ML20098B184
| ML20098B184 | |
| Person / Time | |
|---|---|
| Site: | Monticello  |
| Issue date: | 09/26/1995 |
| From: | Hill W NORTHERN STATES POWER CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| TAC-M90979, NUDOCS 9510020348 | |
| Download: ML20098B184 (12) | |
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I Northom States Power Company Monticello Nuclear Generating Plant 2807 West Hwy 75 Monticello, Minnesota 55362-9637 September 26,1995 Response to Request for Additional information TAC No. M90979 U S Nuclear Regulatory Commission Attn:
Document Control Desk Washington, DC 20555 t
MONTICELLO NUCLEAR GENERATING PLANT Docket No. 50-263 License No. DPR-22 Additional Information Main Steam Isolation Valve Testina The purpose of this letter is to respond to the NRC Staffs Request for Additional information (TAC No. M90979) dated August 28,1995. The specific information requested by the Staff is included in Attachment 1.
Your August 28,1995 letter indicated that the failure of main steam isolation valves (MSIVs) to pass 10 CFR 50, Appendix J, testing at the beginning of the 1994 Monticello refueling outage is being reviewed for abnormal occurrence reportability. The details of this event were reported to the Commission in Licensee Event Report 94-10 dated October 19,1994.
In the NRC Staffs evaluation of this event, we believe a number of mitigating factors should be considered. These factors include:
a.
Conservatism of MSIV Testina MSIV testing during the 1994 refueling outage was conservatively performed with the valves closed after establishing a reactor pressure of 0 psig, and with no pneumatic pressure supplied to the actuators. Under design accident conditions, the valves would close with steam flow and pressure assisting to firmly seat them.
b.
Probability of Release of Radioactive Material MSIV leak tightness is required for those accidents in which radioactive material is released as the result of fuel clad damage. Redundant emergency core cooling systems (ECCS) are provided at Monticello which prevent significant fuel clad damage in event of accidents and transients. ECCS equipment was 0 9 f) 1 O n I
9510020348 950926 PDR ADOCK 05000263 g
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i USNRC NORTHERN STATES POWER COMPANY l
September 26,1995 Page 2 i
fully functional. It was unlikely that a significant accidental release of radioactive material cocid occur via the steam lines during the period in question.
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c.
Radic!auical Dose Assessment Conservatism l
I The NRC Staff willindependently assess the radiological consequences of the i
as-found MSIV leakage using methodology similar to that recently developed by i
the Boiling Water Reactor Owners Group (BWROG). While this provides a more realistic treatment of the radiological consequences of iodine over earlier techniques, many conservative " licensing" assumptions continue to be l
commonly used. Conservatisms include:
i In cases of large MSIV leakage, the actual release to the environment will be l
substantially less than the reported total leakage for all main steam lines.
Because main steam and condensate drain piping downstream of the MSIVs I
will remain intact, actual release to the environment will be govemed by the turt>ine stop valve and steam line drain flow paths. Most of the radioactive iodine will reach the main condenser where plateout and condensation will occur. Realistic leakage flow rates via the main steam lines should be used.
A large amount nf radioactive material is generally assumed to be released
{
from the fuel based on the TID-14844, March 23,1962, source term. In J
particular, the quantity and chemical form of radioactive iodine which is l
assumed greatly overestimates the radiological hazard presented by this i
nuclide. A more realistic, but still conservative, radiological" source term" i
j (NUREG-1465) was recently approved by the Commission and should be j
used in this evaluation.
l i
The reactor containment is assumed to remain at peak accident pressure, j
Pa, for the 30-day duration of the event. Leakage of radioactive materialis I
assumed to continue at a rate corresponding to the peak accident pressure.
j in reality the duration of peak pressure is brief and containment pressure j
and leakage is significantly reduced for the duration of the postulated event.
A realistic containment pressure and leakage rates should be used in this 1
evaluation.
Very conservative atmospheric dispersion models are generally used (short term worst case conditions). This often results in calculated doses ten to j
100 times greater than those which would result under more typical meteorological conditions. A realistic assumption for atmospheric dispersion j
should be used in this evaluation.
