Text

License Amendment Request (LAR) to Revise the Loss of Coolant Accident (LOCA) and Main Steam Line Break (MSLB) Accident Dose Consequences Analyses and Affected Technical Specifications (TS)
	ML081790439
Person / Time
Site:	Prairie Island
Issue date:	06/26/2008
From:	Wadley M; Nuclear Management Co
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	L-PI-08-052
	Download: ML081790439 (97)
	v • d • e

Prairie Island Nuclear Generating Plant Operated by Nuclear Management Company, LLC L-PI-08-052 10 CFR 50.90 U S Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Prairie Island Nuclear Generating Plant Units 1 and 2 Dockets 50-282 and 50-306 License Nos. DPR-42 and DPR-60 License Amendment Request (LAR) to Revise the Loss of Coolant Accident (LOCA) and Main Steam Line Break (MSLB) Accident Dose Consequences Analvses and Affected Technical Specifications (TS)

Pursuant to the requirements of 10 CFR 50.90 and 10 CFR 50.59(~)(2), the Nuclear Management Company, LLC (NMC) requests Nuclear Regulatory Commission (NRC) review and approval of proposed licensing basis and TS changes. The proposed changes would amend the Facility Operating Licenses by revising the licensing basis LOCA and MSLB accident radiological dose consequences for PlNGP as currently described in the Updated Safety Analysis Report (USAR) Section 14.5 and Section 14.9. This LAR also proposes concomitant amendments to Appendix A of the Facility Operating Licenses, Technical Specifications (TS) 3.3.5, "Containment Ventilation Isolation Instrumentation", 3.4.17, "RCS Specific Activity", and 3.6.3, "Containment Isolation Valves", which are necessary to implement the proposed revised analyses.

NMC has evaluated the proposed changes in accordance with 10 CFR 50.92 and concluded that they involve no significant hazards consideration. to this letter contains the licensee's evaluation of the proposed changes. provides the supporting licensing report.

NMC requests approval of this LAR within one calendar year of the submittal date.

Upon NRC approval, NMC requests 90 days to implement the associated changes. In accordance with 10 CFR 50.91, NMC is notifying the State of Minnesota of this LAR by transmitting a copy of this letter and enclosures to the designated State Official.

If there are any questions or if additional information is needed, please contact Mr. Dale Vincent, P.E., at 651-388-1 121.

171 7 Wakonade Drive East Welch, Minnesota 55089-9642 Telephone: 651.388.1 121

Document Control Desk Page 2 Summary of Commitments This letter contains no new commitments and no revisions to existing commitments.

I declare under penalty of perjury that the foregoing is true and correct.

Executed On JUN 2 6 2008 Michael D. Wadley v

Site Vice President, Prairie Island Nuclear Generating Plant, Units 1 and 2 Nuclear Management Company, LLC

Enclosures:

(2) cc:

Administrator, Region Ill, USNRC Project Manager, Prairie Island, USNRC Resident Inspector, Prairie Island, USNRC State of Minnesota

ENCLOSURE Evaluation of the Proposed Changes License Amendment Request (LAR) to Revise the Loss of Coolant Accident (LOCA) and Main Steam Line Break (MSLB) Accident Dose Consequences Analyses and Affected Technical Specifications (TS)

1.

SUMMARY

DESCRIPTION

2.

DETAILED DESCRIPTION 2.1 Proposed Changes

2.2 Background

3.

TECHNICAL EVALUATION

4.

REGULATORY SAFEN ANALYSIS 4.1 Applicable Regulatory Requirementslcriteria 4.2 Precedent 4.3 Significant Hazards Consideration 4.4 Conclusions

5.

ENVIRONMENTAL CONSIDERATION

6.

REFERENCES ATTACHMENTS:

1.

Technical Specification Pages (Markup)

2.

Bases Pages (Markup) (For information only)

3.

Technical Specification Pages (Retyped)

Page 1 of 21

Enclosure Dose Analyses NMC

1.

SUMMARY

DESCRIPTION This LAR is a request to amend Operating Licenses DPR-42 and DPR-60 for the Prairie Island Nuclear Generating Plant (PINGP), Units 1 and 2.

