ML20116D816
| ML20116D816 | |
| Person / Time | |
|---|---|
| Site: | Monticello  |
| Issue date: | 07/26/1996 |
| From: | Hill W NORTHERN STATES POWER CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML20116D820 | List: |
| References | |
| NUDOCS 9608020338 | |
| Download: ML20116D816 (34) | |
Text
1 Northem States Power Company Monticello Nuclear Generating Plant 2807 West Hwy 75 Monticello, Minnesota 55362-9637 July 26,1996 10 CFR Part 50 Section 50.90 U S Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555 MONTICELLO NUCLEAR GENERATING PLANT Docket No. 50-263 License No. DPR-22 l
License Amendment Request Dated July 26,1996 Reactor Coolant Eouivalent Radioiodine Concentration and Control Room Habitability Attached is a request for a change to the Technical Specifications, Appendix A of the Operating License for the Monticello Nuclear Generating Plant. This request is submitted in accordance with the provisions of 10 CFR Part 50, Section 50.90.
This proposed amendment changes Technical Specification sections 3.6.C, Coolant Chemistry and 3/4.17.B, Control Room Emergency Filtration System. In addition, changes are submitted for the bases for these sections. The changes are proposed to establish Technical Specification requirements consistent with modified analysis inputs used for the evaluation of the radiological consequences of the main steam line break accident.
Exhibit A contains a description of the proposed changes, the reasons for requesting the changes, a Safety Evaluation, a Determination of Significant Hazards Consideration, and an Environmental Assessment. Exhibit B contains the current Technical Specification pages marked up with the proposed changes. Exhibit C contains revised Monticello Technical Specification pages. Exhibit D,"MNGP MSLBA Analysis Summary," provides a summary of the modified main steam line break analysis and reports the results of the re-evaluation of this postulated accident scenario.
Please contact Marvin Engen, Sr Licensing Engineer, (612-295-1291) if you require further information related to this request.
M/dAp William J Hill Plant Manager Monticello Nuclear Generating Plant
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9608020338 960726 _
PDR ADOCK 05000263 P
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USNRC NORTHERN STATES POWER COMPANY July 26,1996 Page 2
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c: Regional Administrator-Ill, NRC NRR Project Manager, NRC Resident inspector, NRC State of Minnesota Attn: Kris Sanda Attachments: Affidavit to the US Nuclear Regulatory Commission Exhibit A - Evaluation of Proposed Changes -
Exhibit B - Proposed Changes Marked Up on Existing Technical Specification Pages Exhibit C - Revised Monticello Technical Specification Pages Exhibit D - MNGP MSLBA Evaluation Summary l
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UNITED STATES NUCLEAR REGUl.ATORY COMMISSION NORTHERN STATES POWER COMPANY l
l MONTICELLO NUCLEAR GENERATING PLANT DOCKET NO. 50-263 REQUEST FOR AMENDMENT TO OPERATING LICENSE DPR-22 i
LICENSE AMENDMENT REQUEST DATED JULY 26,1996 REACTOR COOLANT EQUIVALENT RADIOlODINE CONCENTRATION 1
AND CONTROL ROOM HABITABILITY Northem States Power 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 Hazards Consideration and an Environmental Assessment. Exhibit B contains current Technical Specification pages marked up with the proposed changes. Exhibit C contains the affected Monticello Technical Specifications pages with the proposed changes incorporated. Exhibit D, provides a summary of the analysis which supports the proposed change.
This letter contains no restricted or other defense information.
NORTHERN STATES POWER COMPANY By M) 4 William J Milf Plant Manager Monticello Nuclear Generating Plant On this 2 dayof Mv l 7'llo before me a notary public in and for said County, personally appeared' William J Hill, Plant Manager, Monticello Nuclear Generating Plant, and being first duly swom acknowledged that he is authorized to execute this document on behalf of Northam 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.
