ML20249C509
| ML20249C509 | |
| Person / Time | |
|---|---|
| Site: | Monticello  |
| Issue date: | 06/19/1998 |
| From: | Schibonski C NORTHERN STATES POWER CO. |
| To: | NRC |
| Shared Package | |
| ML20249C506 | List: |
| References | |
| NUDOCS 9806300203 | |
| Download: ML20249C509 (16) | |
Text
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I UNITED STATES NUCLEAR REGULATORY COMMISSION i
NORTHERN STATES POWER COMPANY MONTICELLO NUCLEAR GENERATING PLANT DOCKET NO. 50-263 i
REQUEST FOR AMENDMENT TO OPERATING LICENSE DPR-22 I l
REVISION TWO TO LICENSE AMENDMENT REQUEST DATED JULY 26,1996 l REACTOR COOLANT EQUIVALENT RADIOIODINE CONCENTRATION I AND CONTROL ROOM HABITABILITY
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Northern States Power Company, a Minnesota corporation, requests authorization for changes I 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 Evoluat 'n, a Determination of No Significant Hazards Consideration and an Environmental /4 tssment. 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 and Exhibit E provide a summary of the analysis which supports the proposed change. Exhibit F contains an EFT system testing commitment and a list of exceptions from ASME N510-1989.
This letter contains no restricted or other defense information.
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NO THERN S PO P NY By / [ -
Craig A. Scfii6onski Acti Plant Manager ticello Nuclear Generating Plant On this I9 day ofIm M98 before me a notary public in and for said County, personally appeared Craig A. Schibonski, Acting 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, 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.
'S6muel I. Shirey SAMUEL l. SMREY Notary Public - Minnesota nownrusuc ammmasora l Sherburne County , W Comm. Em M. 31,2000 ,
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l 9806300203 980619 PDR ADOCK 05000263 P PDR
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EXHIBIT A Evaluation of Proposed Changes to the Technical Specifications
- 1. Reason for Proposed Changes 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 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 a postulated Main Steam Line Break (MSLB) accident, and of a postulated line break in the Reactor Water Cleanup (RWCU) system.
Main Steam Line Break Evaluation The postulated MSLB accident involves an instantaneous circumferential break of a main steam line outside primary containment. The current licensing basis for the MSLB accident is provided in USAR Section 14.7.3. The analysis of the MSLB radiological consequences presented in the USAR determined that the offsite doses are well below the guidelines of 10 CFR 100.
As part of the power rerate program, NSP 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 full power. During the evaluation of this accident, it was identified that a MSLB from a hot standby condition would provide a greater potential dose than a full power MSLB.
The hot standby MSLB accident has a greater mass release and represents a more conservative cordition for radiological analysis compared to the full power condition with less mass release. Hot standby is operation with the reactor critical in the startup mode at a power level sufficient to maintain reactor pressure and temperature. Hot standby is not a normallong term operating condition, and the plant is in this condition only briefly during normal plant startups and shutdowns. The postulated MSLB from the hot standby condition results in a high rate of depressurization and a rapid rise of water level to the main steam line inlet, resulting in a maximum coolant mass released through the break.
NSP analyzed the hot standby MSLB using the appropriate contemporary regulatory guidance with updated analysis inputs . A summary of this analysis is provided as Exhibit D, MNGP MSLBA Evaluation Summary. This analysis similarly determined conformance to the guidelines of 10 CFR 100 and 10 CFR 50 Appendix A, General Design Criterion 19.
The analysis inputs used a value for the reactor coolant radioiodine concentration of 2 Ci/gm and a value for Control Room Emergency Filtration system filter efficiency of 98%
The proposed changes to the plant Technical Specifications described in this license amendment request are conservative with respect to the radiciodine concentration and A-1 L____.-..-_.-___________ _ _ _ _ _ _ _
consistent with the Control Room Emergency Filtration Train iodine removal efficiency used as inputs for the MSLB evaluation summarized in Exhibit D.
