ML20217R052
| ML20217R052 | |
| Person / Time | |
|---|---|
| Site: | Prairie Island  |
| Issue date: | 08/29/1997 |
| From: | Sorensen J NORTHERN STATES POWER CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML20217R053 | List: |
| References | |
| PROJECT-689 TAC-M98944, TAC-M98945, NUDOCS 9709040148 | |
| Download: ML20217R052 (10) | |
Text
.
Northern States Power Company Prairie Island Nuclear Generating Plant 1717 Wakonade Dr. East Welch, Minnesota $5089 August 29,1997 U S Nuclear Regulatorv Commission Attn: Document Control Desk Washington, DC 20555 PRAIRIE ISLAND NUCLEAR GENERATING PLANT Docket Nos. 50-282 License Nos. DPR-42 50-306 DPR-60 Response to Request for Additional Information Related to License Amendment Request Dated May 15,1997 incorporation of Voltage-Based Steam Generator Tube Repair Criteria (TAC Nos. M98944 and M98945)
By letter dated August 13,1997, ?.he NRC Staff requested additional information regarding our License Amendment Request Dated May 15,1997 whicI proposed changes to Section 4.12 of the Prairie Ish,d Technical Specifications that would allow the use of voltage-based steam generator tube repair criteria. Our response to the August 13,1997 request for additional information is provided as Attachment 1.
In a related matter, NSP has determined that the allowable main steam line break leakage in gallons per minute, utilized in the May 15,1997 License Amendment Request, was calculated at operating temperature and the projected main steam line break leakage is calculated at room temperature. As described in Attachment 2, all references to 6.4 and 6.5 GPM in the May 15,1997 License Amendment Request should be changed to 4.67 and 4.77 GPM, respectively (at 70 degrees F).
In this letter we have made no new Nuclear Regulatory Commission commitments.
Please contact Gene Eckholt (612-388-1121) if you have any questions related to this response.
}r g
i el P Sorensen
(,'j^7 Plant Manager Prairie Island Nuclear Generating Plant b!!$fl$!!!lfhlfl!flflllllll 9709040148 970829 PDR ADOCK 05000282 P
PDR u.
t.
i USNRC.
NORTHERN STATES POWER COMPANY August 29,1997 -
Page 2'--
j c: Regional Administrator - Region Ill, NRC :
Senior Resident inspector, NRC NRR Project Manager, NRC J E Silberg Attachments:
Affidavit- - Response to August 13,1997 NRC Request For Additional Information l- - Correction to Allowable Main Steam Line Break Leakage
, - NEl Letter Dated October 15,1996, " Response to NF.C Letter Dated February 9,1996, RegardinD New Probe Variability Criteria"
(
i UNITED STATES NUCLEAR REGULATORY COMMISSION NORTHERN STATES POWER COMPANY PRAIRIE ISLAN'D NUCLEAR GENERATING PLANT DOCKET NO. 50-282 50-306 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION RELATED TO REQUEST FOR AMENDMENT TO OPERATING LICENSES DPR-42 & DPR-60 LICENSE AMENDMENT REQUEST DATED May 15,1997 i
Northern States Power Company, a Minnesota corporation, by this "DD-97-18, Directors Decision DD-97-18 Re 970528 Petition Filed by Prairie Island Indian Community Requesting That USNRC Take Action to Determine That NSP Violated Requirements of [[CFR" contains a listed "[" character as part of the property label and has therefore been classified as invalid..Request Denied Due to Listed Reasons|letter dated August 29,1997]], with Attachments 1,2 and 3 provides a response to an NRC Staff request for additional information in support of the subject License Amendment Request dated November 27,1996. Attachment 1 provides the response to the August 13,1997 NRC Request for Additional Information._ Attachment 2 provides a discussion of a correction to the May 15,1997 License Amendment Request. Attachment 3 provides a copy of an NEl letter referenced in the response to the August 13,1997 request for additional information.