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,dW MARVIN RICHARD ENGEN
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Mahlin R EngenNotary Public-Minn/
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l EXHIBIT A 1
i Monticello Nuclear Generating Plant License Amendment Reauest Dated July 26.1996 i
Evaluation of Proposed Changes to the Technical Specifications j
for Operating License DPR-22 1.
Reason for Proposed Changes i
Changes are proposed to Technical Specification sections 3.6.C, Coolant Chemistry and 3/4.17.B, Control Room Emergency Filtration System. In addition, changes are submitted j
for the bases for these sections. The changes are proposed to establish Technical Specification requirements consistent with modified analysis inputs used for the evaluation j
of the radiological consequences of the Main Steam Line Break (MSLB) accident. -
The postulated MSLB accident involves an instantaneouc circumferential break of a main steam line outside primary containment. Break flow is limited by a main steam line flow restrictor. Main Steam isolation Valve (MSIV) closure is initiated due to low steam line pressure and a reactor scram is initiated when the MSIVs begin to close. The current licensing basis for the MSLB accident is provided in USAR Section 14.7.3.2. The analysis of the MSLB radiological consequences presented in the USAR determined that the offsite doses are well below the guidelines of 10CFR100..
l Monticello has re-evaluated the radiological consequences of the MSLB accident taking into consideration current regulatory guidance and the analysis inputs of the current licensing basis. The current licensing basis analysis assumes a MSLB with the reactor at i
full power. During the re-analysis of this accident, it was identified that a postulated MSLB from a hot standby condition would provide a greater potential hazard than a full power MSLB. Hot standby is operation with the reactor critical in the startup mode at a power
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level sufficient to maintain reactor pressure and temperature. Hot standby is not a normal long term operating condition and the plant is in this condition only briefly during normal plant startups and shutdowns. Analysis of the MSLB from the hot standby condition results in a high rate of depressurization and rapid rising of water level to the main steam line inlet, thus maximum coolant mass is released through the brea.k. The hot standby MSLB accident has a greater mass release and is thus a more conservative condition for analysis j
of the radiological consequences of this accident as compared to the full power condition.
The radiological consequences of the more conservative hot standby MSLB accident were analyzed using the current licensing basis analysis inputs for the source term, control room i
ventilation filter bypass leakage and control room filtration system efficiencies while taking into consideration the appropriate contemporary regulatory guidance and revised inputs for dose conversion factors and atmospheric dispersion factors. This analysis of the hot standby MSLB determined that the calculated doses do not exceed the exposure guidelines of 10CFR100 and 10CFR50 Appendix A, General Design Criterion 19. These A-1
licensing basis analysis inputs provide a conservative analysis of the radiological consequences to the MSLB accident.
The licensing basis inputs used in the analysis of the hot standby MSLB provide a conservative evaluation of this design basis accident. However, the regulatory guidance for the analysis of these events has continued to evolve since the initiallicensing of MNGP.
MNGP recognizes the modifications to the guidance provided in the NRC Regulatory Guides as well as the guidance provided in NUREG-0800, " Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants." MNGP has further analyzed the hot standby MSLB taking into consideration the appropriate contemporary regulatory guidance using updated analysis inputs. A summary of this analysis is provided as Exhibit D,"MNGP MSLBA Evaluation Summary." This further analysis similarly determined conformance to the guidelines of 10CFR100 and 10CFR50 Appendix A, General Design Criterion 19.
To provide additional margin to these exposure criteria for the hot standby MSLB radiological consequences analysis, the analysis inputs (consistent with the regulatory guidance) used values for the reactor coolant radioiodine concentration and Control Room Emergency Filtration system filter efficiency consistent with the Technical Specification changes proposed herein. The proposed changes to the plant Technical Specifications described in this license amendment request are consistent with the analysis inputs for the MSLB presented in Exhibit D.