Pursuant to the requirements of 10 CFR 50.90 and 10 CFR 50.59(~)(2), the Nuclear Management Company, LLC (NMC) requests Nuclear Regulatory Commission (NRC) review and approval of proposed licensing basis and TS changes. The proposed changes would amend the Facility Operating Licenses by revising the licensing basis LOCA and MSLB accident radiological dose consequences for PINGP as currently described in the Updated Safety Analysis Report (USAR) Section 14.5 and Section 14.9. This LAR also proposes concomitant amendments to Appendix A of the Facility Operating Licenses, Technical Specifications (TS) 3.3.5, "Containment Ventilation Isolation Instrumentation", 3.4.17, "RCS [Reactor Coolant System] Specific Activity",

and 3.6.3, "Containment Isolation Valves1', which are necessary to implement the proposed revised analyses. These licensing basis and TS changes assure that the plant analyses results meet the applicable regulatory limits and guidance, and the plant is operated in a safe manner.

2.

DETAILED DESCRIPTION 2.1 Proposed Changes A brief description of the associated proposed TS changes is provided below along with a discussion of the justification for each change. The specific wording changes to the TS are provided in Attachments 1 and 3 to this enclosure.

TS 3.3.5, "Containment Ventilation Isolation Instrumentation": This TS and associated Bases will be removed from the Technical Specifications. This TS provides requirements for instrumentation which isolates the Containment Inservice Purge System (CIPS) during Modes 1, 2, 3, and 4, the plant operating Modes. This change is acceptable because this system has not been operated in these Modes in many years and NMC does not intend to operate this system in these Modes in the future. The system will be blind flanged during the plant operating Modes and valve isolation will not be required.

TS 3.4.17, "RCS Specific Activity", Surveillance Requirement (SR) 3.4.17.2 and Figure 3.4.17-1: This LAR proposes to revise the action limit in Condition A and the acceptance criteria limit in SR 3.4.17.2 from I

.O pCi/gm dose equivalent 1-131 (DEI) to 0.5 pCi/gm. Also in Figure 3.4.17-1, the Acceptable Operation DEI limit is reduced from 60 pCi/gm to 30 pCi/gm for plant operations 80 to 100% of rated thermal power (RTP). These changes are acceptable because they assure that the control room dose consequences from an MSLB accident are within the applicable regulatory limits.

Page 2 of 21

Enclosure Dose Analyses NMC TS 3.6.3, "Containment Isolation Valves": Changes to this TS are proposed which will require ClPS to be blind flanged during the plant operating Modes and remove ClPS operational requirements. The changes include an addition to the Actions Notes prohibiting intermittent unisolation of the ClPS penetrations, removal of Condition D requirements for valve leakage limits not met, revision of Surveillance Requirements (SR) to require blind flange installation prior to entering the plant operating Modes and deletion of the SR requiring leakage rate testing for the ClPS containment isolation valves. These changes are acceptable because they assure that the ClPS containment penetrations are blind flanged during plant operating Modes and containment integrity is maintained in accordance with the accident analyses.

Although Bases changes are not a part of this LAR, Attachment 2 to this enclosure includes marked up Bases pages for information. The changes proposed in are directly related to the LOCA and MSLB accident licensing basis changes and the proposed revisions to TS 3.3.5, 3.4.17, and 3.6.3.

In summary these changes are acceptable because they assure the plant analyses and operations meet the applicable regulatory requirements.

2.2 Background

The current PINGP licensing basis for radiological dose consequence analyses of accidents discussed in Chapter 14 of the USAR is based upon methodologies and assumptions that are primarily derived from Technical Information Document (TID)

-14844 and other early industry guidance.

An effort to validate the current design and licensing basis of the radiological accident analysis was initiated in support of future facility changes at PINGP. This validation effort revealed several conditions adverse to quality with respect to the current licensing basis LOCA and MSLB accident radiological analyses.

Three dose significant issues were identified with the LOCA radiological analysis: 1) a nonconservative modeling factor was used to address the activity buildup rate in the control room prior to reaching equilibrium concentrations; 2) the shield building elemental iodine filter efficiency assumed in the analysis did not incorporate the safety factor of two discussed in Regulatory Guide 1.52, Revision 2; and 3) the control room analysis described in the USAR was not based on the limiting unfiltered in-leakage determined from tracer gas testing.