RWCU High Energy Line Break Evaluation NSP identified a discrepancy in the mass and energy slease calculated for a postulated line j, break in the RWCU piping during power rerate engineering evaluations. The assumed 3 RWCU break mass flow rate for the RWCU line break evaluation was 244 lbm/sec. NSP has determined that the break flow rate is 719 lbm/sec. See LER 96-008, Reactor Water Clean Up Line Break Reanalysis Due to an Error Discovered During Re-evaluation, dated September 16,1996 for details. Plant operation with RWCU in service was shown to be acceptable and bounded by existing radiological analyses with an administrative limit of 0.25 Cl/gm for reactor coolant dose equivalent radiciodine concentration. This license amendment request proposes to incorporate the administrative limit into the plant Technical Specifications. )
i An evaluation of the radiological consequences of the postulated limiting RWCU line break was completed. The release was assumed to occur at ground level without Standby Gas Treatment System (SGTS) filtration. Results of the evaluation show that the radiological t consequences of the RWCU line break are well below the guidelines of 10 CFR Part 100 for offsite doses and below the guidelines of 10 CFR Part 50, Appendix A, GDC 19 for control room doses. This evaluation used the source term for reactor coolant radiciodine provided in the Monticello Updated Safety Analysis Report (USAR). The evaluation credits operator action to isolate the postulated RWCU line break, by closure of the containment isolation valves, ten (10) minutes after initiation of the postulated break. This assumption is very conservative because a safety related RWCU isolation system was installed during the 1998 refueling outage that will automatically isolate the break within one minute of a high room temperature or high flow signal. NSP had previously committed to install this isolation system by letter dated June 10,1997. NSP will submit a license areendment to incorporate j
requirements for the automatic RWCU isolation equipment into the technical specifications a following approval of this amendment request.
An additional evaluation was performed which demonstrated that the radiological consequences for the full spectrum of postulated RWCU line breaks remains bounded by the radiological consequences of the Main Steam Line Break as evaluated in Section 4.4 of the NRC Safety Evaluation Report, dated March 18,1970, supporting the Monticello provisional operating license. This evaluation used a source term based on a reactor coolant dose equivalent radioiodine concentration of 0.25 Ci/gm. ;
1 The atmospheric dispersion (X/Q) factors used in the analyses provided in Exhibit D and Exhibit E differ from those previously reviewed by the NRC staff. The atmospheric dispersion factors were determined using Monticello site meteorological data and the 1 methodology of NUREG/CR-5055, Atmospheric Diffusion for ControlRoom Habitability Assessments, by J.V. Ramsdell, Pacific Northwest Laboratory,1988; and Regulatory Guide 1.145, Atmospheric Dispersion Models for Potential Accident Consequence Assessments at Nuclear Power Plants, Revision 1, November 1982. The determination of the control room A-2 l l
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X/Q factors used the methodology of NUREG/CR-5055 for ground release from the Reactor Building plenum and the Turbine Building, and used the elevated stack release model from Regulatory Guide 1.145 for elevated releases from the plant Offgas Stack. The l
determination of offsite X/Q factors used the methodology of Regulatory Guide 1.145 for I gmund release from the Reactor Building plenum and the Turbine Building, and for the elevated releases from the plant Offgas Stack.
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- 11. Description of Proposed Changes l
Pursuant to 10CFR50.90, Northem States Power Company hereby proposes the following changes to the Monticello Technical Specifications. Proposed changes are indicated by bolded text.
A. Reactor Coolant Chemistry Equivalent Radiciodine Concentration j 1
- 1. Technical Specification Section 3.6, PRIMARY SYSTEM BOUNDARY, Specification 3.6 c. ., Coolant Chemistry, page 123.
. a) Specification 3.6.C.1 states: '
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- 1. The steady state radioiodine concentration in the reactor coolant shall not exceed 5 microcuries of I-131 dose equivalent per gram of water.
b) The specification is proposed to be changed to state:
- 1. The steady state radioiodine concentration in the reactor coolant shall not exceed 0.25 microcuries of I-131 dose equivalent per gram of water.
- 2. Technical Specification Section 3.6 and 4.6 Bases, Section C, Coolant Chemistry, page 148.
l 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 10 CFR 100 using the analysis input of 2 microcuries of lodine-131 dose equivalent per gram of water in the reactor I coolan'. foi the steady state radiciodine concentration limit. The bases will reflect that l
the radiological consequences of a postulated high energy line break in the RWCU system outside the drywell are well within the guidelines of 10CFR100 using the evaluation input cf 0.25 microcuries of lodine-131 dose equivalent per gram of water in the reactor coolant for the steady state radiciodine concentration limit.
