ML20247K336

Responds to 980424 RAI to Complete Review of License Amend Request Re Rerate Program

ML20247K336
Person / Time
Site:	Monticello
Issue date:	05/12/1998
From:	Hammer M NORTHERN STATES POWER CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
TAC-M96238, NUDOCS 9805220071
Download: ML20247K336 (10)

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I ik,,it.e,T. States Power Company

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• Monticello Nuclear Generating Plant l 2807 West Hwy 75 Monticello, Minnesota 55362-9637 i

I May 12,1998 i

US Nuclear Regulatory Commission

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Attn: Document Control Desk j Washington, DC 20555 MONTICELLO NUCLEAR GENERATING PLANT Docket No. 50-263 License No. DPR-22 Response to April 24,1998 Request for Additional Information (RAI) on Monticello Power Rerate License Amendment (TAC No. M96238)

Ref.1 Letter from NRC to R.O. Anderson, NSP, "Monticello Nuclear Generating Plant - Request for Additional Information on License Amendment Request Entitled ' Supporting the Monticello Nuclear Generating Plant (MNGP) Power Rerate Program' (TAC No.

M96238)," April 24,1998 By letter dated April 24,1998 (Ref.1), the NRC staff provided a Request for Additional Information (RAl) to complete its review of NSP's license amendment request for the Monticello Nuclear Generating Plant (MNGP) Power Rerate Program. NSP's response to Ref.1 is provided as Attachment 2 to this letter.

Please contact Joel Beres at (612) 295-106 if additional information is required.

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Michael F. Hammer -

Plant Manager Monticello Nuclear Generating Plant

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c: Regional Administrator - 111, NRC NRR Project Manager, NRC Sr, Resident inspector, NRC State of Minnesota, Attn: Kris Sanda J. Silberg, Esq.

Attachments Attachment 1 Affidavit to it.e US Nuclear Regulatory Commission Attachment 2 NSP Response to Staff Power Rerate RAI Dated April 24,1998 i _;

UNITED STATES NUCLEAR REGULATORY COMMISSION NORTHERN STATES POWER COMPANY

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l MONTICELLO NUCLEAR GENERATING PLANT DOCKET NO. 50-263 {

Response to April 24,1998 Request for Additional Information (RAI) on Monticello Power Rerate License Amendment (TAC No. M96238)

Northern States Power Company, a Minnesota corporation, by letter dated May 12,1998 i provides its response for the Monticello Nuclear Generating Plant to a US Nuclear Regulatory Commission (NRC) letter dated April 24,1998, with the subject "Monticello Nuclear Generating Plant - Request for Additional Information on License Amendment Request Entitled ' Supporting the Mont! cello Nuclear Generating Plant (MNGP) Power Rerate Program' (TAC No. M96238)." ,

This letter contains no restricted or other defense information.

f NORTHERN STATES POWER COMPANY By h A1/l44LL/ .

Michael'F. Hammer Plant Manager Monticello Nuclear Generaang Plant A

On this IN ay d of I\ Axe \D8 before me a notary public in and for said County, personally appeared 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.

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. SamuelI. Shire ,

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Notary Public - Minnesota Sherburne County Q)"

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Attachment 2 NSP Responss to Staff Power Rerate RAI Dated April 24,1998 t

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1. NSP concluded that based on its review of scram irip setpoints (and their changes) po.werrerate showed no significant effect on trip (initiating event) frequency. Provide a basis / discussion for this conclusion.

NSP Response l

The GE setpoint methodology used at Monticello include's a spurious trip avoidance calculation. The calculation determines whether the setpoint value is sufficiently removed from a trip given allowances for instrumentation uncertainties such as drift.

This trip avoidance calculation was completed for all scram setpoints affected by power rerate, and the calculations demonstrated that the appropriate acceptance criteria for trip avoidance was met.

NSP also reviewed several events arising from equipment failures and operational maneuvers that involve integrated plant responses in regard to scram trip avoidance.

These events involve equipment setpoints in systems that could eventually result in an inadvertent scram. For example, reactor water high level following a single recirculation pump trip would trip the reactor feed pumps and cause a reactor water low level scram.

For these situations, NSP conducted an evaluation using the core transient models used for cycle analysis. This evaluation demonstrated that the setpoints in question had sufficient trip avoidance margin at rerate conditions.

The PRA team relied on these evaluations in reviewing the impact on the potential for initiating events.

Exhibit D of the rerate license amendment dated December 4,1997 contains a listing of hardware changes and setpoint modifications which support NSP's conclusion that there is no significant potential for increases in the frequency of initiating events modeled in the PRA. For example, the following changes were made to assure reliability of balance of plant systems.

Turbine Trip increase capacity of moisture separator drain valves increase generator isophase bus cooling and stator water cooling capacities Loss of Feedwater - Replace condensate pumps with higher capacity pumps Loss of Condenser Raise condenser low vacuum setpoint These modifications are intended to maintain margin on trip setpoints at rerate power levels and to maintain equipment operation within existing design constraints. As a result, it was concluded that the potential for these transient initiators was no greater due to the rerate.