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- 11. Description of Proposed Changes Pursuant to 10CFR50.90, Northern States Power Company hereby propose the following changes:
A. Reactor Coolant Chemistry Equivalent Radioiodine Concentration
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- 1. Technical Specification Section 3.6, PRIMARY SYSTEM BOUNDARY. Specification 3.6.C.1, Coolant Chemistry, page 123.
a) Specification 3.6.C.1 states:
- 1. The steady state radiolodine concentration in the reactorcoolant shallnot exceed 5 microcunes ofI-131 dose equivalent per gram of water.
i b) The specification is proposed to be changed to state:
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- 1. The steady state radioiodine concentration in the reactor coolant shall not exceed 2 microcuries of I-131 dose equivalent per gram of water.
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- 2. Technical Specification Section 3.6 and 4.6 Bases, Section C, Coolant Chemistry.
l page 148.
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The bases discussion contained in the first paragraph on page 148 is to be revised to reflect the analysis performed of the main steam line break accident radiological consequences. The bases will reflect that the main steam line break accident radiological consequences analysis demonstrated that the resulting dose consequences are well within the guidelines of 10CFR100 using the analysis input of 2 microcuries of I-131 dose equivalent per gram of water in the reactor coolant for the steady state radioiodine concentration limit.
B. Control Room Habitability i
Specification 3.17.B.2, Control Room Emergency Filtratica System Performance Requirements, page 229w.
a) Specification 3.17.B.2 states:
- 2. Perfonnance Requirements
litters shall show $1% DOP penetration.
(2) The results ofin-place halogenated hydrocarbon tests at 1000 cfm (210%) on charcoal banks show S1% penetration.
(3) The results oflaboratory carbon sample analysis shallshow >98%
l methyllodide removal efficiency when tested at 80 C, 95% R.H.
b) The specification is proposed to be changed to state:
- 2. Performance Requirements
- a. Periodic Requirements (1) The combined results of the in-place DOP tests at 1000 cfm (110%)
for the HEPA filters shall show $ 0.3% DOP penetration.
(2) The results of in-place halogenated hydrocarbon tests at 1000 cfm (110%) on charcoal banks show 5 0.3% penetration.
(3) The results of laboratory carbon sample analysis shall show the methyllodido penetration S 0.4% when tested at 30 *C and 95%
- relative humidity.
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Specification 3.17.B.3, Post Maintenance Testing Requirements, page 229x.
a) Specification 3.17.B.3 states:
- 3. Post Maintenance Requirements
- a. Afterany maintenance ortesting that could affect the HEPA MIteror i
HEPA Miter mounting frame leak tight integrity, the results of the in-place l
DOP tests at 1000 cfm (110%) on HEPA filters shall show S 1% DOP penetration.
- b. After any maintenance or testing that could affect the charcoal adsorber leak tight integrity, the results ofin-place halogenated hytitocarbon tests at 1000 cfm (110%) on charcoal adsorber banks shall show $1%
b) The specification is proposed to be changed to state:
- 3. Post Maintenance Requirements
- a. After any maintenance or testing that could affect the HEPA filter or-HEPA filter mounting frame leak tight integrity, the combined results of l
the in-place DOP tests at 1000 cfm (110%) on HEPA filters shall show1 l
0.3% DOP penetration.
- b. After any maintenance or testing that could affect the charcoal adsorber
- leak tight integrity, the results of in-place halogenated hydrocarbon tests at 1000 cfm (110%) on charcoal adsorber banks shall show 5 0.3%
- 3. Technical Specification Section 3.17, Control Room Emeraency Filtration System.
l Bases, page 229y. Technical Specification Section 3.17, Section B, Control Room Emeroency Filtration System. Ha.s.elt, page 229z.
a) The bases for section 3.17.A, page 229y, are revised such that detailed design information which does not directly support the Technical Specification is removed.
b) The bases for section 3.17.B. page 229y, are revised to reflect an emergency l
filtration system iodide removal efficiency of 98% as an input in the analysis l
performed of the main steam line break accident radiological consequences.