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B. Control Room Habitability
- 1. Technical Specification Section 3.17, CONTROL ROOM HABITABILITY, l Specification 3.17.B.2, Control Room Emergency Filtration System Performance Requirements, page 229w.
1 a) Specification 3.17.B.2 states: !
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- 2. Perfonnance Requirements l
- a. Periodic Requirements (1) The results of the in place DOP tests at 1000 cfm (+10%) on HEPA filters shall show$1% DOP penetration.
(2) The results ofin-place halogenated hydrocarbon tests at 1000 cfm (110%) on chamoal banks show $1% penetration.
(3) The results oflaboratory carbon sample analysis shall show >98%
methyliodide 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
(1) The results of the in-place DOP tests at 1000 cfm (+10%) shall show l 51% DOP penetration on each individual HEPA filter and shall show l
$ 0.05% DOP penetration on the combined HEPA filters. I (2) The results of in-place halogenated hydrocarbon tests at 1000 cfm (110%) shall show$ 0.05% penetration on the combined charcoal I banks. I (3) The results of laboratory carbon sample analysis shall show < 0.4%
methyl iodide penetration when tested at 30 *C and 95% relative humidity.
- 2. Technical Specification Section 4.17, CONTROL ROOM HABITABILITY, i Specification 4.17.B.2, Control Room Emergency Filtration System Performance Requirement Test, page 229w.
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a) Specification 4.17.B.2 states:
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- 2. Performance Requirement Test a At least once per 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> . .
b) The specification is proposed to be changed to state: l
- 2. Performance Requirement Test The in-place performance testing of HEPA filter banks and charcoal adsorber banks shall be conducted in accordance with Sections 10 and 11 of ASME N510-1989 with exceptions as described in Section 6.7 of the USAR. The carbon sample test for methyl lodide removal shall be !
conducted in accordance with ASTM D 3803-89.
I a) At least once per 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br />.. j 1
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- 3. Technical Specification Section 3.17, CONTROL ROOM HABITABILITY, Specification 3.17.B.3, Post Maintenance Testing Requirements, page 229x.
a) Specification 3.17.B.3 states:
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- 3. Post Maintenance Requirements i
- a. After any maintenance or testing that could affect the HEPA filter or HEPA filter mounting frame leak tight integrity, the results of the in-place DOP l tests at 1000 cfm (210%) on HEPA filters shall show < 1% DOP
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penetration.
- b. After any maintenance or testing that could affect the charcoal adsorber l leak tight integrity, the results ofin-place halogenated hydrocarbon tests 1
at 1000 cfm (210%) on charcoal adsorber banks shall show <1%
penetration.
b) The specification is proposed to be changed to state: l
- 3. Post Maintenance Requirements i
- a. After any maintenance or testing that could affect the HEPA filter or HEPA filter mounting frame leak tight integrity, the results of the in-place DOP l tests at 1000 cfm (+10%) shall show < 1% DOP penetration on each l Individual HEPA filter and shall show < 0.05% DOP penetration on the combined HEPA filters.
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- 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 (+10%) shall show < 0.05% penetration on the combined charcoal adsorber banks.
- 4. Technical Specificatie n Section 3.17, Section A, Control Room Ventilation System, Bases, page 229y. Technical Specification Section 3.17, Section B, Control Room Emergency Filtration System, Bases, page 229y. Technical Specification Section 4.17, Section B, Control Room Emergency Filtration System, Bases, page 229z.
a) The bases for section 3.17.A, page 229y, are revised to reflect operation of the Control Room Ventilation system as moamed to reduce control room ventilation l
j bypass leakage. In addition, the bases are revised such that detailed design '
information which does not directly support the Technical Specification is removed.
b) The bases for section 3.17.8, page 229y, are revised to reflect an emergency filtration system iodine removal efficiency of 98% as an input in the analysis performed of the Main Steam Line Break Accident radiological consequences.