The most dose significant issue identified with the MSLB radiological analysis was nonconservatism in the equilibrium iodine appearance rate which was identified in Westinghouse Nuclear Safety Advisory Letter (NSAL)00-004 (Reference 1).

Page 3 of 21

Enclosure Dose Analyses NMC The issues identified by the validation effort were entered into the PlNGP corrective action program (CAP). Operability was evaluated and documented in accordance with the PlNGP operability determination process. No systems were declared inoperable; however, the RCS specific activity level was administratively limited to a specific activity lower then allowed by TS to ensure that control room and offsite dose acceptance criteria are met.

The dose consequence analyses for the LOCA and MSLB accident have been re-evaluated in accordance with the current PlNGP licensing basis methodology while addressing the issues identified in the validation effort. As discussed below in Section 3.0, "Technical Evaluation", the LOCA accident re-evaluation dose consequence results, for control room iodine doses, are more than a minimal increase in the consequences of an accident previously evaluated in the USAR. Therefore, under the requirements of Title 10 Code of Federal Regulations Section 50.59 (10 CFR 50.59),

NRC review and approval is required prior to incorporation into the plant licensing basis.

This LAR proposes to revise the LOCA and MSLB accident dose consequence analyses presented in USAR Chapter 14, Sections 14.5 and 14.9, remove the provisions in TS 3.3.5 and 3.6.3 which allow the ClPS to operated during Modes 1, 2, 3, and 4, and revise the DEI limits in TS 3.4.17.

With the TS changes proposed in this LAR the plant will continue to operate safely and the health and welfare of the public is protected.

3.

TECHNICAL EVALUATION PlNGP is a two unit plant located on the right bank of the Mississippi River approximately 6 miles northwest of the city of Red Wing, Minnesota. The facility is owned by the Northern States Power Company (NSP) and operated by NMC. Each unit at PlNGP employs a two-loop pressurized water reactor designed and supplied by Westinghouse Electric Corporation. The initial PlNGP application for a Construction Permit and Operating License was submitted to the Atomic Energy Commission (AEC) in April 1967. The Final Safety Analysis Report (FSAR) was submitted for application of an Operating License in January 1971. Unit 1 began commercial operation in December 1973 and Unit 2 began commercial operation in December 1974.

The PlNGP was designed and constructed to comply with NSP's understanding of the intent of the AEC General Design Criteria (GDC) for Nuclear Power Plant Construction Permits, as proposed on July 10, 1967. PlNGP was not licensed to NUREG-0800, "Standard Review Plan (SRP)."

System Descriptions The radiological dose consequence analyses evaluate the post accident doses at the exclusion area boundary (EAB), low population zone (LPZ) and control room (CR) based on plant structure shielding and holdup, mitigation equipment performance and meteorological data. Plant building and mitigation equipment background information is Page 4 of 21

Enclosure Dose Analyses NMC provided to facilitate NRC review of this submittal. The relative locations of the site buildings and control room are shown in USAR Figures 1.I-3, and 1.I-6.

Containment The containment is a free standing steel pressure vessel surrounded by a reinforced concrete shield building. The containment vessel, including all its penetrations, is a low leakage steel shell designed to contain radioactive material that may be released from the reactor core following a design basis LOCA. Additionally, the containment and shield building provide shielding from the fission products that may be present in the containment atmosphere following accident conditions.

The containment vessel is a vertical cylindrical steel pressure vessel with a hemispherical dome and ellipsoidal bottom, completely enclosed by a reinforced concrete shield building. A 5 ft wide annular space exists between the walls of the steel containment vessel and the concrete shield building and 7 ft clearance exists between the roofs of the containment vessel and shield building to permit inservice inspection and collection of containment outleakage.

The shield building provides shielding and allows controlled release of the annulus atmosphere under accident conditions. The inner steel containment and its penetrations establish the leakage limiting boundary of the containment. Maintaining the containment operable limits the leakage of fission product radioactivity from the containment to the environment.

Shield Building Ventilation System (SBVS)

The SBVS ensures that radioactive materials that leak from the primary containment into the shield building (secondary containment) following a design basis accident (DBA) are filtered and adsorbed prior to exhausting to the environment.