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it was also concluded that none of the remaining initiating events modeled in the PRA are affected by the rerate. The following are examples.

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LOCAs Frequency dictated by the potential for passive pipe failures, which are independent of power level.

Transient initiators . Transient initiators other than those noted above are I

support systems, which do not directly interface with the primary system and would not be affected by changes in power.

DC power AC power l Service water RBCCW Instrument Air Loss of Offsite Power Dictated by the reliability of the grid and the switchyard which are not degraded by the rerate.

ATWS Rod insertion takes place under the same primary system conditions as existed prior to the rerate.

Combined with the conclusion that there is no change in the transient initiator frequency, the potential for ATWS is also unchanged.

Intemal Flooding Dictated by passive piping system failures and the frequency of maintenance activities which might lead to flooding conditions, neither of which is affected by the rerate.

Intemal Fires Dictated by combustible loading within each fire area and not by the rerate.

2. NSP indicated that the likelihood of all eight SRVs [ safety relief valves] failing is not significantly different from the likelihood of at least seven out of eight SRVs failing.

l Provide a besiddiscussion for this conclusion.

NSP Response l ' If only independent valve failures were considered, there would be a difference between the probabilities of seven failures and eight failures. However, the probability of seven and eight valves failing is dominated by common cause. Given that seven valves have

, failed due to common causes, it is highly likely that the eighth is similarly affected and will fail from the same causes. Therefore, the common cause failure probability for eight of l eight failures is virtually the same as for at least seven of eight failures, and the overall probability of eight failures is virtually the same as for at least seven failures.

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3. NSP concluded that the failure rates of equipment / component will not change due to power relate. Provide a basis / discussion for this conclusion.

NSP Response NSP conducted and documented comprehensive reviews of all plant systems and associated equipment that had the potential to be affected by power rerate. These reviews were conducted by cognizant system engineers and by subject matter experts.

Each system evaluation received secondary reviews by technical experts and management review teams as described in NSP's response to Question 19 of our letter to the NRC dated March 26,1998.

Changes in equipment service conditions and process changes, if any, were identified.

For the great majority of plant equipment, power rerate does not result in any change to the system processes or service conditions that would involve operation outside of the design constraints that were used to establish the present capabilities of the equipment, such as code allowabies, temperature, pressure, cooling requirements, run times, and electrical loading. In a few cases, these reviews identified modifications which would improve the performance of certain equipment and systems under rerate conditions.

These modifications are included in Exhibit D of our letter to the NRC dated December 4, 1997. Most of these modifications are complete, and all will be completed prior to rerate operation. The F~tA team relied extensively on these deterministic reviews to conclude that plant systems and equipment would continue to be operated within the existing design constraints.

A number of accident sequences would generate conditions which go beyond normal design. Moreover, the failure probability of some components depends on the number of demands the component may receive during the accident. Therefore, in addition to the system reviews described above, the most risk significant components from the PRA were reviewed to determine whether any components were expected to operate in accident sequences that went beyond their design basis or whether the number of demands would change.

To perform the review, all components whose failure could potentially contribute to accident sequences on the order of 1E-6/yr were identified. This threshold for risk significance corresponded to a risk achievement worth (RAW) of 1.07. The following summarizes the results of the review.

Level 1 PRA considerations: In loss of decay heat removal sequences (Accident i Class ll), the containment is allowed to approach its design pressure before containment venting is initiated. The wetwell and drywell vent and purge valves are credited in the PRA as one means of venting. These valve operators are capable of actuating at the containment design pressure. As rerate has no effect on this pressure, the vent valve reliability was rat changed in the rerate accident sequence quantification.

Level 2 PRA considerations: Post Core Damage Environment. A number of i plant components are expected to continue operating after core damage and are credited in the Level 2 PRA even when exposed to the environment associated 3

, with a core melt. This equipment includes the SRVs inside containment and the injection systems located in the reactor building which may be exposed to shine

. from the containment or from coolant recirculated through piping systems or to high temperatures and pressures following failure of the drywell.

A general assumption was made throughout the PRA regarding environmental qualification. Any equipment which was exposed to an environment that exceeded its qualification limits was assumed to be unavailable. For equipment operating within its qualification temperatures, pressures, and radiation levels, random failure probabilities were used. All environmentally qualified equipment will also be qualified for rerate conditions. As a result, no changes were made to the failure probabilities for equipment modeled in the Level 2 PRA.

Changes in the Number of Demands due to Rerate: A higher decay heat levelis expected to slightly increase the number of demands on the SRVs used for pressure relief when the primary coolant system is isolated, thus increasing the potential for a stuck open relief valve during such sequences. It was found that the risk of core damage due to transient initiators followed by a stuck open relief valve is just under 1E-6/yr. The risk of stuck open relief valves is dominated by station blackout sequences because the gradual blowdown of the reactor to the suppression pool will eventually result in a trip of the turbine-driven HPCI and RCIC pumps, leaving only the diesel-driven fire pump as a source of makeup.

