ML20128M710
| ML20128M710 | |
| Person / Time | |
|---|---|
| Site: | Monticello  |
| Issue date: | 02/12/1993 |
| From: | Parker T NORTHERN STATES POWER CO. |
| To: | |
| Shared Package | |
| ML20128M699 | List: |
| References | |
| NUDOCS 9302220253 | |
| Download: ML20128M710 (9) | |
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UNITED STATES NUCLEAR REGULATORY COMMISSION NORTHERN STATES POWER COMPANY MONTICELLO NUCLEAR GENERATING PiANT DOCKET NO. 50 263 REQUEST FOR AMENDMENT TO OPERATING LICENSE DPR 22 LICEME AMENDMENT REQUEST DATED February 12, 1993 Northern States Powe Company, a Minnesota corporation, requests authorization for changes to Appendix A of the Monticello Operating License as. shown on the.
attachments labeled Exhibits A, B, and C.
Exhibit A describes the proposed changes, describes the. reasons for the changes, and contains a Safety Evaluation, a Determination of Significant llaeards Consideration and an Environmental Assessment.
Exhibit B contains current Technical Specification-pa6es marked up with the proposed changes.
Exhibit C is a copy of the Monticello Technical Specifications incorporating the proposed changes.
This letter contains no restricted or other defense information.
NORTHERN # AT S
>4 COMPANY By 9
i
/ homas M' Parker T
Director Nuclear Licensing on this k ay o
/ fk 3 before me a notary public in and for said County, personally apg6, red Thomas M Parker, Director, Nuclear a
Licensing, and being 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 statements made in it are true and that it is not interposed for delay.
i MARQA K. LeCOM NOTARY PUSUC -MINNIS0f 4 '
HENNCPIN COUNTY h
My Commission i$tes Sept 24,199)
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-9302220253 930212 PDR ADOCK 05000263 P
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Exhibit A MONTICELLO NUCLEAR CENERATINC PLANT f
License Amendment Request Dated February 12, 1993 I
J Evaluation of proposed changes to the Technical Specifications f
for Operating License DPR 22 i
1 j
Pursuant to 10 CFR Part 50, Section 50.59 and 50.90, the holders of Operating.
J License DPR 22 hereby propose the following changes to the Monticello I
Technical Specifications:
1 i
a j
Eggg Section Proposed Changes J
52 Table 3.2.2 Revise Core Spray and Low Pressure Coolant Injection trip function A.1.b.ii to delete the word "and" from I
the function description (Reactor Low Pressure Permissive Bypass Timer) and to refer to Required Condition "B" in lieu of Required Condition "C".
i j-Also, correct the spelling of the word " Channels" in the heading for the second column from the right.
I
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53 Table 3.2.2 Revise llPCI System trip functions B.1 (Iligh Drywell Pressure) and B.2 (Low Low Reactor Water Level) to j
refer to Required Condition "A"
in lieu of "B".
l-Also, revise Automatic Depressurization System trip l
functions C.1 (Low Low Reactor Water Level), 0.2 (Auto i
Blowdown Timer), and C 3 (Low Pressure Core Cooling.
Pumps Discharge Pressure Interlock) to refer to Required Condition "B" in lieu of "C".
54 Table 3.2.2 Revise Diesel Generator trip functions D.2 (Low Low Reactor Vater Level) and D.3 (lligh Drywell Pressure) to refer to Required Condition "C" in lieu of "D".
55 Table 3.2.2 Delete the existing Required Condition "B", and re.
identify remaining Required Conditions "C" and "D" as 1
i MB" and "C", respectively.
60d Table 3.2.8 Revise Required Condition "B"
to refer to Specification 3.5.D in lieu of 3.5.E 2 Also, delete redundant " status" from the. descriptica of Required Condition "C." near bottom of the page, t
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1 101 4.5.A.1 Revise the minimum required flow rate of the Core Spray Pumps upwards frorn 2.700 gpm to 2. 800 gpm.
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107 3.5.F.1.a.2 The specification currently reads:
i "The Maximum Average Planar Linear lleat j
Generation Rate (MAPLilGR) will be changed as i
noted in Table 3.11.1."
i Revise this specification to read:
i "The Maximurn Average Planar Linear lleat
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Generation Rate (MAPLilGR) will be changed as i
noted in Table 1 of the Core Operating Litait~
j Report."