l Paragraph two of this section is revised to be consistent with the proposed
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changes to the Technical Specifications which establish conservative testing i
criteria with respect to the iodide removal efficiency. Paragraph three of this section of the bases is proposed to be revised to reflect the inputs used in the
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analysis for control room dose calculations: 85% standby gas treatment system i
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adsorption and filtration efficiency and 98% control room emergency filtration system adsorption. The control room doso calculations confirmed that control room personnel whole body and organ doses remained within the guidelines of 10CFR50 Appendix A, GDC 19.
l c) The bases for section 4.17, page 229z, is proposed to be revised to reflect that the revised guidance of ANSI N510-1989 is to be used for ventilation filter testing.
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i lit. Safety Assessment of the Proposed Changes i
A. Coolant Chemistry Equivalent Radioiodine Concentration Technical Specification Section 3.6, PRIMARY SYSTEM BOUNDARY. Specification 3.6.C.1, Coolant Chemistry, page 123.
Technical Specification Section 3.6 and 4.6 Bases. Section C, Coolant Chemistry.
page 148.
A change is proposed to specification 3.6.C.1 and the associated bases to establish the limiting condition for operation such that the steady state radioiodine concentration in the reactor coolant shall not exceed two (2) microcuries of I-131 dose equivalent per gram of water. This change is proposed to reflect an analysis input used in the evaluation of the radiological consequences of the main steam line break accident.
During circulation, the reactor coolant acquires radioactive material due to activation of corrosion products and release of fission products from potential fuelleaks. The release of coolant during a design basis accident could release radioactive materials into the environment. A limit is established in the plant Technical Specifications on the maximum allowable level of radioactivity in the reactor coolant to ensure that in the event of a release of any radioactive material to the environment during a design basis main steam line break accident, radiation doses are maintained within the guidelines of 10CFR100. The steady state radioiodine concentration in the reactor coolant is an initialinput for analysis of the radiological consequences of an accident due to a main steam line break outside of containment. No fuel damage is postulated in the main steam line break accident, and the release of radioactive material from the break ends shortly after the main steam isolation valves close completely.
MNGP has operated well within the technical specification limits for reactor coolant concentration of dose equivalent 1-131. Data readily available from the last eight operating cycles, shows that the reactor coolant concentration of dose equivalent 1-131 has been a small fraction of the Technical Specification limit with a cycle average of 4.5 X 10'* micro curies per gram over the last eight operating cycles. Even though the reactor core may contain no defective fuel, trace amounts of natural uranium in core construction materials and zircaloy cladding, as well as traces of enriched uranium on the extemal cladding surface, could be a source of fission products in the coolant during power operation.
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i The basis for the current Technical Specification limit for the reactor coolant radiolodine concentration of 5 microcuries per gram of dose equivalent 1-131 is derived from the analysis of the main steam line break accident performed by the Atomic Energy l
Commission (AEC), predecessor to the NRC, in support of the issuance of the MNGP provisional operating license. License Amendment Number 9 to the Provisional Operating License, issued April 10,1975, provided a change to the limiting condition for the reactor coolant activity and revised the technical specification bases to be consistent with the standard technical specifications in effect at the time of issuance of the amendment as well as the inputs used in the AEC analysis of the main steam line break accident.'
4 The Main Steam Une Break accident radiological consequences have been re-analyzed. This analysis was performed using modified inputs consistent with the current regulatory guidance.- The analysis conservatively used an input for the dose equivalent reactor coolant 1-131 concentration of 2 microcuries per gram. This analysis input provides a conservative assessment of the potential radiological consequences.
The results of this analysis are provided in Exhibit D. The analysis demonstrated that
. the dose consequences from a potential main steam line break outside of primary containment are well within the guidelines of 10CFR100 and 10CFR50 Appendix A, i
General Design Criterion 19. The Technical Specification bases are to be revised to delete bases information which is not pertinent to the specification and to reflect the appropriate information which is consistent with the accident analysis.