This letter contains no restricted or other defense information.
NORTH RN STAT S P WER COMPANY i
By e
gel P $3orense'n Plant Manager Prairie Island Nuclear Generating Plant On thid ay of aa 89'7 before me a notary public in and for said County, personally appq6re)f Joel P Sorensen, Plant Manager, Prairie Island Nuclear Generating Plant, and beTrfg first duly sworn acknowledged that he is authorized to execute this document on behalf of Northern States Power Company, that he knows the contents thereof, and that to the best of his knowledge, information, and belief the stagments made in it are true and t jt is not interposed for delay, g)h
[
'i T' /
=:::. ::::.: ::::::.::::::::::::-
MAICIA K. LaCORE NOTARYPUBuc4ApetESOTA HENNEPWCOUNTY 1%Cassimeion ExptuJan.31,t000 w :: : : :::.:.:::::::::::::::.::::
-.. - -._ - ---.---.- -~-
f g
4 ATTACHMENT 1 Response to Auoust 13.1997 NRC Reauest For Additional Information i
I The following information is provided in response to the Request for Additional Information regarding our Licent.,e Amendment Request Dated May 15,1997 which proposed changes to Section 4,12 of the Prairie Island Technical Specifications that would allow the use of voltage-based steam generator tube repair criteria:
RESPONSE TO NRC STAFF QUESTIONS i
ll Question 1:
l Page 2 of Exhibit D. Northern States Power (NSP) states that the latest at roved
[
EPRI [ Electric Power Research Institute) database (7/8-inch diameter tubing) will be used in performing Generic Letter (GL) 95-05 (" Voltage-Based Repair Criteria for Westinghouse Steam Generator Tubes Affected By Outside Diameter Stress Corrosion Cracking") specified calculations. The staff has agreed,in principle, with the Nuclear Energy Institute (NEI) on a protocol by which the industry will periodically update the database used to perform GL 95-05-specified calculations
(
Reference:
Letter from Brian Sheron, NRC, to David Modeen, NEl, dated April 10, 1997). Licensees, including NSP, should follow the guidance of the protocol and 1
provide the staff with the latest database used in performing GL 95-05 specified -
l calculations.
Response
Following final NRC epproval of the NEl _ voltage based alternate repair criteria
[
database protocol via letter to NEl. NSP will follow the protocol and use the latest
_ industry database following that protocol. Until that time, NSP will follow the guidance i
of the protocol provided to the NRC by NEl on January 15,1997 and describe the database used in performing Generic Letter 95-05 specified calculations as part of.the L
Generic Letter 95-05 90-day report. In the future, under the proposed protocol, each l
updated version of the database will be provideo to the NRC by the industry and NSP would only provide reference to the version used.
i
- Questbn 2:
i.
Page 2 of Exhibit D. NSP states that tube integrity evaluation will be consistent k
with Westinghouse WCAP-14277, "SLB Leak Rate and Tube Burst Probability Analysis Methods for ODSCC at TSP intersections." However, on page 9 of i-
Attachmint 1 August 29,1997 Page 2 Exhibit A, NSP references WCAP 14277, Revision 1. Please varlfy the WCAP ravision used to evaluate tube integrity.
Response
NSP wili use the methodology of Revision 1 or later of WCAP-14277 to evaluate tube integrity.
Question 3:
Address how guidelines in Section 2 of Attachment 1 to GL 95-05 would be implemented. Specifically, (a) discuss the differential pressure (across the tube walls) required for conditional tube burst probability; [The differential pressure is dependent on the operability requirements of the pressurizer power-operated relief valves (PORVs) during power operation.] and (b) verify that Prairie Island satisfies the requirements of GL 90-06, " Resolution of Generic issue 70, ' Power-Operated Relief Valve and Block Valve Reliability,' and Generic issue 94,
' Additional Low temperature Overpressure Protection for Light-water Reactors,'
pursuant to 10 CFR 50.54(f)."