I i
l Radiological doses at the site boundary and low population zone outer boundary are calculated for 10 CFR Part 100 compliance purposes. No W2W95 NsP H3DATAWRCCoRRWASIVLET. Doc
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USNRC NORTHERN STATES POWER COMPANY September 26,1995 Page'3 credit is given for sheltering or evacuation over the assumed 30-day duration of the event. Dose calculations are therefore not representative of a realistic situation and should be carefully qualified when reported.
As explained in Licensee Event Report 94-10, several measures have been taken to improve 1
MSIV performance at the Monticello Plant. All four outboard MSIVs were replaced during the i
1994 refueling outage with improved design gate valves and actuators which are independent of instrument air for closure. In addition, a major overhaul of inboard MSIVs was completed and a safety-related instrument air supply was installed. These improvements will make a repetition of MSIV problems at Monticello unlikely in the future.
j This submittal does not contain any new commitments to the NRC, nor does it modify any existing NRC commitments. Please contact Marvin Engen, Sr Licensing Engineer, (612-295-1291), if you have any questions conceming the information we have provided.
l h / b m ) h *ft William J Hill Plant Manager Monticello Nuclear Generating Plant c:
Regional Administrator-Ill, NRC l
NRR Project Manager, NRC Resident inspector, NRC I
State of Minnesota Attn: Kris Sanda l
Attachments: 1 - Response to NRC Letter Dated 8/28/95 Request for Additional Information j
Monticello Nuclear Generating Plant Regarding Leakage of Main Steam i
isolation Valves l
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W2W96 NSP H2ATAV4RCCoRRn4SIVLET. Doc
RESPONSE TO NRC LETTER DATED 8/28/95 REQUEST FOR ADDITIONAL INFORMATION MONTICELLO NUCLEAR GENERATING PLANT REGARDING LEAKAGE OF MAIN STEAM ISOL VALVES
- 1. Provide the following information and references which will allow the staff to calculate off-site doses from increased MSIV leakage:
(a drawing may be helpful)
Data Reference a) Main steam line (MSL) inside and outside 16.12 inches (ID)
Monticello Line Designation Table M-163A pipe diameters 18.00 inches (OD) b) MSL pipe length between the MSIV and the drain line Note 1 Note 2 c) MSL pipe material Carbon Steel (A-1068)
Monticello Specification M-40 d) MSL initial operating and ambient 550 P operating Nominal full power steam temperature temperatures 120 T ambient Nominal area temperature e) Drain line inside and outside Note 1 Note 2 pipe diameters f) Drain line pipe length to the drain header Note 1 Note 2 i
g) Drain line pipe material h) Drain line header inside and outside pipe diameters I) Drain line headerlength to common l Note 1 l Note 2 drain line to condenser j) Drain line header material Carbon Steel (A-106B)
Monticello Specification M-40 k) Drain line initial (operating) and ambient 300T operating Nominal temperature tempE,rature 120 T ambient Nominal area temperature
(
I) Common drain line length to the condenser Note 1 Note 2 3
m) Condenser total volume 59,600 Ft Vendor proposal technical supplement 3
n) Condenser air space volume 44,400 Ft l
3 o) Condenser hotwell (liquid) volume 15,200 Ft p) Condenser temperature 95 T operating Saturation temperature for design pressure q) Type of material used to insulate the main steam Fiberglass Monticello Specification M-43A and drain lines r) Thickness of insulating material used on the l Note 1 l Note 2 main steam and drain lines 9/26/95 Page 1
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RESPONSE TO NRC LETTER DATED 8/28/95 REQUEST FOR ADDITIONAL INFORMATION MONTICELLO NUCLEAR GENERATING PLANT l
REGARDING LEAKAGE OF MAIN STEAM ISOL VALVES s) The MSIV leakage rate used in your calculabons and the justification for that value.
l The MSIV leakage rate used was the sum of the minimum pathway leaka0e rates for the four steemlines. Minimum i
pathway leakage rate was used since there was no failure of any of the MSIVs to close.
