ML20217K160

Provides Supplemental Info on Certain Containment Equipment, Internal Flooding & Turbine Missiles,Per 980417 Telcon

ML20217K160
Person / Time
Site:	Monticello
Issue date:	04/22/1998
From:	Hammer M NORTHERN STATES POWER CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
TAC-M96238, NUDOCS 9805010113
Download: ML20217K160 (6)

Text

l Northem States Power Company Monticello Nuclear Generating Plant 2807 West Hwy 75 Monticello, Minnesota 55362-9637 April 22,1998 US Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555 l

MONTICELLO NUCLEAR GENERATING PLANT Docket No. 50-263 License No. DPR-22 l

Submittal of AdditionalInformation Regardingh Monticello Power Rerate Program (TAC No. 96238)

A conference call was held on April 17,1998, between the NRC staff and NSP with regard to the Monticello power rerate program. NSP stated that it subsequently would provide supplemental information on certain containment equipment, internal flooding, and turbine missiles. This information is provided as Attachment 2 to this letter, j l

! Please contact Joel Beres at (612)'295-1436 if additional information is required.

l l Michael F. Hammer l Plant Manager

! Monticello Nuclear Generating Plant c: Regional Admirfistrator-Ill, NRC l

NRR Project Manager, NRC  !

Sr. Resident inspector, NRC l State of Minnesota, Attn: Kris Sanda l

J. Silberg, Esq.

Attachments i Attachment 1 NRC A.ffidavit f Attachmer,t 2 Supplemental Power Rerate Information k I q

l t p,1) I l 980501C113 980422 PDR ADOCK 05000263 p PDR

< ,a n

,# V 'wI d 'N

n UNITED STATES NUCLEAR REGULATORY COMMISSION NORTHERN STATES POWER COMPANY MONTICELLO NUCLEAR GENERATING PLANT DOCKET NO. 50-263 Submittal of Additional Information Regarding the Monticello Power Rerate Program (TAC No. 96238) i Northern States Power Company, a Minnesota corporation, by letter dated April 22,1998 provides supplementalinformation regarding the Monticello Nuclear Generating Plant power rerate program subsequent to a conference call between the NRC staff and NSP on April 17,1998. This letter contains no restricted or other defense information. ,

l NORTHERN STATES POWER COMPANY 1

By d.441A44/; , .

~"

Michael F. Hammer Plant Manager Monticello Nuclear Generating Plant On this 9B day of doY k _ d l8 before me a notary public in and for said County, personally dppeared Michael F. Hammer, Plant Manager, Monticello Nuclear Generating Plant, and being first duly sworn acknowledged that he is authorized to execute this document on behalf of Northern States Power Company, and that to the best of his knowledge, information, and belief the statements made in it are true.

t c r.

Samuel I. Shirey

"" "'"7%

Notary Public - Minnesota {^" 3 Sherburne County h >'. :nm a

!w:::ht,.:::::am...- aj ann..M...*.0N 1

l l

L

i 1 .

t .

=

Attachment 2 Supplemental Power Rerate Information 1

l l

l l

l l

j l l

i i

i l

j Environmental Qualification of Drywell Equipment  !

l A. DG O' Brien Penetrations l

The staff has expressed a concern over a particular portion of the temperature profile from j approximately 7,000 seconds to 102,000 seconds for these penetrativns. Over this time period '

of approximately 26 hours3.009259e-4 days <br />0.00722 hours <br />4.298942e-5 weeks <br />9.893e-6 months <br />, the accident profile is s 246 F, and the accident profile is about 21 F higher than the test profile. While NSP demonMrated that qualification can be established for j this portion of the profile using Arrhenius techniqi;es, the staff asked NSP to provide additional l information to supplement this conclusion. This information is provided below.

The normal bulk ambient air temperature in the drywell is less than 135 F. The DG O' Brien penetration was thermally aged prior to LOCA testing for 168 hours0.00194 days <br />0.0467 hours <br />2.777778e-4 weeks <br />6.3924e-5 months <br /> to a value of 121 C [249.8

• F). The accident profile over the area of concern is below 250 F.

The actual temperature response of the containment penetration, which is located at the containment wall, is expected to be lower than the predicted drywell air temperature response I used for the DBA LOCA EQ accident profile. This introduces an additional margin of conservatism in the accident profile. The temperature gradient applied to the penetration would be lower than the predicted drywell air space temperature gradient, and the peak would also be lower because the mass of the containment pressure boundary and the associated concrete shielding will act as a heat sink. Moreover, the DG O' Brien penetrations are uniquely positioned relative to the containment atmosphere. Each penetration is appended to an existing spare penetration. This spare penetration extends beyond the containment into the reactor building.

The portion of the penetration that is environmentally qualified, the connector, is actually located in the reactor building. See simplified diagram below.

I REACTOR BLDG CONTAINMENT Connector - ,,,, - ,,,,,,

\ y Cable E Spare j Penetration DG O' Brien ./

Penetration in addition to the conservatisms identified above, the drywell air temperature response is calculated using simultaneous application of worst case conditions including an initial power assumption of 1917 MWt which is 8% above the requested rerate power level of 1775 MWt.