Paragraph two of this section is revised to be consistent with the proposed ;
changes to the Technical Specifications which establish conservative testing !
criteria with respect to the iodine removal efficiency. Paragraph three of this I section of the bases is proposed to be revised to reflect the inputs used in the analysis for control room dose calculations: 85% standby gas treatment system adsorption and filtration efficiency, and 98% control room emergency filtration system adsorption. The control room dose calculations confirmed that control room personnel whole body and organ doses remained within the guidelines of 10CFR50 Appendix A, GDC 19.
c) The bases for section 4.17.8, page 229z, are revised to reflect the performance requirements for EFT filter testing. I i
in addition, the bases are revised to establish consistency between the Technical l Specification bases and Technical Specification 3.17.B.2.b(1). Technical '
Specification 3.17.B.2.b(1) states that the Control Room Emergency Filtration :
System shall be shown to be operable with a combined filter pressure drop of less than or equal to eight (8) inches of water. The Technical Specification bases for specification 3.17.B.2.b(1) on page 229z incorrectly states that a pressure I drop of less than eight (8) inches of water is indicative that the filters and adsorbers are not clogged by excessive amounts of foreign matter. The bases 4 l on page 229z are revised to state that a pressure drop of less than or equal to eight (8) inches of water is indicative of acceptable system performance, j A-6 t
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111. Safety Assessment of the Proposed Changes A. Coolant Chemistry Equivalent Radiciodine Concentration l 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. I 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 0.25 Ci of lodine-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 a postulated line break in the Reactor Water Cleanup (RWCU) system.
Background
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 event 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 radioactive material to the environment due to a postulated high energy line break outside of the primary containment up to and including a design basis Main Steam Line Break Accident, radiation doses are maintained within the guidelines of 10 CFR 100. 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 l 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 operates well within the current and proposed technical specification limits for reactor coolant concentration of dose equivalent lodine-131. Data from the last eight operating cycles show that the reactor coolant concentration of dose equivalent lodine-l 131 has been a small fraction of the Technical Specification limit with a cycle average of l 4.5 X 10' Ci/gm over the last eight operating cycles. Even though the reactor core may l contain no defective fuel, trace amounts of natural uranium in core construction materials l
and zircaloy cladding, as well as traces of enriched uranium on the external cladding surface, could be a source of fission products in the coolant during power operation.
The basis for the current Technical Specification limit of 5 microcuries per gram of dose equivalent lodine-131 is derived from the analysis of the Main Steam Line Break accident performed by the Atomic Energy 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, A-7 1
l 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 l analysis of the Main Steam Line Break Accident.
1 Analytical Basis for Proposed Changes to Dose Equivalent I-131 l The Main Steam Line Break Accident radiological consequences have been reanalped l as part of the power rerate effort. This analysis was performed using inputs that are consistent with the current regulatory guidance. The analysis used an input for the dose equivalent reactor coolant lodine-131 concentration of 2 Ci/gm. The operating dose equivalent reactor coolant lodine-131 concentration at Monticello is normally well below 2 Ci/gm, and this analysis input is conservative with respect to actual operating values.
l The results of the MSLB analysis are provided in Exhibit D. The analysis demonstrated that the dose consequences from a postulated design basis Main Steam Line Break outside of primary containment are well within the guidelines of 10 CFR 100 and 10 CFR 50 Appendix A, General Design Criterion 19. The Technical Specification bases are revised to delete bases information which is not pertinent to the specification and to reflect the revised accident analysis. The proposed change to the Technical Specification to limit the dose equivalent reactor coolant lodine-131 concentration to 0.25 Ci/gm is conservative with respect to the value of 2 Ci/gm dose equivalent reactor coolant lodine-131 used for the evaluation of the postulated design basis Main Steam Line Break Accident.
An evaluation of the radiological consequences of the postulated limiting RWCU line break was completed. The release was assumed to occur at ground level without SGTS filtration. The evaluation credits operator action to isolate the postulated RWCU line break, by closure of the RWCU containment isolation valves, ten (10) minutes after initiation of the postulated break. The motor operated containment isolation valves, which are remotely controlled by the control room operator, are credited as having a 29 second closure time. This assumption is conservative because a safety related RWCU containment automatic isolation was installed during the 1998 refueling outage which would effect an isolation within one minute of the break. The break flow is assumed as linearly proportional to the valve closure upon initiation of valve closure. This evaluation used a source term based on a reactor coolant dose equivalent radioiodine concentration of 0.25 Ci/gm. This source term is consistent with the change to Technical Specification 3.6.C.1 proposed by this License Amendment request. The results of this evaluation are provided in Exhibit E.