The SBVS establishes a negative pressure in the annulus between the shield building and the steel containment vessel following a DBA. Filters in the system then control the release of radioactive contaminants to the environment.

The SBVS consists of two separate and redundant trains. Each train includes a heater, a prefilter, moisture separators, a high efficiency particulate air (HEPA) filter, an activated charcoal adsorber section for removal of radioiodines, a recirculation fan and an exhaust fan. The ventilation system for each shield building includes a vent stack which penetrates the shield building dome and discharges to the atmosphere. The HEPA filter and the charcoal adsorber section are credited in the analysis. The system initiates and maintains a negative air pressure in the shield building by means of filtered exhaust ventilation of the shield building following receipt of a safety injection (SI) signal.

Page 5 of 21

Enclosure Dose Analyses NMC Auxiliary Building Special Ventilation System (ABSVS)

The ABSVS is a standby ventilation system, common to the two units, that is designed to collect and filter air from the Auxiliary Building Special Ventilation (ABSV) boundary following a LOCA. The ABSV boundary contains those areas within the auxiliary building which have the potential for collecting significant containment leakage that could bypass the shield building and leakage from systems which could recirculate primary coolant during LOCA mitigation such as the residual heat removal (RHR) system.

The ABSVS consists of two independent and redundant trains. Each train consists of a heater, a prefilter, a HEPA filter, an activated charcoal adsorber section for removal of gaseous activity (principally iodines), and a fan. The system initiates filtered ventilation of the ABSV boundary following receipt of a safety injection (SI) signal, high radiation signal or manual initiation. The radiation signal is not credited for accident mitigation.

Control Room Special Ventilation System (CRSVS)

The control room envelope consists of the control room and the two chiller rooms. The control room is a common structure that contains the controls for both Unit 1 and Unit 2.

The control room is located at elevation 735 ft. within the Auxiliary Building approximately equidistance between Unit 1 and Unit 2. The chiller rooms are located directly above the control room at elevation 755 ft. The control room ventilation system is entirely located within the two chiller rooms (one train of ventilation system in each room), with the exception of the outside air supply. The outside air supply ducting is routed through the Auxiliary Building. The outside air supply dampers are located at the envelope boundary. There are no other ventilation systems that penetrate the control room envelope.

The CRSVS provides a protected environment from which operators can control the unit following an uncontrolled release of radioactivity.

The CRSVS consists of two independent, redundant trains that recirculate and filter the control room air. Each train consists of an air handling unit, a prefilter, a HEPA filter, an activated charcoal adsorber section for removal of gaseous activity (principally iodines),

and a cleanup fan. The CRSVS is an emergency system, parts of which may also operate during normal unit operation.

Upon receipt of the actuating signal(s), normal air supply to the control room is isolated, and the stream of ventilation air is recirculated through the system filter trains. The prefilters remove any large particles in the air, and any entrained water droplets present, to prevent excessive loading of the HEPA filters and charcoal adsorbers.

Actuation of the CRSVS is initiated by:

a.

High radiation in the control room ventilation duct; or Page 6 of 21

Enclosure Dose Analyses NMC

b.

Safety injection signal.

Actuation of the system closes the unfiltered outside air intake and unfiltered exhaust dampers, and aligns the system for recirculation of a portion of the control room air through the redundant trains of HEPA and the charcoal filters. The operating condition initiates filtered ventilation of the air supply to the control room.

Containment lnservice Purge System (CIPS)

The CIPS refers to that portion of the Spent Fuel Special and Containment lnservice Purge system that provides the containment air cleanup function. The CIPS operates to: a) reduce the concentration of noble gases within containment prior to and during personnel access; and b) provide low volume normal purge and ventilation. Two containment automatic isolation valves and an automatic shield building ventilation damper are provided on each supply and exhaust line. The system includes charcoal filters which are shared with the SFPSVS and exhausts to the Shield Building Vent Stack.

Containment ventilation isolation (CVI) instrumentation closes the containment isolation valves in the CIPS. This action isolates the containment atmosphere from the environment to minimize releases of radioactivity in the event of an accident. The CIPS may be in use during reactor operation (Modes 1, 2, 3, and 4) and with the reactor shutdown except during handling of recently irradiated fuel (i.e., fuel that has occupied part of a critical reactor core within the previous 50 hours5.787037e-4 days 0.0139 hours 8.267196e-5 weeks 1.9025e-5 months ). Containment ventilation isolation initiates on an SI signal, by manual actuation of containment isolation, or by manual actuation of containment spray.

