ML20199K784
| ML20199K784 | |
| Person / Time | |
|---|---|
| Site: | Monticello  |
| Issue date: | 11/18/1997 |
| From: | Hammer M NORTHERN STATES POWER CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| NUDOCS 9712010201 | |
| Download: ML20199K784 (8) | |
Text
y Northem States Power Company Morticello Nuclear Generating Plant 2007 West Hwy 75 Monticetto, Minnesota $5362 9637 November 18,1997 US Nuclear Regulatory Commission NUREG-0737 Attn: Document Control Dosk Supplement 1 Washington, DC 20555 MONTICELLO NUCLEAR GENERATING PLANT Dockot No. 50-263 License No. DPR 22 Roquest for Deviation From Emergoney Procedure Guidelines. Revision 4. NEDO-31331. March 1987 The purpose of this letter is to request Nuclear Regulatory Commission (NRC) review and approval of a deviation from the Boiling Water Roactor (BWR) Owners' Group Emergency Proceduto Guidelines (EPGs), Revision 4, NEDO 31331, March 1987. Revision 4 of the EPGs was approvod by the NRC Staff in a Safety Evaluation Report (SER) issued on September 12, 1988. This dovlation is necessary to permit the Monticello Emergency Operating Procedures (EOPs) to recognizo 2/3 coro holght as adoquato for core cooling following a largo break loss of coolant accident.
BackaroWDd in January 1997, the Nuclear Regulatory Commission (NRC) completed a System Operational Perfoimance inspection (SOPI) of the Monticello Residual Heat Removal (RHR) system (Reforenco 1). During this inspection, unroviewed satety questions associated with the containment long term cooling analysis and the not positivo suction head (NPSH) requirements of the RHR and core spray pumps were identified.
The unroviewed safoty questions were resolved by the NRC with the issuanco of Amendment No. 98 to the Monticello full term operating license (Reference 2). One of the conditions imposed by the NRC with this licenso amendment was the following requiremor t related to reactor vossol water levol for adequate cora cooling following a loss of cocknt accident (1.00' :
g Process a 10 CFR 50.59 evaluation to change the EOP definition of adequate core cooling to 2/3 coro height. The corresponding EOP chan0es and the required operator training shall also be complotod. Final implomontatici.s shall be completed when r0 the 1") CFR l
i 50.59 evaluation requirem9nts are satisfied.
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USNRC; Northern States Power Company
' Novi:nber 18,1997 Page 2 The Monticello EOPs conform to Revision 4 of the NRC approved EPGs If reactor water level following a LOCA cannot be restored to above the top of the active fuel (TAF), the EPGs require the drywell to be flooded and the reactor vessei vented to raise the water level in the vessel to above TAF. This is inconsistent with the licensing basis analysis in the Monticello Updated Safety Analysis Report (USAR). The USAR recognizes 2/3 core height as being adequate for core cooling. After water levelin the reactor is restored to 2/3 core height by the emergency core cooling system (ECCS) following a LOCA, the USAR assumes one or more RhA pumps and RHR service water pumps are used for long term cooling without the need for containment flooding.
Tho differences between the plant licensing basis and the Revision 4 of the LPGs have been reviewed in detail. This included:
Review of the Monticello licensing basis
. Review of current regulatory requi.ements for emergency core cooling Review of the basis foc current EOP requirements Monticello Licensino Basis The design and licensing basis for Monticello emergency core cooling system performance in the event of a large break LOCA is described in the following sections of the USAR:
Cora Spray System Section 6.2.2.3 Low Pressure Core Section 6.2.3.2.2 injection (LPCI)
The USAR discussion of core spray system response to a large break LOCA ir derived from the original FSAR. Following the LOCA, it is stated that the system rapidly restores water level inside the shroud to the top of the jet pumps (2/3 core helght). The subcooled water leg in the jet pumps provides a differential pressure which supports a higher swollen level inside the shroud, effectively cooling the core.
The USAR discussion of LPCI core flooding in response to a LOCA is also derived from the original FSAR. Following the LOCA, LPCI flow rapidly accumulates in the lower plenum of the reactor vessel, and level inside tne shroud increases until the level reaches the top of the jet
, pumps. The cooler fluid in the jet pumps supporta a higher two phase level inside the shroud.