110 3.5/4.5 Bases The second sentence of the second paragraph of Part A the ECCS Bases currently reads i
1 "The Core Spray purnp is designed to deliver i
greater than or equal to 3,020 gpm (safety 1
analysis assumed 2700 gpin) against a system head corresponding to a reactor pressure 130 psi l
greater than containment pressure."
Revise the above sentence to read:
l "The Core Spray pump is designed to deliver greater than or equal ~ to 3,020 spin (the i
SAFER /GESTR IACA safety analysis assumed a Core j
Spray pump flow of 2,800 gpm, or 2,700'gpm flow into the core + 100 gpm to account for ECCS bypass-leakage) against a system head corresponding to a reactor pressure 130 psi i
j greater than containment pressure."
l Also, in the fifth paragraph on this page, reference l
Specification 3.5.A.), in lieu of 3.5.A.2.
i 113 3.5/4.5 Bases Revise the.last paragraph of this page to refer to Part E Specification 3.5.E.7 in lieu of 3.5.E.s.
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127 3'.6.E.1 The specification currently reads::
i
- E.
Safety /Rel'ief Valves 1.
During power operating conditions-p i
and whenever. reactor coolant.
l pressure is greater than 110 psig_
and temperature is greater than i
345'F:
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a.
The safety valve function (self actuation) of seven safety / relief valves shall be
- operable, b.
The solenoid activated relief function (Automatic Pressure Relief) shall be operable as required by Specification 3.5.E.
c.
The Low Low Set Function for three non Automatic Pressure Relief Valves shall be operable as_ required by Specification 3.2.11. "
Revise this speaification to read as follows:
"E.
Safety / Relief Valves 1.
During operating conditions and whenever reactor coolant pressure is less than 100 psig and temperature is greater than 345+F, the safety valve function (self actuation) of seven safety / relief valves shall be operable (Note: Low Low Set and ADS requirements are located in Specifications 3.2.11 and 3.5. A.
respectively)."
3.6.E.2 Revise this specification to refer to Specification 3.6.E.1 in lieu of 3.6.E.1 n.
151 3.6/4.6 Bases The first paragraph on this page currently reads:
"The safety / relief valves have two functions; i.e. power relief or self actuated by high pressure. The solenoid actuated function (Automatic Pressure Relief) in which_ external instrumentation signals of coincident high drywell pressure and low low water level initiate opening of the valves.- This function is discussed in Specification 3.5.E.
In addition, the valves can be operated manually."
Revise this_ paragraph to read:
"The. safety / relief valves have two functions; 1) over pressure relief (self actuated by high' pressure), and 2) Depressurization/ Pressure A3 i
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Control (using air actuators to open the valves J
via ADS, Low Low Set system, or manual operation).
The Low Low Set and ADS functions i
j are discussed further in Sections 3.2 and 3.5."
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156 3.7.A.1 Revise this specification to refer to Specification 3.5.L2 in lieu of 3.5.M.
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Reaton for Chance
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The portion of the change related to increasing the required Core Spray Pump i
flow from 2,700 gpm to 2,800 gpm is intended to account for the flow losses j
(bypass leakage paths) inherent to the Emergency Core Cooling Systems-(ECCS) j design.
Increasing the required flow rate for the Core Spray Pumps will assure that the total flow entering the core (ECCS pucp flow minus bypass leakage) during a Loss of Coolant Accident (LOCA) is consistent with the value j
assumed in the Monticello SAFER /CESTR 14CA Analysis.
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The changes to the 3.6/4.6 Bases discussion on page 151 are intended to j
clarify i.nd correct existing statements that are both confusing and misleading, The current wording states, incorrectly, that coincident high
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drywell pressure and low low water level signals initiate automatic actuation of the safety relief valves.
This is no longer true because of a modification-j performed in response to NUREG 0737, item II.K.3.18 (
Reference:
License Amendment No. 62 dated March 31, 1989).
The correct discussion of this 4
function is provided in Section 3.2 of the. Technical Specifications.
The l
proposed change will address this discrepancy and reference the proper information.
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j Similarly, the changes to Specification 3.6.E are intended to clarify the intent of the specification with respect to Automatic Depressurization System j-(ADS) and Low Low Set system requirements.
As presently written, Specifications 3.2.H, 3.5 A and 3.6.E cross reference each other in a manner j
that could lead to misinterpretation of the governing requirements for these 1
systems. The language of the proposed change is intended to alleviate this concern.
i l
The remaining changes are editorial in nature and are intended primarily to correct branching errors that occurred in previous License Amendments. Most of these errors resulted from License Amendment 79 (SAFER /CESTR), dated April 9, 1991, in which Section 3.5/4.5 (Core and containment Cooling Systems) was substantially rewritten and reorganized.