Monticello has evaluated the thermal-hydraulic effects of the postulated RWCU system line break. The effect on Environmental Qualification of equipment was reviewed and it was concluded that the ability to safely shutdow1 the plant is maintained with the limiting line break. The internal flooding from the break was evaluated and no adverse effects were identified. The effect on the integrity of structures was evaluated, and it was found that the break had no adverse effect on the primary containment or Reactor Building structures. The effect on the integrity of the RWCU Room structure was evaluated, and A-8
it was determined that block walls associated with the room may fail, but safe shutdown equipment needed to mitigate the consequences of the event would not be impacted by debris. Motor operated valve operability was examined, and all affected valves were confirmed to be capable of performing their design basis functions.
Coolant Chemistry Bases Changes The bases for Sections 3.6/4.C are being changed to reflect the analyses for the MSLB and
, the RWCU break as described above.
Summary The proposed changes to the plant Technical Specifications and Bases are consistent with the analyses performed of the Main Steam Line Break Accident and Reactor Water Cleanup line break 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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l 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 HABITABILITY, l Specification 3.17.B.3, Post Maintenance Testing Requirements, page 229x.
l Technical Specification Section 3.17, Section A, Control Room Ventilation System, Bases, page 229y. Technical Specification Section 3.17, Section B, Control Room ,
Emergency Filtration System, Bases, page 229y. Technical Specification Section i 4.17, Section B, Control Room Eriergency Filtration System, Bases, page 229z. I l
Changes are proposed to the plant Technical Specifications and bases to modify the limiting conditions for operation and surveillance requirements for the control room j 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.
System Function and Design Basis The function of Control Room Ventilation-Emergency Filtration Train (CRV-EFT) system is to maintain the environment of the Main Control Room, thereby ensuring its 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 provide the control room and the first and second floors of the EFT building with j conditioned air to maintain acceptable temperature conditions during normal operation.
The EFT subsystem provides for manual isolation of the control room and the first and second floors of the EFT building from outside air during a toxic chemical release or an accident where high levels of activity may be released. During a radiological accident, the EFT provides for immediate automatic pressurization of the Control Room with l
filtered air to minimize the activity, and therefore the radiological dose., inside the control j room. The redundant air filtration units consist of the following components in series: a l 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, and the HEPA filters remove particulate matter.
- The EFT system is designed to satisfy the criteria of NUREG 0737, Section Ill.D.3.4 l which imposes the criteria of 10CFR50 Appendix A, GDC 19. The EFT system 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 A-10 i
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i of the accident. Section 6.7 of the MNGP USAR provides additionalinformation on the i CRV-EFT system. l Technical Specification Changes
- The Main Steam Line Break Accident radiological consequences have been re-analyzed.
l The inputs used in this analysis provide a conservative assessment of the potential L radiological consequences. The results of the analysis are provided in Exhibit D. The l radiological analyses reflect an improvement for the control room emergency filtration l system filter efficiency and reduced control room ventilation bypass leakage. 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. The proposed changes to the limiting conditions for operation and surveillance requirements for the EFT system are more restrictive than the current requirements. These changes provide L assurance that the EFT system removal efficiency will exceed the assumed removal efficiency in the analysis for dose consequences of postulated events.
l The proposed changes to performance criteria and post maintenance testing criteria l specified in technical specification requirements 3.17.B.2.a(1),3.17.B.2.a(2),3.17.B.3.a, and 3.17.B.3.b establish in-place testing criteria consistent with regulatory guidance. For '
in-place penetration testing of the HEPA filters with dioctyl phthalate (DOP), the HEPA filters upstream and downstream of the charcoal adsorbers are tested individually and in combination to satisfy the proposed criteria for Technical Specifications.
The surveillance requirements of Section 4.17.B.2 have been revised to include an I explicit reference to the testing criteria of Section 10, HEPA Filter Bank In-Place Test, L and Section 11, Adsorber Bank in-Place Test, of ASME N510-1989, Testing of Nuclear Air Treatment Systems and to ASTM D 3803-89, Standard Test Method for Nuclear- ,
Grade Activated Carbon. Because of the design vintage and configuration of the EFT
!. system, certain exceptions from the requirements to ASME N510-1989 are requested.
These exceptions are documented in Exhibit F of this submittal. The exceptions will be included in Section 6.7 of the USAR during the next regularly scheduled update.