The supply and exhaust lines are designed to have blind flanges installed where the lines pass through the shield building annulus. Normally, during MODES 1, 2, 3, and 4 the blind flanges provide the containment penetration isolation function. If this system was to be used for ventilation of containment in MODES 1, 2, 3, and 4, the valves would be leak tested, and the blind flanges removed and replaced with a spool piece.

Prior to system use, the automatic isolation valves and dampers would be verified to be operable and a debris screen installed on each line preventing foreign material from inhibiting the proper closing of the valves. When purge of containment is completed and CIPS operation is no longer required, the system would be returned to its normal operating configuration with the spool pieces removed and the blind flanges installed.

LOCA and MSLB Accident Radiological Analyses Changes and Bases for Changes The LOCA and MSLB accident radiological dose consequence analyses have been revised to address nonconservatisms which are discussed in detail in Enclosure 2, Section 2, "Regulatory Approach".

As discussed in Enclosure 2, Section 3, "Computer Codes", the new analyses were Page 7 of 21

Enclosure Dose Analyses NMC performed using an industry computer code, ARCON96, and Shaw Stone & Webster (SS&W) computer codes PERC2 and SW-QADCGGP (see Enclosure 2 for a complete citation to these codes). The NRC previously approved ARCON96 for use at PINGP in license amendments 166 and 156, for Units 1 and 2 respectively, in a Safety Evaluation (SE) dated September 10,2004, Accession Number ML042430504. The NRC has previously reviewed and approved license amendments utilizing the computer codes PERC2 and SW-QADCGGP to support their analyses. Two example license amendments are cited as References 2 and 3.

In addition to addressing nonconservatisms, these analyses have also been performed at 102% of the plant rated thermal power and use fuel source terms which bound new types of fuel which NMC plans to use at PINGP. These considerations are discussed in, Section 4, "Radiation Source Terms".

The revised LOCA and MSLB accident radiological dose consequences analyses inputs, assumptions and methods are provided in Enclosure 2, Sections 2 though 7 The LOCA and MSLB accident analyses were based on the TS changes discussed below.

The results of these revised LOCA and MSLB accident analyses are provided in, Section 8, "Summary of Results and Conclusions". These analyses demonstrate that the doses are within the applicable regulatory limits and guidance, assuming adoption of the proposed TS changes. Tables 1, 2 and 3, below, provide a comparison of the current licensing basis dose consequences to the proposed reanalysis dose consequences. For the MSLB accident, the proposed dose consequences are less than the current licensing basis assuming the RCS specific activity limits for DEI are reduced as proposed in this LAR. However, for the LOCA reanalysis, the control room thyroid dose resulted in more than a minimal increase in the consequences and, under the criterion of 10 CFR 50.59(c)(2)(iii), a license amendment is required from the NRC prior to implementation; thus, this license amendment request has been submitted for NRC review and approval.

Table 1 MSLB Accident - Pre-accident Initiated Spike Page 8 of 21 Receptor Location EAB LPZ CR Reanalysis Thyroid (rem) 3.9 3.3 6.9 Current Licensing Basis Whole Body (rem)

~ 0. 0 1

~ 0. 0 1 Thyroid (rem) 8.02 7.62 11.67 Whole Body (rem) 0.027 0.07

~ 0. 0 1

Enclosure Dose Analyses NMC Table 2 MSLB Accident - Accident Initiated Spike Table 3 LOCA Receptor Location EAB LPZ CR Current TS Requirements, Basis and Limitations Reanalysis Receptor Location EAB LPZ CR The current ClPS TS requirements were originally incorporated into PlNGP TS by license amendments 63 and 57, for Units 1 and 2 respectively, issued March 23, 1983 to address Containment Purge Generic Issue B-24 and NUREG-0737 Item ll.E.4.2.

The TS require ClPS to be blind flanged during Modes 1, 2, 3, and 4, except that ClPS may be operated if the CVI instrumentation is operable and the CIPS valves meet applicable containment penetration leakage rate limits. Blind flanges must be installed after each use of ClPS and their seals must meet applicable containment penetration leakage rate limits.