The bases for the Technical Specifications state that radioiodine concentration can change rapidly in the reactor coolant during transient reactor operations such as reactor shutdown, reactor power changes and reactor startup if failed fuelis present. As specified in the Technical Specifications, additional reactor coolant samples shall be taken and analyzed for reactor operations in which steady state radioiodine concentrations in the reactor coolant indicate iodine release from the fuel. The capability to detect fuel element failures is inherent in the radiation monitors in the off-gas system. Based on operating experience, if fuel defects were to occur, appropriate plant actions would be taken based upon off-gas release values which would ensure the integrity of the MSLB safety analys:s. Maintaining off-gas releases to within acceptable values provides adequate operating constraints such that it is unlikely for transient reactor operations to result in radioiodine concentrations greater than the proposed Technical Specification value.
The proposed change to the plant Technical Specifications and Bases is consistent with the
. analysis performed of the main steam line break accident radiological consequences. The proposed changes do not result in a significant increase in the probability or consequences of postulated accidents previously analyzed, an accident not previously analyzed, or in a significant reduction in the margin of safety.
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i B. Control Room Habitability Technical Specification Section 3.17, CONTROL ROOM HABITABILITY.
Specification 3.17.B.2, Control Room Emergency Filtration System Performance Requirements, page 229w.
Technical Specification Section 3.17, CONTROL ROOM HABITABILIT(.
Specification 3.17.B.3, Post Maintenance Testing Requirements, page 229x.
Technical Specification Section 3.17, pontrol Room Emeroency Filtration System.
Bases. page 229y. Technical Specification Section 4.17, Section B, Control Room Emeraency Filtration System. Bases. page 229z.
Changes are proposed to the plant Technical Specifications and Bases to modify the limiting conditions for operation and surveillance requirements for the control room emergency filtration system. The changes are proposed to establish conservative limiting conditions and testing criteria with respect to the filter and charcoal adsorber efficiencies used in the analysis of control room operator doses for design basis accidents.
The function of Control Room Ventilation-Emergency Filtration Train (CRV-EFT)
System is to maintain the environment of the Main Control Room, thereby ensuring it's habitability during normal and accident conditions. The CRV-EFT System is composed of two subsystems; the Control Room Ventilation (CRV) subsystem and the Emergency Filtration Train (EFT) subsystem. The function of the CRV portion of the system is to pwvHe the control room and the first and second floors of the EFT building (Reactor E Ing Adtz.tn) with conditioned air to maintain acceptable temperature conditions during normal operation. The EFT subsystem provides forisolation of the control room and the first and second floors of the EFT building from outside air during a toxic chemical release er an accident where high levels of activity may be released. During a radiological accident, the EFT provides for automatic isolation with immediate automatic pressurization of the Control Room with filtered air to minimize the activity, and therefore the radiological dose, inside the control room. The redundant air filtration units consist of the following components in series: a low efficiency filter, an electric heating element, a High Efficiency Particulate Air (HEPA) filter, two 2-inch charcoal adsorber beds, a HEPA filter, and a centrifugal fan. The charcoal adsorber removes gaseous iodine, while the HEPA filters remove particulate matter.
The EFT system is designed to satisfy the criteria of NUREG 0737, Section Ill.D.3.4 which imposes the criteria of 10CFR50 Appendix A, General Design Criterion (GDC)
- 19. The EFT is designed to provide adequate radiation protection to permit access and occupancy of the control room under accident conditions without personnel receiving radiation exposures in excess of 5 rem whole body, or its equivalent to any part of the body, for the duration of the accident. Section 6.7 of the MNGP USAR provides additionalinformation on the CRV-EFT system.
The main steam line break accident radiological consequences have been re-analyzed.
