Group (WOG), the Electric Power Research Institute (EPRI) and the Nuclear Energy Institute (NEI) to understand the operational experience, identify technical issues, cause factors, relative importance, and solutions. One of these tasks was the development of safety evaluations that characterized the initiation of damage, propagation and consequences. These safety evaluations are contained in WCAP 13565' and are applicable to Prairie Island. The NRC reviewed the safety evaluations and issued a safety evaluation report (SER) to NEl on November 19,1993. The safety evaluations and the SER establish the basis for Prairie Island continued operation.
Addressee is required to provide the following information:
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Response
1.
Regarding inspection activities:
1.1 A description of allinspections of CRDM nozzle and other VHPs performed to j
the date of this generic letter, including the results of these inspections.
ASME Boiler & Pressure Vessel Code (B&PVC)Section XI Inservice inspection (ISI) (B4.12] VT-2 visual inspections of 100% of the CRDM surface area above the head insulation during Inservice and Hydrostatic System Pressure Tests.
ASME B&PVC Section XI ISI [B15.10,11] VT-2 visual inspections of the general head area, penetrations and canopy seals, above the head insulation during Inservice and Hydrostatic System Pressure Tests.
ASME B&PVC Section XI ISI [B14.10] Volumetric or Surface inspection of 10% Peripheral CRDM housing welds.
GL-88-05 Shutdown visual inspections for leakage on the penetrations and canopy seals above the head insulation.
Visualinspections of the head and penetrations at insulation removal areas during canopy seal weld repair work.
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July 29,1997 i
Page 2
'j To this date - no indication of a Reactor Vessel Head, or Head Penetration, leak has been identified.
1.2 If a plan has been developed to periodically inspect the CRDM nozzle and other VHPs:
l A. Provide the schedule for first, and subsequent, inspections of the CRDM nozzle and other VHPs, including the technical basis for this schedule.
i See 1.4 response below.
B. Provide the scope for the CRDM nozzle and other VHP inspections, including the total number of penetrations (and how
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many will be inspected), which penetrations have thermal sleeves, j
which are spares, and which are instrument or other penetrations.
See 1.4 response below and:
1 Each Prairie Island Reactor Vessel has 40 CRDM head penetrations and 1 Head Vent penetraticn with the following i
breakdown:
TYPE REMARKS 1,10 thru 33 Full Length CRDM Thermal Sleeves 38 thru 41 Full Length CRDM Thermal Sleeves i
1 6,7,8,9 Part Length CRDM 2,3,4,5 Spare No Thermal Sleeves 34,35,37 Inst Port T/C No Thermal Sleeves i
Head Vent No Thermal Sleeves I
- Part length CRDMs are not used and the drive shafts have been retracted to full up positions and locked in place. No Thermal Sleeves are provided on PL CRDMs.
97-01_2. DOC
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July 29,1997 Page 3 1.3 If a plan has not been developed to periodically inspect the CRDM nozzle and other VHPs, provide the analysis that supports why no augmented inspection is necessary.
4 This question is not applicable.
j 1.4 in light of tne degradation of CRDM nozzle and other VHPs described above, j
provide the analysis that supports the selected course of action as listed in either j
1.2 or 1.3, above. In particular, provide a description of all relevant data and/or tests used to develop crack ini.tiation and crack growth models, the methods and l
data used to validate these models, the plant-specific inputs to these models, and how these models substantiate the susceptibility evaluation. Also, if an l
Integrated industry inspection program is being relied on, provide a detailed j
description of this programs.
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l The data, tests, and methods usedin developing the crack initiation and j
crack growth models on which Northem States Power Company management strategy for addressing the RPVHP cracking issue is based are provided in Sections 2 and 3 of WCAP-14901.
Prairie Island is a participant in the Westinghouse Owners Group analysis program in which a plant specific probability analysis using the methodology describedin Section 4 of WCAP-14901 has been performed. The plant specific input parameters to the analysis are
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provided in Attachment 3. The analysis results will be incorporated into the WOG/NElintegrated inspection program for use in determining the need for a plant specific inspection. This integrated program includes all three PWR owners groups, the Electric Power Research Institute, and the Nuclear Energy Institute who are cooperatively working to compile information on the estimated operating time from January 1,1997, needed to initiate and propagate a crack 75% through wallin a vesselpenetration for all of the heads in the United States. This information will be evaluated to determine if an adequate number of plants have or are planning to inspect in the near future. This evaluation willbe completed and detailed inspection plans for the industry will be provided to the NRC by the end of 1997.
2.
Provide a description of any resin bead intrusions, as described in IN 96-11, that have exceeded the current EPRI PWR Primary Water Chemistry Guidelines recommendations for primary water sulfate levels, including the following information:
97-01,2. DOC w.
1 July 29,1997 Page 4 2.1 Were the intrusions cation, anion, or mixed bed?
See 2.6 Response below.
2.2 What were the durations of these intrusions?
See 2.6 Response below.
2.3 Does the plant's RCS water chemistry Technical Specifications follow the EPRI guidelines?
See 2.6 Response below.