Current Licensing Basis Thyroid (rem) 6.4 22.4 27.2 TS 3.3.5 and the ClPS related provisions in TS 3.6.3 are included in TS to provide operational flexibility to reduce airborne radioactivity levels in containment during plant operation in Modes 1, 2, 3, and 4. This system would most likely be operated during the plant operating Modes if containment entry is required when operating with leaking fuel.

Thyroid (rem) 6.6 23.34 29.78 Whole Body

<O. 1 x0.1

<0.01 TS 3.4.17 Condition A requires the plant to take the Required Actions when the RCS DEI concentration exceeds 1.0 ~Cilgm and SR 3.4.17.2 requires periodic and event based verification during operation in Mode 1 that the RCS DEI is less than 1.0 pCi1gm.

The plant may operate at DEI concentrations above 1.0 ~Cilgm in the "Acceptable Whole Body (rem) 0.027 0.07

<0.01 Current Licensing Basis Page 9 of 21 Reanalysis Skin (rem)

NIA NIA 26.7 Thyroid (rem) 20.5 16.0 28.5

' Thyroid

@em) 18.8 20.5 8.78 Whole body 3.7 3.2 1.6 Whole body (rem) 2.0 2.0 1.5 Skin (rem)

NIA NIA 16.5

Enclosure Dose Analyses NMC Operating" region of Figure 3.4.17-1 for limited periods of time as permitted by TS 3.4.17. The Figure limits are based on limiting the RCS DEI concentration to 60 times the normal limit, which is 60 pCiIgm, when the reactor power is at or above 80% RTP.

Permitting power operation to continue for limited time periods with the primary coolant's specific activity greater than 1.0 pCiIgm DEI, but within the allowable limits shown in Figure 3.4.17-1, accommodates the possible iodine spiking phenomenon which may occur following change in reactor thermal power.

These requirements were originally introduced into the PINGP TS with license amendments 52 and 46 issued December 4, 1981 based on NRC guidance in the standard TS at that time. The limits on the specific activity of the RCS were established to ensure that the resulting 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months doses at the site boundary will not exceed a small fraction of the 10 CFR 100 dose guideline limits following a steam generator tube rupture (SGTR) accident with an existing reactor coolant steam generator (SG) tube leakage rate of 1 gpm. The values for the limits on specific activity represent limits based upon a parametric evaluation by the NRC of typical site locations. These values were considered conservative in that specific site parameters of PINGP, such as site boundary location and meteorological conditions, were not considered in this evaluation. However, the analyses presented above in this LAR demonstrate these limits were nonconservative in that the supporting analyses did not account for the equilibrium iodine appearance rate as identified in Reference 1.

Proposed Changes This LAR proposes to revise the TS which will require the containment inservice purge system penetrations to be blind flanged at all times during plant operating Modes 1, 2, 3, and 4. The blind flanges will fall under the requirements of SR 3.6.1. I and thus be required to meet containment penetration leakage rate limits.

Other requirements in TS 3.6.3 that have also been revised, due to the requirement to install blind flanges on the lines, include an addition to the Actions Notes prohibiting intermittent unisolation of the ClPS penetrations, removal of Condition D requirements for valve leakage limits not met, and deletion of the SR requiring leakage rate testing for the ClPS containment isolation valves.

This LAR proposes to remove TS 3.3.5, which provides requirements for instrumentation to close the ClPS isolation valves, in its entirety.

This LAR proposes to revise TS 3.4.17 Condition A to require taking the Required Actions when the RCS DEI limit exceeds 0.5 pCi/gm and likewise the acceptance criterion in SR 3.4.17.2 is also revised to 0.5 pCi/gm. In keeping with the current Figure 3.4.17-1 "Acceptable Operation" limit of 60 times the normal limit, the figure is revised to limit power operation to 30 pCi/gm when the reactor power is at or above 80% RTP.