The inputs used in this analysis provide a conservative assessment of the potential A-7 1
radiological consequences. The results of the analysis are provided in Exhibit D. The radiological analyses reflect an improvement for the control room emergency filtration system filter efficiency and reduced control room ventilation bypass leakage.
l A modification to the control room emergency filtration system is to be performed to establish the improved control room ventilation bypass leakage. With this modification, pressurization of the control room during normal system operation will not be a system function; however filtered pressurizing air will be supplied to the main control room upon automatic initiation of the EFT system. This modification will enhance the isolation of the envelop supp!!ed by the EFT system. In addition, it will minimize the potential for the introduction of non-filtered air during short term puff releases. The lack 1
l of normal pressurization for the control room does not have an adverse effect on l
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A 98% iodine removal efficiency for the EFT system was used as an input for the dose l
consequence analysis. A 98% removal efficiency is within the capability of the EFT ~
system charcoal adsorber as demonstrated in previous surveillance tests. Changes are proposed to the limiting conditions for operation and surveillance requirements for the EFT system consistent with the analysis input for the iodine removal efficiency used in the main steam line break radiological consequences evaluation. The proposed changes are consistent with the applicable regulatory guidance provided in NRC l
Generic Letter 83-013; Regulatory Guide 1.52, Revision 2; and NUREG-1433.
A performance criterion of less than or equal to 0.4% methyl iodide penetration for I
laboratory testing is proposed for specification 3.17.B.2.a(3) consistent with the 98%
charcoal adsorber efficiency. A safety factor of 5 was used in determining this test criterion consistent with applicable regulatory guidance for an EFT system with heaters in line with the process stream. The laboratory test of the EFT system charcoal adsorber is to be performed in accordance with the guidelines provided in Regulatory Guide 1.52, Revision 2 and ANSI /ASME N510-1989. The test conditions are proposed j
to be revised to 30*C and 95% relative numidity consistent with ASTM D3803-1989 and NUREG-4960 section 5.2.3.3.f. Thes9 test conditions are more representative of the service conditions for the EFT systera and establish a more conservative test condition.
The current Technical Specification test condition of 80*C is consistent with ASTM D3803-1979; however surveillane,e test have been performed at both the 80 C and 30'C test conditions for several surveillance cycles. A review of the test data for the I
test conditions of 80 C and 30 C supports the 30 C test condition as a more conservative test condition. Guidance provided in ASTM D3803-1989 states that the 30*C, 95% relative humidity methyl iodide test is the most reliable test method to establish iodide removal efficiency of any adsorbent.
The proposed changes to performance criteria and post maintenance testing criteria specified in technical specification requirements 3.17.B.2.a(1),3.17.B.2.a(2),
3.17.B.3.3, and 3.17.B.3.b establish in-place testing criteria consistent with the regulatcry guidance of NRC Generic Letter 83-013 for a system with a HEPA and charcoal adsorber efficiency of 98%. For in-place penetration testing of the HEPA j
filters with dioctyi phthalate (DOP), the HEPA filters upstream and downstream of the A-8 l
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l charcoal adsorbers are tested individually, with the individual tests then factored together to reflect the efficiency of the two HEPA filters in combination to satisfy the proposed criteria for Technical Specifications. The bases of the technical specifications are proposed to be revised to reflect the revised testing criteria and test methods.
The proposed changes to the Technical Specifications are provided to ensure testing is s
performed which provides a high level of assurance of the capability of the EFT system 1
to perform as analyzed in the evaluation of the control room operator doses resulting i
from a postulated main steam line break. The analysis determined that control room 4
operator doses remain below the regulatory guidelines of 10CFR50 Appendix A, GDC j
- 19. The proposed changes to the Technical Specifications ensum continued i
compliance with NUREG-0737, item lli.D.3.4 which requires that nuclear power plants i
be equipped with a control room from which actions can be taken to operate the plant l
safely under normal and accident conditions. The testing criteria has been established in accordance with the applicable regulatory guidance while providing conservative margin to the analytical inputs used in the safety analyses, thus the propossd changes are acceptable. The proposed changes do not result in a significant increase in the probability or consequences of postulated accidents previously analyzed, an accident not previously analyzed, or in a significant reduction in the margin of safety.