A performance criterion of less than or equal to 0.4% methyl iodide penetration for l 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 FFT 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 ASTM D 3803-89. The test conditions are proposed to be revised to 30 C and 95% relative humidity co,1sistent with ASTM D 3803-
- 89. ' Guidance provided in ASTM D 3803-89 states that the 30 C,95% relative humidity methyl iodide test is the most reliable test method to establish iodide removal efficiency
- l. of any adsorbent.
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l Control Room Habitability Bases Changes The bases are revised to establish consistency between the Technical Specification bases and Technical Specification 3.17.B.2.b(1). Technical Specification 3.17.B.2.b(1) states that the Control Room Emergency Filtration system shall be show to be operable !
with a combined filter pressure drop ofless than or equal to eight (8) inches of water. By l letter dated July 5,1988, with subject " Revision 4 to License Amendment Request Dated April 3,1984," NSP responded to an NRC staff question regarding the appropriate criterion for specification 3.17.B.2.b.1 for the combined pressure drop across the HEPA filters and charcoal adsorber. NSP stated that the system was designed to have a pressure drop of eight (8) inches of water. Amendment 65 was issued May 30,1989 and specified an acceptance criterion of less than or equal to eight (8) inches of water for the combined filter pressure drop; however, the bases for the specification incorrectly ;
discusses the acceptable value for the pressure drop in that it omits that values equal to 8 inches are acceptable. A review of design basis information concerning the Emergency Filtration Train confirms that the appropriate acceptance criterion for the combined pressure drop across the HEPA filters and charcoal adsorbers is less than or equal to eight (8) inches of water. The bases on page 229z are revised to state that a !
pressure drop of less than or equal to eight (8) inches of water is indicative of acceptable system performance.
The bases for sections 3.17 and 4.17.B are revised to reflect the performance of in-place filter testing for the Emergency Filtration Train system in accordance with the above standards. In addition, a modification to the control room emergency filtration system
' has been performed to establish the reduced control room ventilation bypass leakage.
, With this modification, pressurization of the control room during normal system operation I- will not be a system function; however, filtered pressurizing air will be supplied to the main control room upon initiation of the EFT system. The modification enhances the isolation of the envelope supplied by the EFT system. This modification minimizes the ;
potential for the introduction of non-filtered air during short term puff releases. The lack of normal pressurization for the control room does not have an adverse effect on control :
room habitability. The bases for sections 3.17 are being changed to reflect this modification. i 1
l Summary of Control Room Habitability Changes The proposed changes to the Technical Specifications are provided to ensure testing is ;
l-performed which provides a high level of assurance of the capability of the EFT system to perform as analyzed in the evaluation of the control room operator doses resulting from a postulated Main Steam Line Break. The analysis determined that control room operator doses remain below the regulatory guidelines of 10CFR50 Appendix A, GDC
- 19. The proposed changes to the Technical Specifications ensure continued compliance with NUREG-0737, item Ill.D.3.4 which requires that nuclear power plants be equipped with a control room from which actions can be taken to operate the plant 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 analyticalinputs used in the safety analyses, thus the proposed changes are acceptable.
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The proposed changes do not result in a significant increase in the probability or consequences of oostulated accidents previously analyzed, an accident not previously analyzed, or in a significant reduction in the margin of safety.
IV. Determination of No Significant Hazards Consideration The pmposed change to the Operating License has been evaluated to determine whether it constiQtes a significant hazards consideration as required by 10 CFR 50.91 using standards provided in 10 CFR 50.92. The analysis demonstrating no significant hazards consideration 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 radiciodine concentration in the reactor coolant to ensure that in the event of a release of radioactive material to the environment due to a postulated high energy line break up to and including a design basis Main Steam Line Break Accident, radiation doses are maintained well within the regulatory guidelines. 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 and postulated high energy line breaks. 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 radioiodine concentration. This value is conservative with respect to the value used in the Main Steam Line Break dose consequences analysis and is consistent with the dose consequences evaluation of a postulated Reactor Water Cleanup (RWCU)line break. Changes are proposed to the limiting conditions for operation and surveillance requirements for the Control Room Emergency Filtration Train iodine removal efficiency. These changes are consistent with the inputs used in the analysis of the radiological consequences of the postulated RWCU line break and the Main Steam Line Break Accident. Changes to testing requirements are more restrictive and in accordance with the applicable regulatory guidance. These proposed requirements maintain operating restrictions for analytical inputs used in the analysis of the Main Steam Line Break Accident. Evaluation of these events has demonstrated that the postulated radiological consequences will also remain within the licensing basis established in the AEC Provisional Operating License Safety Evaluation Report, dated March 18,1970, thus the proposed changes do not result in an increase in the consequences of previously evaluated accidents.