The leakage rates were measured at test pressures sli htly Oreater than 25 psig and conected to 3
0 equivalent leakage rates at 42 psi. This allowed all containment leakage test results to be evaluated at 42 psig (Pa).
i 0
The conelation used was formula A-3 in Appendix A to Franidin Research Center Repost TER-C5257-30 which was transmitted to Northem States Power Company with NRC letter dated June 3,1984. This transmettal was related to the NRC Staffs evaluation of Appendix J exemptions. The formula is.
m/mi = (P. + P )2, (p,)2 (P. + P,)' - (P.)*
j m/mi = ratio of mass flow of air at accident pressure to mass f!ow at test pressure P. = peak accident pressure. Gauge l
P. = test pressure, gauge Pn = one atmosphere, absolute
[
Valve Leaka0e at Test Pressure Mimimum Path Le&XaSe
(>25 psig) scfh (42 psig) scfh t
AO-2-80A 95 scfh at 25.5 psig 203
[
AO-2-86A 9284 scfh at 25.0 psig AO-2-808 6109 scfh at 25.0 psig
}
AO-2-868 2298 scfh at 25.0 psig 5067
?
AO-2-80C 10 scfh at 25.5 psig 21 AO-2-86C 88 scfh at 25.0 psig
[
AO-2-80D 114 scfh at 25.5 scfh 245 AO-2-860 173 scfh at 25.0 psig Total Reported Leakage 5536 9/26/95
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RESPONSE TO NRC LbI itR DATED 8/28/95 REQUEST FOR ADDITIONAL INFORMATION MONTICELLO NUCLEAR GENERATING PLANT REGARDING LEAKAGE OF MAIN STEAM ISOL VALVES l
- 2. Provide the following information and references which will allow the staff to calculate control room operator doses from increased MSIV leakage:
Data Reference a) lodine protection factor 3.38 - 52.9 Note 3 l
b) Control room geometry factor 37.3 Note 4 i
c) Recirculation flow rate (cfm) 0 No recirculation cleanup d) Charcoal adsorber thickness in inches Two 2-inch beds Monticello Operations Manual i
e) Adsorberefficiency(%)
90 - 99 Note 3 I
NOTES:
1)
Refer to the attached tables for information related to MSIV piping downstream of the last main steam isolation valve (MSIV) and the three drain paths for each steam line.
2)
There are four main steam lines (PS1, PS2, PS3, & PS4) and three drain paths for each steam line.
Drain paths are the main steam line drains, bypass valve manifold drain, and equalizing line drain.
P&lD drawings M-115 and M-102 show the affected piping.
Pipe lengths for each drain path were obtained from isometric drawings NF-36271. FSK-826, NX-13142-15, FSK-625, NQ-86956, NQ-86957, and NX-13142-42. Drawings are available on site for review.
Piping diameter, schedule, and insulation class were obtained from Monticello Line Desegnation Table M-163A.
Insulation thickness was obtained from Specification M-43A. These documents are available on site for review.
lodine protection factor is calculated using equation (11) from the paper " Nuclear Power Plant Control Room Ventilation 3)
System Design for Meeting General Criterial 19," K G Murphy and K M Camts USAEC,13th AEC Air Cleanin0 Conference.
Realistic charcoal iodine removal efficiency is 99% for the Monticello design. A conservative " design" value is 90%.
Filtered intake air is 1000 cfm i.10%. Unfiltered infiltration is assumed to vary from a conservative 250 cfm to 10 cfm.
9/26/95 Page 3 i
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RESPONSE TO NRC Lt:11tR DATED 8/28/95 REQUEST FOR ADDITIONAL INFORMATION MONTICELLO NUCLEAR GENERATING PLANT REGARDING LEAKAGE OF MAIN STEAM ISOL VALVES 4
NOTES (contd):
4)
Geometry factor is calculated using equation (9) of the Murphy Campe paper referenced above.
Control room volume is estimated to be 27,000 cf.