Several ' specifications were either i
deleted or re munbered at that time and rela,ted changes to associated cross-1 references were missed.
i Safety Evaluation!
1.
Increase in Core Sorav Pumo Reauired Flow Rate:
Technical Specification 4.5.A.1 currently requires that the Core Spray l
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l Pumps develop a flow rate of 2,700 gpm against a system head 1
corresponding to a reactor pressure of 130 psi greater than the I
containment pressure. Technical Specification 4.5.A.2 requires that the 1
Low Pressure Coolant Injection (LPCI) pumps develop a flow rate of 3,870 gpm, corresponding to two pumps delivering 7,740 gpm, at a reactor pressure of 20 psi greater than containment pressure.
The SAFER /GESTR.
IDCA analysis prepared for Monticello by General Electric incorrectly utilized the above flow rates to represent actual flow into the core.
Due to the design of the Core Spray and LPCI Systems, there are minor flow losses-(bypass leakage paths) that cause the actual flow rate into the core to be slightly less than the measured discharge flow rate of the pumps. The Core Spray System is assumed to have 20 gpm leakage from i
a 1/4 inch vent hole in the T box which is located between the inner reactor vessel wall and the core shroud.
The LPCI system is assumed to have 50 gpm leakage from slip joints on the jet pump assemblies. These flow diversions are treated as leakage paths because the associated coolant goes into the annulus region of the vessel and would flow out the postulated Design Basis Loss of Coolant Accident (DBA LOCA) j Recirculation System suction-line break, t
An evaluation was performed (
Reference:
Nonconforming Item Report 92-a 037) which confirmen that the actual flow rates for individual ECCS pumps minus assumed leakage was adequate to meet the flow rates assumed in the SAFER /GESTR LOCA analysis, therefore there were no immediate
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operability concerns. 11owever, the discrepancy between the flow rates required by the Technical-Specifications and the valuss assumed in the 4
SAFER /CESTR 1DCA analysis remains. To resolve this issue, we propose to
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increase the required Core Spray flow rate by 100 gpm (20 gpm to account i
for Core Spray leakage + 50 gpm to account for LPCI leakage + 30 gpm for margin) to account for all of the assumed ECCS bypass leakage paths.
i The LPCI flow rate currently required by the Technical Specification l
(3,870 gpm per pump / 7,740 gpm total) would remain unchanged.
This issue has been discussed with General Electric, who performed the i
Monticello SAFER /CESTR 1DCA analysis.
General Electric has concluded that with respect to the analysis, it is of no significance whether the assumed 70 gpm bypass leakage (increased to 100 gpm to provide 30 gpm i
margin) is accounted for by increasing Core Spray flow, LPCI flow, or i
both, llowever, when the trade off between increasing Cor. Spray or LPCI flow is con =idered, increased Core Spray flow is preferred for the following reasons:
l a.
In addition to replenishing vessel water inventory lost during the DBA-LOCA, Core Spray flow (which is injected into the vessel above i
the core) is more effective in collapsing any steam bubble that might form in the vessel, b.
The Core Spray pumps-deliver flow to the reactor vessel at higher reactor pressures than the Residual 11 eat Removal (RHR) pumps operating in the LPCI mode, which is beneficial in mitigating a.
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An additional factor in our decision to account for all ECCS assumed t
J bypass leakage by increasing Core Spray flow involves the relative l'
l capacities of the Core Spray and RilR pumps.
Each of the four RilR pumps i
(which provide LPCI flow) is currently capable of consistently meeting i
the existing Technical Specification flow rate requirement of 3,870 gpm.
i A review of recent surveillance test results has confirmed that the pumps are also capable of toeeting the slightly higher flow rate assumed i
by the SAFER /CESTR-LOCA analysis (3,895 gpm, which equates to an i
additional 25 gpm per operating pump assuming only two pucps are running, to account for the total LPCI bypass leakage of 50 gpm).
llowever, the higher value (3,895 spm) is very near the upper limit of RilR pump capacity, and there is insufficient margin remaining to ensure the pumps would consistently achieve this higher flow in the future.