Response
NSP will evaluate the conditional tube burst probability at an assumed differential pressure across the tube walls equal to the pressurizer safety valve setpoint plus 3 l
percent for the valve accumulation, less atmospheric pressure. The nominal pressurizer safety valve setpoint at Prairie Island is 2485 psig. With 3% accumulation, the assumed differential pressure for the tube integrity evaluations is 2560 psig. This is consistent with page 10 of E::hibit A.
Question 4:
Page 4 of Exhibit D. It is stated that the industry methodology will be used for new probe variability as approved by the NRC. Discuss which industry methodology will be used and provide reference documents that contain the methodology.
Response
The methodology to be used foi new probe variability wi:1 utilize the procedures described in the October 15,1996 letter (Attachment 3) from NEl (Alexander Marion) to the NRC (Dr. Brian Sheron), " Response to NRC Letter Dated February 9,1996, Regarding New Probe Variability Criteria" as approved by the NRC in the July 29,1997 letter from NRC (Brian Sheron) to NEl (David Modeen), " Probe Variability Criteria." This methodology requires that the primary frequency and mix frequency voltage responses i
Attachmont 1 August 29,1997 Page 3 of new probes be within i 10% of the nominal voltage responses when voltages are normalized to the 20% flaw values. The nominal voltage responses were established as the average voltages obtained from ASME standard drilled hole flaws for at least 10 production probes.
In order to be used for the voltage repair criteria, new production probes must have a voltage response within 10% of the nominal voltage values for the ASME hole flows at both the primary and mix frequencies. The nominal (reference) voltages were included in the October 15,1996 NF.I letter to the NRC (Attachment 3). Use of these acceptance criteria by the vendor supplyin0 bobbin coil probes to be used for the voltage based alternate repair criteria inspection fulfills the requirements of Section 3.c.2 of Generic Letter 95-05 for new probe variability.
Question 5:
j Page 4 of Exhibit D. It is not clear to the staff that NSP has committed to the guidelines in Sections 3.b.2,3.b.3,3.b.4,3.c.1,3.c.2,3.c.5,3.c.6,3.c.7, and 3.c.8 of to GL 95 05. Please confirm and/or address whether NSP has committed to the above mentioned guidelines.
Response
Paragraph 8 on Page 4 of Exhibit A addresses the requirements of Section 3 of Generic Letter 95-05 and commits NSP to the requirements of Sections 3.b.2,3.b.3, and 3.b.4.
Paragraph 3 on Page 3 of Exhibit A and Section 3 of Exhibit D commits NSP to the requirements of Section 3.c of Generic Letter 95-05 unless exception was taken.
l To confirm this commitment, NSP will:
3.c.1: Calibrate the bobbin coils against a transfer standard 3.c.2: Use bobbir' coil probes supplied in accordance with the methodology described in the October 15,1996 letter (Attachment 3) from NEl (Alexander Marion) to the NRC (Dr. Brian Sheron), " Response to NRC Letter Dated February 9,1996, Regarding New Probe Variability Criteria" as approved by the NRC in the July 29,1997 letter from NRC (Brian Sheron) to NEl (David Modeen);" Probe Variability Criteria."
3.c.5: Establish quantitative acceptance criteria for noise criteria.
i
e Att: chm:nt i
~
August 29,1997 Page 4 3.c.6: Include guidance in the analysis guidelines for reviewing and minimizing mixed residuals on the transfer standard.
3.c.7: Conduct a site specific demonstration for probe diameters other than 0.720 inch and submit this data to the NRC for approval or utilize an NRC approved industry submittal for 7/8 inch tubes prior to leaving tubes in service using smaller probes.
3.c.8: Train data analysts on the potential for primary water stress corrosion cracking at TSP intersections.