Page 1 0 of 2 1

Enclosure Dose Analyses NMC Technical Bases for Changes This LAR proposes to remove ClPS operating requirements from the TS. Normally this system is blind flanged during plant operating Modes; revisions are proposed for TS 3.6.3 which will require the ClPS penetrations to be blind flanged during Modes 1, 2, 3, and 4. The ClPS is not required for support of any plant operating evolutions and its purge and filtration function is not credited in any accident analysis. The system may be operated during plant operating Modes to reduce airborne radioactivity in containment, particularly when there is significant fuel leakage. This system has not been operated during Modes 1,2, 3, and 4 for many years and NMC does not see benefits from operation of the system in these Modes in the future. In the current nuclear plant operating environment, the plant would shutdown to address fuel leakage issues rather than continue to operate and depend on ClPS operation to clean up the containment atmosphere during plant operating Modes. As discussed in Enclosure 2, Section 7.2, the LOCA and MSLB accident analyses were performed assuming the ClPS is isolated and demonstrated that the applicable regulatory requirements and guidance were met.

The containment ventilation isolation instrumentation is credited for isolation of ClPS following an accident in containment during plant operation in Modes 1, 2, 3 and 4.

With blind flanges installed on the system penetrations, TS requirements for penetration isolation instrumentation are not required and TS 3.3.5 can be removed from TS.

ClPS may be operated in Modes 5 and 6, however, the ClPS penetrations will continue to be isolated during movement of recently irradiated fuel in accordance with the provisions of TS 3.9.4, "Containment Penetrations" and the clarifications of the Bases 3.9.4 which states, "In MODE 6, during handling of recent irradiated fuel, the Containment Purge and Containment Inservice Purge systems must remain closed."

As discussed in Enclosure 2, Sections 2.1 and 4.2, the MSLB accident analyses were performed assuming the DEI limits in TS 3.4.17 and SR 3.4.17.2 were reduced as proposed in this LAR. The dose consequence results presented in Enclosure 2, Section 8 demonstrate that when equilibrium iodine appearance rate is properly addressed and the TS limits are reduced, the doses are within the applicable regulatory limits and guidance.

Conclusions LOCA analyses were performed to address nonconservative modeling of the control room activity buildup rate and control room in-leakage, and shield building filter efficiency. These analyses demonstrate that, although doses have increased, they remain below the allowable regulatory limits and guidelines for exclusion area boundary, low population zone and control room doses. This LAR proposes to revise the plant licensing basis to adopt these revised LOCA analyses. An MSLB accident Page 1 1 of 21

Enclosure Dose Analyses NMC analysis was performed to address nonconservative modeling of equilibrium iodine appearance and demonstrated that the regulatory control room dose limits are met when the TS 3.4.17 limits are reduced. Accordingly, this LAR proposes to reduce the TS 3.4.17 action limits and SR acceptance criterion to meet the regulatory control room dose limits. This LAR also proposes to revise the plant licensing basis to adopt the revised MSLB accident analysis. The LOCA and MSLB accident radiological dose consequences analyses assumed that the containment inservice purge valves are closed; accordingly, changes are proposed in TS 3.6.3 which will require ClPS to be blind flanged in Modes 1, 2, 3, and 4 and other TS requirements supporting operation of this system in these Modes will be removed, including TS 3.3.5. Operation, maintenance and testing of the Prairie Island Nuclear Generating Plant with the proposed licensing basis and TS revisions will continue to protect the health and safety of the public.

4.

REGULATORY SAFETY ANALYSIS 4.1 Applicable Regulatorv Requirementslcriteria Title 10 Code of Federal Regulations Section 50.36, "Technical specifications

/I 0 CFR50.36):

(d) Technical specifications will include items in the following categories:

2) Limiting conditions for operation.

(i) Limiting conditions for operation are the lowest functional capability or performance levels of equipment required for safe operation of the facility.

When a limiting condition for operation of a nuclear reactor is not met, the licensee shall shut down the reactor or follow any remedial action permitted by the technical specifications until the condition can be met.

(ii) A technical specification limiting condition for operation of a nuclear reactor must be established for each item meeting one or more of the following criteria:

(6) Criterion 2. A process variable, design feature, or operating restriction that is an initial condition of a design basis accident or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

(C) Criterion 3. A structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a design basis accident or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

Page 12 of 2 1

Enclosure Dose Analyses NMC

3) Surveillance requirements. Surveillance requirements are requirements relating to test, calibration, or inspection to assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the limiting conditions for operation will be met.