The core is effectively reflooded, and subsequent LPCI Injection replaces water lost by steam generation in the core.
Once the core has been reflooded, the USAR then describes realignment of one RHR pump and starting of one or two RHR service water pumps for long term contairunent cooling.
Therefore, as described in the USAR, either the core spray or LPCI system will provide adequate core cooling with the core covered to 2/3 core height.
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USNRC Northern Statea Power Company
' November 18,1997 Page 3 It should be no'ed that 2/3 core height refers to " collapsed" subcooled level as measured by the fuel.?one water level instrumentation. Actual " swollen" water level inside the core shroud can be substantially higher than measured level.
Current Reaulatorv Reauiremento For Emeroency Core Cooling The acceptance criteria for ECCS performance is contained in 10 CFR Part 50, Section 50.46(b). Criterion (b)(5) contains the acceptance criterion for long term cooling:
After any calculated successfulInitial operation of the ECCS, the calculated core temperature shall be maintained at an acceptably low value and decay heat shall be removed for the extended period of time required by the long-lived radioactivity remaining in the core.
As noted in the Monticello EAFER/GESTER LOCA Analysis Basis Documentation (Heference 3), documentation of compliance with Criterion (b)(5) was performed generically for all bolling water reactors (BWRs) by General Electric in Volume 2 of NEDO 20566A (Reference 4).
As noted i'n Volume 2 of NEDO 20566A, long-term cooling is assured for:
Pioe Breaks Other Than in the Recirculation System. The reactor vessel refloods for all pipe breaks other than in the recirculation system, and the fuel cladding quickly cools to sat" ration temperdure. No further perforation nor metal-water reaction will result.
Recirculation Line Baaks in Non-Jet Pumo BWRs. As demonstrated in Section 1,.
the fuel rods will be wetted by the core spray in a matter of minutes following the accident. The cladding surface area will quickly reium to saturation temperature. No further perforations nor metal-water reaction will result.
Recirculation Line Breaks in Jet Pumo BWRs. When the core refloods shortly following the postulated LOCA, the fuel rods will retum quickly to saturation temperature over their entire length. For large pipe breaks, the heat flux in the core will eventually be inadequate to maintain a two-phase water flow over the entire length of the core since the static water level inside the core shroud is approximately that of the jet pump suctions. When at least one spray system is available long term, the upper third of the core will remain wetted by the core spray water as in non-Jet pump BWRs, and there will be no further perforation or metal-water reaction.
Recirculation Line Break in Jet Pumo BWRs With LPCI Inlection into Recirculation Piping. Even if a core spray system is not available long term, for those plants with LPCI injection into the recirculation piping, the upper region of the core will be cooled by convection to the steam generated in the still-covered region. Cladding temperatures will not reach values resulting in further perforation, significant additiontl oxidation, or
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OSNRC Northern States Power Company November 18,1997 Page 4 significant additional metal water reaction. This level swell cooling mechanirm has been investigated experimentally and analytically over a wide range of conditions.
Table 5 5 of Reference 3 cummarized the ECCS systems available for alllimiting break locations and limiting ECCS single failures. At least one core spray system will be available, except for a break in one of the two core spray lines. A core spray lins break is not a concem since the core spray vessel penetrations are located well above TAF and thus the core will remain covered for this event.
At least one LPCI system is available except for a recirculation line break with f ailure of the LPCI injection valve. In this case, two core spray systems will be available. Water levelin the reactor sessel will be maintained at 2/3 core height and long term cooling of the entire core will be assurod.
- Basis For Current EOP neauirements Symptomatic Emergency Procedure Guidelines (EPGs) were developed by the BWR Owners' Group following the accident at Three Mile Island (TMI) as required by the NRC TMI Action Plan, Revision 4 of the EPGs (Reference 5) serve cs the basis of the current Monticello EOPs contained in Section C.5 of the Operations Manual. Revision 4 of the EPGs has been reviewed and approved by the NRC (Reference 6) Refer to Section 13.7 of the Monticello USAR.