The analytical containment temperature response profile provides a conservative bounding prediction of worst case drywell temperatures with respect to the actual post-LOCA response.

There are two DG O' Brien penetrations in use at Mcnticeijo. These penetrations contain the instrument cables for the Ger,eral Atomic radiation monitors. As discussed in part B below, a 2

j - .

'ailure of the radiation monitors would not affect plant safety. The radiation monitor instrument

, cables are used for low signal strength applications, and the current levels are very low. A failure of these cables, however unlikely, would not involve a damaging arc. Given the rerate accident profile, the penetrations are unlikely to fail in a manner that would compromise

, containment integrity. Both the inner and outer pressure boundaries of the penetration would l have to be compromised.

B. General Atomic Radiation Monitors NSP obtained an exemption from the requirements of IEEE 323-1974 for these radiation monitors. By letter from L.O. Mayer, NSP, to Director, Nuclear Reactor Regulation, " Post TMI Requirements - NUREG - 0737," dated December 30,1980, NSP requested an exemption to qualify the radiation monitors to IEEE 323-1971. The staff approved this exemption by letter dated June 3,1982. Section 5.4 of the 1971 IEEE 323 Standard, of itself, does not preclude a qualification and profile analysis of the type previously submitted to the staff for power rerate. 3 l Section 2.4, Other Qualification Methods, of NUREG-0588, " Interim Staff Position on l Environmental Qualification of Safety-Related Electrical Equipment," provides a comparison of the margin requirements for the 1971 and 1974 standards. This section allows for a qualification method using margins other than those proposed in Section 6.3.1.5 of IEEE 323-1974.

l The radiation monitors are not original plant equipment. The monitors were installed to meet the i post-accident monitoring requirements of Section ll.F.1.3 of NUREG-0737 and the guidelines of i Revision 2 of Regulatory Guide 1.97. This equipment is not required to function to mitigate any j design basis accident, and the failure of these' monitors in an accident environment would not affect plant safety.

The equipment can be used to detect increases in radiation levels above background during normal plant operations. Slight changes in the post-accident LOCA profile do not affect this capability. The instrument is also used to determine gross estimates of fission product I concentrations in the containment after a low-probability severe accident involving core damage. j Cther equally viable substitute methods exist, such as containment sampling and software models, to estimate these concentrations.

For the area of the accident profile that the staff has identified, the temperature gradient is not limiting, and a failure of metal components is not indicated. The radiation monitor is constructed almost entirely of metal parts. The most susceptible component of the detector assembly is the 4 non-metallic Raychem cable splice. The splice material was separately te sted, independent of the General Atomic qualification testing, to a longer and higher profile that bounds the post-LOCA containment temperature profile. The splice was also thermally aged to the following values in the Raychem test program.

7318 hrs @ 136 C [276.8 F]

830 hrs @ 162 C [323.6 F]

3

. . . . . i

• I Turbine Missiles and internal Flooding )

A. Turbine Missiles  ;

Tho High Pressure (HP) turbine modifications at Monticello established the necessary rerate steam flow path. The HP turbine modifications replaced the HP turbine rotor and diaphragms.

The Low Pressure (LP) turbine modifications replaced the LP turbine rotors, inner casings, diaphragms, shaft packing boxes and steam guides. This work was completed during the 1996 refueling outage. The LP turbine modifications were not a necessary part of the rerate program, and these modifications would have been accomp!ished irrespective of power level. During the 1998 refuel outage, the first 3 rows of stationary diaphragms and the first row of buckets on the HP rotor were replaced to assure adequate steam flow capacity for rerate operation.

Power rerate does not involve an increase in the probability or consequences of a turbine missile event. A missile analysis was conducted for the replacement of the original stacked rotors with monoblock rotors. This analysis is presented in Section 12.2.3 of the Monticello USAR. Power rerate does not change the conclusions of this section. The limiting component is the LP tb.aine L-0 buckets, and the turbine modifications for power rerate. which are confined to the less susceptible HP turbine, do not affect this component. Under rerate conditions, the j turbine speed remains at 1800 rpm, and steam quality will increase which will reduce the erosion rate of turbine components. The material strength of the LP buckets and rotors are unaffected by power rerate. In addition, power rerate does not involve any changes to the turbine overspeed protection and, consistent with present turbine operation, power rerate does i not involve sustained operation at critical turbine frequencies.

B. Intemal Flooding l Operation at power rerate does not affect the bases for intemal flooding at Monticello. Under  !

present and power rerate conditions, no single non-Class I equipment failure with the resultant flooding thereof will prevent a safe shutdown or result in a common mode failure of redundant  ;

~

engineered safety systems.

Power rerate does not involve increases in the probability or consequences of an intemal flooding event at MN3P. There are no configuration changes to water collection areas, flood barriers, or flood mitigation equipment. Under rerate operating conditions, only the reactor feedwater and condensate systems experience significant process flow increases. There are no significant changes to the capacity or flow characteristics of any other non-Class I systems that  !

could serve as a potentialintemal flooding source. The analysis of the flooding effects of

. feedwater and condensate line breaks were evaluated at rerate operating conditions using GOTHIC .' This analysis demonstrated that these breaks do not result in flooding levels that would prevent a safe shutdown or cause the loss of redundant engineered safety systems.

4