2.4 Identify any RCS chemistry excursions that exceed the plant administrative limits for the following species: sulfates, chlorides or fluorides, oxygen boron, and lithium.
See 2.6 Response below.
2.5 Identify any conductivity excursions which may be indicative of resin intrusions. Provide a technical assessment of each excursion and any e
follow up actions.
See 2.6 Response below.
2.6 Provide an assessment of the potential for any of these intrusions to result in a significant increase in the probability for IGA of VHPs and any associated plan for inspections.
Northern States Power Company has reviewed the plant historical records to determine if any incident of resin ingress similar to those which occurred in 1980 aiid 1981 at the Jose Cabrera (Zorita) plant has occurred at Prairie Island. This data search is structured to identify all resin intrusion events into the primary coolant system that were of a magnitude greater than 1 ft* (30 Liters). The threshold of 1 ft' was chosen as a conserva'ive lower bound since it represents less than 15% of the estimated volume of resin 97-01_2. DOC
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July 29,1997 Page 5 released into the reactor coolant system during the two events at Jose Cabrera.
For the period of plant operation prior to the routine analysis for sulfate in reactor coolant, the data search was based on a review of the plants reactor coolant chemistry records relative to specific conductance of the reactor coolant. An elevation cf a 28 micro S/cm increment in specific conductance was the value used as an indicator of cation resin ingress equivalent to a volume of 1 ft.
2 Routine analysis for sulfate in reactor coolant was performed for plant operation from February 3,1984 to the present. A sulfate concentration in the range of 15 to 17 ppm peak concentration was used as the indicator of cation resin ingress. This concentration is approximately equivalent to a volume of 1 ft.
Had either specific conductance or sulfate increases indicated resin ingress to the magnitude of the threshold quantity identified above, additional data evaluation would have been conducted to look for a corresponding depression in pH or elevation in lithium as corroborating information of the incident. In the case of the use of sulfate data as the indicator, specific conductance would also have been included as confirmatory data had a significant in-leakage event been identified.
It is unnecessary to review plant records for boron, chlorides, fluorides and oxygen because these species are not viewed as valid indicators of cation resin ingress and degradation within the primary coolant system of a PWR. Borate, chloride and fluoride anions could be associated with the anion portion of mixed bed resin (cation plus anion); however, if mixed bed resin leakage to the RCS occurred, the cation portion of the resin would contain the sulfate indicator doscribed above. Detectable dissolved oxygen in reactor coolant, during power operation with appropriate hydrogen overpressure on he volume control tank and specified residual dissolved hydrogen in the reactor coolant, could not occur and, therefore, could r.M be associated with resin in-leakage.
Prairie Island has followed the EPRI PWR Primary Water Chemistry Guidelines since March 1989 and has implemented revisions when issued.
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July 29,1997 Page 6 i
The following exception to the EPRI guidelines exist at Prairie j
island; l
RCS suspended solids performed only prior to rod movement followino outr.ges when the l
RCS is opened.
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Based on the above criteria Prairie Island has not experienced a resin intrusion into the reactor coolant system. Note -- NSP has previously responded to these concerns raised in NRC Request For Additional Information letter dated l
September 25,1995 with an October 24,1995 response letter.
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References:
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' Westinghouse WCAP-13565, Revision 1 - Alloy 600 Reactor Vessel Head Adaptor i
Tube Cracking Safety Evaluation.
2 Westinghouse WCAP-14901, Revision 0 - Background and Methodology for Evaluation of Reactor Vessel Closure Head Penetration Integrity for the Westinghouse Owners Group.
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97-01 2, DOC
l Prairie Island Units 1 and 2 Input Values for Probalistic Analysis 1
TABLE 5-2 PRAIRIE ISLAND UNIT 1 INPUT VALUES FOR FROBABILISTIC ANALYSIS Case Pen.No.
Temp.
Set-up Y.S. (ksi)
GBC (%)
Angle (')
1 34,35,37 580.2*F 42.9 37.3 82.2 2
26,27 36.5 37.3 82.2 3
28,29
'3'6.5 36.1 91.3 4
30 thru 33 36.5 36.7 89.1' 5
22 thru 25 31.4 36.1 91.3 6
14 thru 17 31.4 41.7 83.5 7
18 thru 21 29.7 36.7 89.1 8
12,13 27.8 36.7 89.1 9
10 11 27.8 36.1 91.3 10' 5
19.3 37.3 82.2
- This case is also used to bound penetrations 1 through 4,6 through 9, and 38 through 41.
i TABLE 5-2 PRAIRIE ISLAND UNIT 2 INPUT VALUES FOR PRoBABILISTIC ANALYSIS Caso Pen.No.
Temp.
Set-up Y.S. (ksi)
GBC (%)
Angle (*)
1 34,35,37 580.2*F 42.9 47.5 86.5 2
26 thru 33 36.5 47.5 86.5 3
14 thru 17 31.4 47.5 86.5 22 thru 25 4
18 thru 21 29.7 47.5 86.5 5
10 thru 13 27.8 47.5 86.5 6
2 thru 5 19.3 47.5 86.F 7*
6 thru 9 13.5 47.5 86.5
- This case is also used to bound penetrations 1 and 38 through 41.
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