The most recent vorsion of the EPGs, the Emergency Procedures and Severe Accident Guidelines (EPG/ SAGS), are currently being implemer.ted by BWRs. They will be implemented at Monticolio by the end of 1998.
Revision 4 of the EPGs requires that reactor water level following a LOCA be restored to abovo TAF before adequato coro cooling is assumed to be achioved. If level cannot be restored to TAF, the containment is immediately flooded using all available sources of water. The reactor vossolis vented to allow the coro to be floodod through the break. The EPG/ SAGS differ slightly, and are loss conservative than the EPGs, in that they specify reactor water level must be rostored only to the minimum steam cool'ng reactor water level (MSCRWL) to avoid containment flooding. For Monticello, those reactor water levels are:
Lgyrl Inches
- 7AF 126 MSCRWL
-150 2/3 Core Height
-174
- Vessel rero is 477.5 inches from inner clad on the bottom of vessel The conflict between the EPGs and the UGAR was evaluated when Revision 4 of the EPGs was adopted at Monticello. The conflict w.A found to be acceptable because reactor water level above TAF was viewed as a more conservative requirement than 2/3 core height. As the result of investigations prompted by the recent NRC SOPl inspection, it is now apparent that
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USNRC Northern States Power Company Novem6er 18,1997 Page 5 requiring reactor water level to be above TAF for adequate core cooling may be inappropriato.
In the event of a large break LOCA, RHR and RHR service water pumps would not be aligned for long term containment cooling as described in the USAR. Instead, operators would be dlrected to begin flooding the containment.
In discussions with General Electric and NSP personnel involved in developing the EPGs, three reasons were identified for specifying TAF instead of 2/3 core height for long term heat removal:
1.
During development of the EPGs, there were concerns about the adequacy of core spray cooling, it was known that the spray pattem of BWR 4 and BWR 5 core spray nozzles narrowed in a steam environment. The spray distribution to certain assemblies was found to be less than the design value whenever only one core spray pump was available. BWR 3 and BWR-E plants, including Monticello, use spray nozzles of a different design. Spray nozzles in these p' ants are not significantly affected by a stearit environment (Reference 7).
2.
Flooding of containment and the reactor vessel to above TAF places the reactor into a stable condition for long-term cooling. Reliance on pumps and other active equipment to maintain this condition is minimized. Conditions are stable and required operator reaction times are long.
3.
Following a LOCA, leve! will be quickly restored abeve TAF, and the core will remain covered for long term cooling except for the largest breaks in the recirculation piping. A break of this size will require containment to be flooded anyway for accident recovery.
8 Resolution of EOP and USAR Conflict The EOPs place a high priority on ensuring the core is covered with water following any accident.
In the event of a large break in the recirculation sptom, water levelin the vessel cannot be restored to above TAF. The EOPs now direct the operators to begin flooding containment and the reactor vessel to above TAF within the first few minutes of initiation of the accident. This creates several conflicts with the licensing basis plant response described in the Monticello USAR:
By adhering to the EOPs, operators will not align RHR and RHR service water pumps for long-term suppression pool cooling when reactor level is restored to 2/3 core height as assumed in the USAR. The containment would be flooded, g
USAR environmental qualifcation, shlolding, and radiological analyses may no a
longer be applicable because conditions are different than those assumed.
Pooding of containment and the m clor vessel requires venting the reactor to the condenser (and ultimately to the environment) in a manner not considered in the USAR mdiological analyses.
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USNRC Northern States Power Company
' Noviem, er 18,1997 b
Page 6 We believe the current EOPs are to) conservative in not crediting core spray or LPCl injection for providing adequate core cooling with water level at 2/3 core height. In the event of a large break LOCA, operators would be directed to begin flooding containment and venting the reactor vessel within minutes. With core spray or LPCI operable, it is not necessary to begin flooding containment immediately.
In many situations the decision to flood containment may represent the best option available for cooling the core. This decision, however, can be made following a thorough analysis of the particular plant conditions and consideration of all alternatives.
Reauested NRC Action Based on the above considerations, it is requested that the NRC approve a deviation from
- Revision 4 of the BWROG EPGs and the new EOP/ SAGS. This deviation will permit a change to l
the Monticello EOPs to recognize 2/3 core height as providing adequate core cooling.