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9/26/95 Page 4
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RESPONSE TO NRC LETTER DATED 8/28/95 REQUEST FOR ADDITIONAL INFORMATION l
MONTICELLO NUCLEAR GENERATING PLANT REGARDING LEAKAGE OF MAIN STEAM ISOL VALVES MAIN STEAM LINE A - PS1 PATHWAY (NOTE 1)
LINE SEGMENT LENGTH (FEET)
PIPE ID(INCHES)
PIPE OD(INCHES)
INSULATION (INCHES)
STEAM LINE DRAIN PS1 TO D1 2.4 16.1 18.0 3.0 D1 TO D4 2.6 1.3 1.9 2.0 D4 TO PS15 15.2 1.7 2.4 2.5 PS15 TO D4 22.1 2.9 3.5 1.5 D4 TO CONDENSER 118.7 5.8 6.6
2.0 REMARhS
Orifice RO-2567 3-inch Bypass Motor Operated Valve, MO-2565 BYPASS MANIFOLD PS1 TO PS11 81.2 16.1 18.0 3.0 l
PS11 TO PS7 38.3 5.8 6.6 3.0 l
PS7 TO D26 63.0 9.6 10.8 3.0 i
l D26 TO D4 86.0 0.8 1.3 2.0 D4 TO CONDENSER 34.0 5.8 6.6 2.0 l
l REMARKS:
Orifice RO-2569 l
1-inch Bypass Motor Operated Valve, MO-1739 EQUALIZING LINE DR PS1 TO PS30 65.0 16.1 18.0 3.0 PS30 TO D40 25.2 16.1 18.0
[0_
D40 TO D4 35.5 1.3 1.9 2.0 t
D4 TO CONDENSER 18.1 5.8 6.6
2.0 REMARKS
Orfice RO-4001 1-inch Bypass Motor Operated Valve, MO-4000 l
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RESPONSE TO NRC LETTER DATED 8/28/95 REQUEST FOR ADDITIONALINFORMATION l
MONTICELLO NUCLEAR GENERATING PLANT REGARDING LEAKAGE OF MAIN STEAM ISOLATION VALVES MAIN STEAM LINE B - PS2 PATHWAY (NOTE 1)
LINE SEGMENT LENGTH (FEET)
PIPE ID(INCHES)
PIPE OD(INCHES)
INSULATION (INCHES)
STEAM LINE DRAIN PS2 TO D2 2.4 16.1 18.0 3.0 D2 TO D4 2.6 1.3 1.9 2.0 D4 TO PS15 25.2 1.7 2.4 2.5 PS15 TO D4 22.1 2.9 3.5 1.5 D4 TO CONDENSEF!
118.7 5.8 6.6
2.0 REMARKS
Orifice RO-2567 3-inch Bypass Motor Operated Valve, MO-2565 BYPASS MANIFOLD PS2 TO PS12 72.2 16.1 18.0 3.0 PS12 TO PS7 14.5 5.8 6.6 3.0 PS7 TO D26 E3.0 9.6 10.8 3.0 D26 TO D4 86.0 0.8 1.3 2.0 D4 TO CONDENSER 34.0 5.8 6.6
2.0 REMARKS
Orifice RO-2569 1-inch Bypass Motor Operated Valve, MO-1739 l
EQUALIZING LINE DR PS2 TO PS30 56.0 16.1 18.0 3.0 PS30 TO D40 28.2 16.1 18.0 3.0 D40 TO D4 35.5 1.3 1.9 2.0 D4 TO CONDENSER 18.1 5.8 6.6
2.0 REMARKS
Orfice RO-4001 1-inch Bypass Motor Operated Valve, MO-4000 l
9/26/95 Page 6
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RESPONSE TO NRC LETTER DATED 8/28/95 REQUEST FOR ADDITIONAL INFORMATION MONTICELLO NUCLEAR GENERATING PLANT REGARDING LEAKAGE OF MAIN STEAM ISOLATION VALVES MAIN STEAM LINE C - PS3 PATHWAY (NOTE 1)
LINE SEGMENT LENGTH (FEET)
PIPE ID(INCHES)
PIPE OD(INCHES)
INSULATION (INCHES) l STEAM LINE DRAIN PS3 TO D3 2.4 16.1 18.0 3.0 D3 TO D4 2.6 1.3 1.9 2.0 D4 TO PS15 35.2 1.7 2.4 2.5 PS15 TO D4 22.1 2.9 3.5 1.5 D4 TO CONDENSER 118.7 5.8 6.6
2.0 REMARKS
Orifice RO-2567 3-inch Bypass Motor Operated Valve, MO-2565 BYPASS MANIFOLD PS3 TO PS13 63.2 16.1 18.0 3.0 PS13 TO PS7 24.2 5.8 6.6 3.0 PS7 TO D26 58.4 9.6 10.8 3.0 D26 TO D4 86.0 0.8 1.3 2.0 D4 TO CONDENSER 34.0 5.8 6.6