IV. Determination of Significant Hazards Considerations The proposed change to the Operating License has been evaluated to determine whether it constitutes a significant hazards consideration as required by 10CFR50.91 using standards provided in 10CFR50.92. This analysis is provided below:
A. The proposed amendment will not involve a significant increase in the probability or consequences of an accident previously evaluated.
A limit is established in the plant Technical Specifications for steady state radiolodine concentration in the reactor coolant to ensure that in the event of a release of radioactive material to the environment during a design basis main steam line break accident, radiation doses are maintained within the guidelines of 10CFR100. The steady state radioiodine concentration in the reactor coolant is an input for analysis of the radiological consequences of an accident due to a main steam line break outside of containment. In addition, requirements are established in the Technical Specifications for control room habitability. During an accident, the control room emergency filtration system provides filtered air to pressurize the Control Room to minimize the activity, and therefore the radiological dose, inside the control room.
A change is proposed for the steady state rad:olodine concentration. This value is consistent with the value used in the main steam line break dose consequences analysis. Changes are proposed to the limiting conditions for operation and surveillane'; requirements for the control room emergency filtration train iodine removal efficiency. Rue 'ahanges are consistent with the inputs used in the analysis of the main steam lim areak radiological consequences. These proposed requirements A-9 l
r maintain operating restrictions for analyticalinputs used in the analysis of the main steam line break accident.
The analysis of the main steam line break accident performed with these inputs demonstrated that radiological consequences of the main steam line break are not changed significantly. The radiological consequences of the main steam line break accident remain within the exposure guidelines of 10CFR100 and 10CFR50 Appendix A, General Design Criterion 19. The offsite dose consequences remain bounded by the licensing basis provided in the AEC Provisional Operating License Safety Evaluation Report, dated March 18,1970. The control room doses calculated for the hot standby main steam line break accident using the TID 14844 dose conversion factors remain bounded by the dose consequences of the comparable design basis loss of coolant accident.
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l The proposed Technical Specification changes do not introduce new equipment l
operating modes, nor do the proposed changes alter existing system inter-i relationships. The system improvements to reduce bypass leakage during postulated accidents do not have an adverse effect on control room habitability. Therefore, this amendment will not cause a significant increase in the probability or consequences of an accident previously evaluated for the Monticello plant.
B. The proposed amendment will not create the possibility of a new or different kind of accident from any accident previously analyzed.
The proposed Technical Specification changes do not introduce new equipment operating modes, nor do the proposed changes alter existing system inter-i relationships. The system improvements to reduce bypass leakage during postulated accidents do not have an adverse effect on control room habitability. The proposed I
j changes are consistent with a modified analysis of the main steam line break accident i
j for a more conservative initial condition for the postulated accident. Therefore, the l
proposed amendment will not create the possibility of a new or different kind of l
accident.
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C. The proposed amendment will not involve a significant reduction in the margin of l
safety.
Surveillance data has demonstrated the proposed requirements are within the current capability of the facility. The proposed changes maintain margins of safety. These l
proposed requirements maintain operating restrictions for analytical inputs used in the analysis of the main steam line break accident. The analysis performed using these inputs determined that the rEdiological consequences of the main steam line break accident remain within the exposure guidelines of 10CFR100 and of 10CFR50 Appendix A, General Design Criterion 10. Therefore, the proposed amendment will not involve a significant reduction in the margin of safety, s
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l V. Environmental Assessment l
Northern States Power Company has evaluated the proposed changes and determined that:
- 1. The change does not involve a significant hazards consideration..
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2 The changes do not involve a significant change in the type or significant increase in the amounts of any effluent that may be released offsite, or
- 3. The changes do not involve a significant increase in individual or cumulative occupational radiation exposure.
Accordingly, the proposed changes meet the eligibility criteria for categorical exclusion set forth in 10 CFR Part 51, Section 51.22(b), an environmental assessment of the proposed changes is not required.
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