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Conversely, Core Spray pump performance is such that the minimum required flow could be increased by 100 gpm to 2,800 gptn without difficulty. The current test criteria for the Core Spray pumps conservatively specifies an acceptance criteria of 3,020 gpin against a system head corresponding to 130 psi greater than containment pressure.
l Thus, the current test criteria provides a margin of more than 200 gpm over the proposed new Technical Specification criteria.
l The combination of ECCS pumps available for each single failure evaluated for a DBA-LOCA by the SAFER /GESTR lhCA analysis includes a Core Spray pump whenever two LPCI (RilR) pumps are available. Therefore, a Core Spray pump would always be available to provide the additional flow necessary to offset the assumed LPCI bypass leakage.
As discussed above, the proposed change will adequately resolve the discrepancy between the current Technical Specification Emergency Core Cooling Systems pump flow rates and the flow into the core assumed by the SAFER /CESTR lhCA analysis. The change is primarily administrative and has no impact on plant safety, since the basic assumptions supporting the SAFER /GESTR lhCA analysis, and therefore the conclusions of the analysis, remain unchanged.
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2.
Edftorial Changes to Correct Branching Errors. Cross Reference Errorn I
and Clarifv/ Correct Bases Information:
The changes to the 3.6/4.6 Bases discussion should have been included as part of License Amendment No. 62, dated March 31, 1989.- Amendment No.
62 reflected ' modifications to the Automatic Depressurization System
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logic that, among other things, removed the High Drywell Pressure interlock in response to NUREG 0737 Item II.K.3.18.
Other portions of Technical Specifications affected by the modification were updated appropriately, but the necessary changes to page 151 were missed.
Safety considerations associated with the Automatic Depressurization System logic change were fully addressed at the time Amendment No. 62 was processed and the proposed correction does not present any.new safety questions or concerns. The proposed change is necessary to ensure the 3.6/4.6 Bases discussion is consistent with the. intent of the-remainder of the Technical-Specifications, A-6 i
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The remaining changes are editorial in nature and do not change the intenc of the existing Technical Specifications. Most of these changes serve to correct internal branching and cross reference errors that occurred during previous license amendments. The remaining changes clarify, but do not change, the intent of existing spacifications.
These changes have no impact on plant safety.
Determination of Sirnificant Hazards Consideration!
This proposed change to the Operating License has been evaluated to determine if it constitutes a significant hazards consideration as required by 10 CFR Part 50, Section 50.91 using the standards provided in Section 50.92. This analysis is provided below:
a.
The cronosed amendment will not involve a sienificant inercase in the probability or consecuences of an accident nreviousiv evaluated.
Increasing the required Core Spray pump flow rate to 2,800 gpm will make the Technical _ Specification consistent with the SAFER /GESTR LOCA analysis.
The change is in the conservative direction (increased ECCS flow) and will not increase the probability or consequences cf a DBA+
LOCA or any other accident previously analyzed.
The remaining changes proposed are editorial or administrative in nature and have no impact on the probability or consequences of any accident previously evaluated, b.
The orgposed amendment will not create the nossibility of a new or different kind of necident from any accident nreviousiv analyzed.
The proposed changes, including the revised Core Spray pump flow rates and the 3.6/4.6 Bases changes, are primarily editorial or administrative in nature. No safety related equipment, safety function, or plant operations will be altered as a result of the proposed changes.
Therefore, the proposed amendment does not in any way create the possibility of a new or different kind of accident from any accident previously evaluated, c.
The nronosed amendment will not involve a significant reduction in the margin of safely.
The proposed amendment will not reduce the margin of safety because the Core Spray pump flow is being conservatively increased so that total ECCS pump flow into the core is consistent with that assumed by the SAFER /GESTR-LOCA analysis. The remaining changes are either editorial-in nature or are based on previously reviewed and approved Technical Specifications and have no impact on plant safety.
Based on the evaluation dsscribed above, and pursuant to 10 CFR Part 50, Section 50.91, Northern States Power Company has determined that operation of the Monticello Nuclear Generating Plant in accordance with the proposed A7
license amendment request does not involve-any significant hazards considerations as defined by NRC regulations in 10 CFR Part 50, Section 50.92.
Environmental AssessmeDI:
Northern States Power has evaluated the proposed changes and determined that:
1.
The changes do not involve a significant hazards consideration.
2.
The changes do not involve a significant change in the types or significant increase in the amounts of any effluentsLthat may be released offsite, or 3.
The changes do not involve a significant increase in individual or
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cumulative occupational radiation exposure.
Accordingly, the proposed changes meet the eligibility criterion for categorical exclusion set forth in 10 CFR Part 51 Section 51.22(c)(9).
Therefore, pursuant to 10 CFR Part 51 Section 51.22(b), an environmental assessment of the proposed changes is est required.
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