Question 6:
Section 5.b of Attachment 1 to GL 95-05 recommends that the effectiveness of monitoring procedures for ensuring the timely detection, trending, and response to rapidly increasing leaks should be assessed. In addition, the appiopriateness of alarm setpoints on the primary to secondary leakage detection Instrumentation and the various criteria for operator action in response to l
l detected leakage should also be assessed. Discuss whether the above assessments have been performed to satisfy Section 5.b of Attachment 1 to GL 95-05.
t Resoonse:
These assessments were conducted when NSP significantly revised the Prairie Island steam generator tube leak abnormal operating procedures to incorporate the EPRI Report TR-104788, "PWR Primary-to-Secondary Leak Guidelines." Prairie Island has incorporated guidance for shutdown at the recommended lower leakage limit of 150 GPD and at 60 GPD/HR change in leak rate. Major aspects of the procedure include action levels at 5 GPD,30 GPD, and 150 GPD leak rates and 60 GPD/HR leak rate change. Plant shutdown is directed when Action Level 2 (Shutdown) takes place based on one quantitative indication confirmed by one qualitative indication. Plant shutdown in about 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> is directed when the 60 GPD/HR leak rate increase is exceeded.
Operators received classroom and simulator training on these revised procedures both in 1996 and 1997. Two sets of alarms are used: Air ejector instrument alarms set at low levels to warn of changing conditions which are reset as leakage increases and plan'. process computer alarms set at just below the Action Levels of 30 GPD and 150 GPD leak rate and 60 GPD/HR rate of change. The abnormal operating procedure also includes procedures for manually assessing steam generator tube leakage when the plant process computer is out of service.
Timely detection is provided by high availability of the air ejector radiation monitor and steam generator blowdown radiation monitors and low alarm setpoints to give early
Attachm:nt 1 August 29,1997 Page5 warning of changing conditions.-- Manual sampling is required at a frequency.
dependent on the leak rate when the air ejector radiation monitor is out of service.
Trending is readily available in the control room using the plant process computer, Manual trending is required by procedure when the computer is out of service.
Trending of the rate of change of leakage is conducted either by computer or manually, if needed. Operators have been trained on response to rapidly changing leak rates.
Trend plots are available in the control room to continuously monitor steam generator
. tube leakage based on air ejector radiation monitors with the capability to monitor as CPM or GPD and GPD/HR.
ATTACHMENT 2 Correction to Allowable Main Steam Line Break Leakage After additional review of the Main Steam Line Break dose calculation performed for NSP, it has been determined that the leak rate of 6.5 GPM (intact plus faulted steam.
generator leakage) esteblished in the Main Steam Line Break dose analysis for the license submittal dated May 15,1997 is at a temperature of 567.3*F and 2250 psi.
Fluid density at those operating conditions is 45.7 lb./cu.ft. At room temperature (70
'F), fluid density is 62.3 lb1cu.ft. Therefore, the allowable total MSLB leakage (intact plus faulted steam generator) becomes 4.77 GPM.
The assumptions for tube leakage are mass flow rates equivalents of 17,736 gm/ min from the faulted steam generator, and 288.2 gm/ min for the intact steam generator. It is conservative to assume the intact steam generator leakage has the same effect as the faulted steam generator leakage and subtract the additional intact steam generator -
leakage due to the desity correction from the faulted steam generator leakage. - Since i
150 GPD at room temperature is 393,9 gm/ min, the allowable faulted steam generator leakage becomes 17736 - (393.9 - 288.2) = 17,630 gm/ min which is 4.67 GPM.
The allowable projected leakage must remain at or below 4.67 GPM in the faulted steam generator, Total allowable leakage for both the intact and faulted loop is 4.77 GPM.
e
.._-m__.
I NEl Letter Dated October 15,1996 t
" Response to NRC Letter Dated February 9 1996 Regarding i
New Probe Variability Criteria" 1-t i
i f
f 5
i
)
i a
4 n
J a
v 1
?
b' i
6 e
i i
i s
e 4
i
.,