Following approval of this request, the following actions will be taken to resolve the conflict described above between the EOPs and the USAR:
Revise the EOPs to change the reactor vessel water level necessary for adequate cora e
cooling from TAF to 2/3 core height. The revised EOPs will direct operators to enter the drywell flooding procedure only if they cannot restore and reliably maintain reactor vessel water level at or above 2/3 core height Containment and reactor vessel flooding will remain an option in the EOPs for accidents which are beyond the licensing basis of the plant, for long term accident recovery, or when judged to be prudent by the emergency response organization, Present necessary tra'ning to all personnel, e
Revise the Monticello lJSAR to describe the NRC approved deviation from Revision 4 of e
the EPGs and the EOP/ SAGS.
This submittal contains no new NRC commitments, nor does it modify any prior commitments.
Please contact Mr Joel Bcres, Licensing Engineer, at (612) 2951436 if you require additional information related to this request.
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Michael F Hammer Plent Manager Monticello Nuclear Generating Plant I
- USNRC, Northern States Power Company Nov* ember 18,1997 Page 7 c:
Regional Administrator-Ill, NRC NRR Project Manager, NRC Resident inspector, NRC State of Minnesota Attn: Kris Sanda J Silberg, Esq.
Attachment:
Affidavit to the US Nuclear Regulatory Commission REFERENCES 1.
NRC Syi, tem Operational Performance inspection (SOPI) Report 50-263/96009(DRS), February 20,1967 2.
Monticello Nuclear Generating Plant losuance of Amendment Re: Updated Analysis of DBA Containment Temperature and Pressure Response and Reliance on Contelnment Pressure to Compensate for Poteritlal Deficiency in NPSH for ECCS Pumps During DBA (TAC No, M97781), July 25,1997 3.
GE'NE 187 02 0392, Revision 1, *Monticello Nuclear Generating Plant SAFER /GESTR LOCA Analysis Basis Documeritation," July,1993 4.
NEDO 20566A," General Electric Company Analytical Model for Loss-Of Coolant Analysis in accordance with 10CFR50 Appendix K Volume t" Septemt,ar 1986 5.
NEDO 31331," Emergency Procedure Guidelines, Revision 4," March,1987 6,
NRC Letter from Ashok C Thadanl to Donald Grace, Chairman BWROO,
- Safety Evaluation of BWR Owners Group - Emergency Procedure Guidelines, Revision 4. NEDO 31331, March,1987," dated September 12,1988 7.
NEDO-20566 3," General Electric Company Analytical Model for Loss-Of-Coolant Analysis in t.ccordanco with 10CFR50 Appendix K Amendment 3 -
Effect of Steam Environment on BWR Core Spray Distribution," April,1977
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UNITED STATES b'UCLEAR REGULATORY COMMISSION NORTHERN STATES POWER COMPANY MONTICELLO NUCLEAR GENERATING PLANT DOCKET NO. 50 263 Roquest for Deviatior; From Emergency Procedure Guidelines. Revision 4. NEDO-31331. March 1997 Northern States Power Company, a Minnesota corporation, requests approval of a dov3stion g
from the Boiling Water Reactor Owners' Group Emergency Procedure Guidelines, Revision 4, NEDO-31331, March 1987, which woro approved by the NRC Staff.
This lottor contains no restricted or other detonso information.
NORTHERN STATES POWER COMPANY By N Ilf M1tW Michao F Hammer Plant Manager Montcollo Nuclear Generating Plant On this I ay of Nedwr, \\%7 before me a notary public in and far said County, personally appeared Michael F Hammer, Plant Manager, Monticello Nuclear Generating Plant, and being first duly sworn acknowledgod that he is authorized to execute this document on behalf of Northern States Power Company, that he knows the content:, thoroof, and that to the best of his knowledge, information, and belief the staternents mado in it are true and that it is not interposed for delay.
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N u ng Samuel i Shiroy SAMUEL l. SMiMY Notary Public-Annnesota mmm rusue.mmatesta Sherburne County My Commission Expiros January 31,2000
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