The analysis of the Main Steam Line Break Accident performed using a reactor coolant radiciodine concentration of 2 pCi/gm dose equivalent lodine-131 and a control room ventilation filter efficiency consistent with the proposed Technical Specifications changes demonstrated that radiological consequences of the Main Steam Line Break are not changed significantly. The radiological consequences of the Main Steam Line Break A-13
l Accident remain within the exposure guidelines of 10CFR100 and 10CFR50 Appendix A, l General Design Criterion 19. The offsite dose consequences remain bounded by the original 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 1 remain bounded by the dose consequences of the comparable design basis loss of I coolant accident. l l
l The evaluation of the postulated RWCU line break, performed using a reactor coolant radiciodine concentration of 0.25 pCi/gm dose equivalent lodine-131 and a control room ventilation filter efficiency consistent with the proposed Technical Specifications changes, demonstrated that the radiological consequences of this event remain within the exposure guidelines of 10CFR100 and 10CFR50 Appendix A, General Design Criterion 19. The offsite dose consequences remain bounded by the Main Steam Line l' Break as established in the licensing basis provided in the AEC Provisional Operating 1
License Safety Evaluation Report, dated March 18,1970.
l The proposed Technical Specification changes do not introduce new equipment operating modes, nor do the proposed changes alter existing system relationships. The l
proposed changes do not introduce new failure modes. 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 l in the probability of an accident previously evaluated for the Monticello plant. '
B. The proposed amendment will not create the possibility of a new or different kind j of accident from any accident previously analyzed. !
l The proposed Technical Specification changes do not introduce new equipment !
operating modes, nor do the proposed changes alter existing system relationships. j Operator action to mitigate the consequences of the postulated RWCU line break is conservative based on the simple action required by the operator to close the containment isolation valves within 10 minutes. Isolation at 10 minutes is very conservative since a safety related RWCU containment isolation system that was installed during the 1998 refueling outage would effect an automatic isolation within one I
minute of the RWCU break.
l The proposed change to the specification for reactor coolant dose equivalent radiciodine is conservative with respect to the re-evaluation of the Main Steam Line Break Accident for the more conservative hot standby initial condition for the postulated accident. The proposed change to the specification for reactor coolant dose equivalent radiciodine is consistent with the postulated high energy line break of a Reactor Water Cleanup line.
l The proposed changes to the limiting conditions for operation and surveillance l requirements for the control room emergency filtration train iodine removal efficiency are consistent with the inputs used in the evaluation of the radiological consequences of the postulated RWCU line break and the Main Steam Line Break Accident. The system improvements to reduce bypass leakage during postulated accidents do not have an A-14 l
l I
L 1
l~
l l l adverse effect on control room habitability. Therefore, the proposed amendment will not l create the possibility of a new or different kind of accident. )
l C. The proposed amendment will not involve a significant reduction in the marg!n of )
safety.
j i
Surveillance data has demonstrated the proposed requirements are within the current j 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 bounding postulated high energy line break of a Reactor Water Cleanup line and the Main Steam Line Break Accident. The proposed change to the specification i for reactor coolant dose equivalent radioiodine is conservative with respect to the re-l evaluation of the Main Steam Line Break Accident for the more conservative hot standby l initial condition for the postulated accident. The proposed change to the specification for l reactor coolant dose equivalent radioiodine is consistent with the postulated high energy i line break of a Reactor Water Cleanup line. The evaluation of these postulated events I determined that the radiological consequences remain within the exposure guidelines of l' 10CFR100 and of 10CFR50 Appendix A, General Design Criterion 19 and within the originallicensing basis contained in the Provisional Operating License. The proposed l changes to the limiting conditions for operation and surveillance requirements for the control room emergency filtration train iodine removal efficiency provide assurance that the system will perform at the filter efficiency as used in the evaluation of the radiological consequences of the postulated events. Therefore, the proposed amendment will not involve a significant reduction in the margin of safety.
V. Environmental Assessment Northern States Power Company has evaluated the proposed changes and determined that:
.1. ' The change does not involve a significant hazards consideration.
l 2 The changes do not involve a significant change in the type or significant increase in the l 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, since 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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