2.0 REMARKS
Orifice RO-2569 1-inch Bypass Motor Operated Valve, MO-1739 EQUALIZING LINE DR PS3 TO PS30 47.0 16.1 18.0 3.0 PS30 TO D40 31.2 16.1 18.0 3.0 D40 TO D4 35.5 1.3 1.9 2.0 D4 TO CONDENSER 18.1 5.8 6.6
2.0 REMARKS
Office RO-4001 1-inch Bypass Motor Operated Valve, MO-4000 9/26/95 Page 7
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RESPONSE TO NRC LETTER DATED 8/28/95 REQUEST FOR ADDITIONAL INFORMATION MONTICELLO NUCLEAR GENERATING PLANT REGARDING LEAKAGE OF MAIN STEAM ISOLATION VALVES MAIN STEAM LINE D - PS4 PATHWAY (NOTE 1)
LINE SEGMENT LENGTH (FEET)
PIPE ID(INCHES)
PIPE OD(INCHES)
INSULATION (INCHES)
STEAM LINE DRAIN PS4 TO D4 2.4 16.1 18.0 3.0 D4(1-1/2 INCH) 2.6 1.3 1.9 2.0 D4 TO PS15 45.2 1.7 2.4 2.5 PS15 TO D4 22.1 2.9 3.5 1.5 D4 TO CONDENSER 118.7 5.8 6.6
2.0 REMARKS
Orifice RO-2567 3-inch Bypass Motor Operated Valve, MO-2565 BYPASS MANIFOLD PS4 TO PS14 54.2 16.1 18.0 3.0 PS14 TO PS7 31 2 5.8 6.6 3.0 PS7 TO D26 54.4 9.6 10.8 3.0 D26 TO D4 86.0 0.8 1.3 2.0 D4 TO CONDENSER 34.0 5.8 6.6
2.0 REMARKS
Orifice RO-2569 1-inch Bypass Motor Operated Valve, MO-1739 EQUALIZING LINE DR PS4 TO PS30 38.0 16.1 18.0 3.0 PS30 TO D40 34 2 16.1 18.0 3.0 D40 TO D4 35.5 1.3 1.9 2.0 D4 TO CONDENSER 18.1 5.8 6.6 2.0 REMARKS:
Orfice RO-4001 1-inch Bypass Motor Operated Valve, MO-4000 9/26/95 Page 8
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TRANSMITTAL MANIFEST NORTHERN STATES POWER COMPANY NUCLEAR LICENSING DEPARTMENT MONTICELLO NUCLEAR GENERATING PLANT NSP Letter Dated 9/26/95; AdditionalInformation Main Steam Isolation Valve Testino Manifest Date:
September 26,1995 Correspondence Date: September 26,1995 Monticello intemal Site Distribution Special Instructions Kaleen Hilsenhoff......USAR File.................Yes No_x_
St:ve Ludders.........NRC Commitment.....Yes No_x_
Lil: Imholte...............Monti OC Sec...........Yes No_x_ - 11, No dist to OC members below if YES Mel Opstad...............Monti SAC Sec.........Yes_x_ No
-5
' Mill Room..............Monti Posting..........Yes No_x_ - 7 Menticello Intemal Site Distribution:
- Monti Document Control File J C Grubb, NGSS, OC W J Hill, Plant MGR, SAC, OC L L Nolan, GSSA, OC C A Schibonski, GSE, OC B D Day, MGR MTC, OC M F Hammer, GSM, OC Operating Experience Coord J E Windschill, GSRS, OC W A Shamla, NQD Monti Site Lic File M W Onnen, GSO, OC AlWojchouski NRC Resident inspector's Office Dave Pennington Pat Tobin Dave Musolf NSP Intemal Distribution E L Watzl R O Anderson, Dir LMI, SAC T E Amundson, Dir NQD, SAC Communications Dept Yes_x__ No i
Extemal NSP Distribution
- Doc Control Desk, NRC Kris Sanda, State of Minn Regional Admin-lli,NRC J E Silberg T J Kim, NRR-PM, NRC
- Advance Distribution made by Site Licensing 4
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