Other transient initiators do not contribute significantly to sequences with stuck open relief valves because in those cases the low pressure systems and the CRD pump are available for coolant makeup once the blowdown is complete.

Station blackout sequences were analyzed to determine the effect of rerate with and without a stuck open safety relief valve. The analyses produced the following insights.

The number of demands on the SRVs does increase, but only slightly. Over the first four hours of a station blackout, the rerate case produced 16 demands on the SRV with the lowest setpoint as compared to 15 demands assuming 1670 MWt power. The change is slight because it is offset by the turbine-driven HPCI or RCIC pumps which draw more steam at the higher power level since they must supply more makeup water to the reactor than at the present powerlevel.

if a relief valve does stick open, the time until the turbine-driven pumps trip off due to low pressure is slightly longer for the rerate case. This is because the higher decay heat level produces more steam, which keeps the reactor above the pump's low pressure trip setpoint longer. Analysis of the time available until the water level drops to the top of active fuel at rerate shows an increase of about 5% for rerate over the 1670 MWt case (2.1 hours1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> versus 2.0 hours0 days <br />0 hours <br />0 weeks <br />0 months <br />).

The smallincrease in the time available to restore a power source or align a diesel-driven fire pump results in a slight improvement in the CDF. The small increase in the chance of a stuck open relief valve because of the increase in the 4

number of SRV demands from 15 to 16 (7% increase) was evaluated, ar 4 the change in the overall CDF was approximately 1 E-8/yr.

The dete'ministic r reviews confirmed that the plant equipment would continue to be operated as designed following the rerate, and the PRA review found no component expected to operate outside its design following an accident. In addition, a search for a meaningful increase in demands on the most risk-significant components identified only the SRV item discussed above. As a result, there was no need to modify component failure rates used in the PRA for rerate conditions.

4. On page 37 of the September 5,1997, submittal, NSP reported that from the baseline Level 2 PRA [probabilistic risk assessment], the potential for a large early release is on the order of 3% of the total CDF [ core damage frequency] and that this frequency remains at about 3% for the rerate CDF. - The large eady release frequency reported in the Monticello IPE[individualplant evaluation] was 4.15E-6/Yr, which represented about 16% ofits estimated CDF (2.6E-5tYr). Provide a basisidiscussion for the difference between LERF [large eady release frequency] estimated for IPE and LERF estimated for baseline and rerate analysis. ,

NSP Response The Monticello IPE submittal does not report a value for LERF. Figure 1.4-5 of the IPE does show the total early releases as being 16% of the total core damage frequency.

Most of these releases, however, are small because they are scrubbed through the suppression pool. The actual large early release frequency for the Monticello IPE would be 2.0E-6/yr, or about 7.4% of the total core damage frequency.

The LERF reported in our response to RAI Question #48 is still less than the IPE value of 2.0E-6/yr. In performing the most recent update for the Monticello PRA, changes to the classification of ATWS sequences were made. In the original IPE, two of the dominant accident sequences were conservatively classified as unscrubbed, when in fact reled:;cc would be scrubbed through the suppression pool. See sequences 9 and 10 of Figure 4.5-6 of the IPE submittal which total 1.5E-6/yr. In these sequences, high and low pressure injection arrest the core melt progression within the vessel once the core has become subcritical. Because the vessel is intact, all releases would be through the SRV tailpipes into the suppression pool before being released through the failed drywell. J

5. Page 10-16 of the December 4,1997, submittal stated in part: 'a sensitivity study was performed for tomado missiles which resulted in a negligible increase in risk." Provide a discussion on what this study entails and the basis for concluding that its effect on risk is

-negligible.

NSP Response The Other Extemal Events analysis for the Monticello IPEEE was largely performed by comparing the plant design with the 1975 Standard Review Plan. The exception was tomado missiles. Safety related structures at Monticello provide protection against missiles with the exception of a few doors and ventilation paths into buildings such as the reactor building. The IPEEE evaluated the potential for tornado missiles entering 5

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1 through these penetrations and quantified the expected core damage frequency assuming the failure of all the equipment in the rooms with these penetrations. The core da. mage frequency associated with the tomado missiles wts shown to be below 1E-7/yr which is well below the screening criteria presented in NUREG-1407.

The tornado missile evaluation was repeated for the rerate analysis. As expected, rerate does not affect the tomado frequency or the potential for missiles striking doors or j

ventilation paths into safety-related structures. However, as in the internal events and fire analyses, the timing of the accident sequences is shorter due to the higher decay heat levels. This has a small effect on the time available for the operator actions credited l

in the tomado missile accident sequences such as emergency depressurization and initiation of decay heat removal systems. The change in core damage frequency due to rerate in the tomado missile analysis remained below 1E-7/yr.

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