The reactor coolant system specific activity is a limiting condition for operation which meets Criterion 2 of 10 CFR 50.36. This license amendment request proposes to reduce the Technical Specification remedial action limit for dose equivalent 1-131 from 1.0 pCi/gm to 0.5 pCi/gm which will require actions sooner, that is, the limits are more restrictive. The Acceptable Operation range in Figure 3.4.17-1 is also reduced from 60

~Cilgm to 30 ~Cilgm for operations at or above 80% of the plant rated thermal power which will require actions sooner, that is, the limits are more restrictive. With these changes, the Technical Specifications will continue to provide limiting conditions for operation with performance levels for safe operation of the facility and appropriate remedial actions.

This license amendment request also proposes to reduce the Surveillance Requirement for reactor coolant system dose equivalent 1-1 31 from 1.0 pCi/gm to 0.5 pCi/gm. With this change, the Technical Specifications provide Surveillance Requirements which assure that the limiting conditions for operation continue to be met.

This license amendment request also proposes revisions to Technical Specification 3.6.3, "Containment Isolation Valves which will require the Containment lnservice Purge System penetrations to be blind flanged during Modes 1, 2, 3 and 4 and the system penetrations shall meet acceptable leakage rate limits to provide containment integrity. These requirements provide the lowest functional capability of equipment required for safe operation of the facility and meet Criterion 3 as equipment that is part of the primary success path to mitigate a design basis accident; thus, the requirements of 10 CFR 50.36 are met. With these changes, the functional capability to isolate the Containment lnservice Purge System valves during plant operation, provided by Technical Specification 3.3.5, "Containment Ventilation Isolation Instrumentation", is not required, does not meet any of the 10 CFR 50.36 Criteria and Technical Specification 3.3.5 can be removed.

Thus with the changes proposed in this license amendment request, the requirements of 10 CFR 50.36 continue to be met.

Title 10 Code of Federal Regulations Section 50.59, "Changes, test and experiments" 110 CFR 50.59) c)(2) A licensee shall obtain a license amendment pursuant to Sec. 50.90 prior to implementing a proposed change, test, or experiment if the change, test, or experiment would:

Page 13 of 21

Enclosure Dose Analyses NMC (iii) Result in more than a minimal increase in the consequences of an accident previously evaluated in the final safety analysis report (as updated);

This license amendment request proposes to incorporate revised loss of coolant accident analyses into the Prairie lsland Nuclear Generating Plant licensing basis which increase dose consequences of this accident as previously evaluated in the Updated

[final] Safety Analysis Report. The analyses demonstrate that the applicable regulatory limits are met; however, the results show that the control room thyroid doses have increased. In accordance with the regulatory requirements of 10 CFR 50.59, these revised dose consequences result in more than a minimal increase in the consequences of an accident previously evaluated in the Final Safety Analysis Report (as updated) and, thus, this license amendment is submitted pursuant to Title 10 Code of Federal Regulations Section 50.90. With the changes proposed in this license amendment request submittal, the requirements of 10 CFR 50.59 are fulfilled.

General Design Criteria The construction of the Prairie lsland Nuclear Generating Plant was significantly complete prior to issuance of 10 CFR 50, Appendix A, General Design Criteria (Appendix A GDC). The Prairie lsland Nuclear Generating Plant was designed and constructed to comply with the Atomic Energy Commission General Design Criteria as proposed on July 10, 1967 (AEC GDC) as described in the plant Updated Safety Analysis Report. AEC GDC proposed criteria 49, 53 and 56 provide guidance applicable to containment integrity which may be affected by the changes proposed in this license amendment request. Furthermore, the Prairie lsland Nuclear Generating Plant is committed to selected Appendix A GDC, including Criterion 19 which provides design guidance for the post accident control room doses. These AEC GDC and Appendix A GDC continue to be met as follows.

AEC GDC Criterion 49 - Containment Design Basis The containment structure, including access openings and penetrations, and any necessary containment heat removal systems shall be designed so that the containment structure can accommodate without exceeding the design leakage rate the pressures and temperatures resulting from the largest credible energy release following a loss-of-coolant accident, including a considerable margin for effects from metal-water or other chemical reactions, that could occur as a consequence of failure of emergency core cooling systems.

AEC GDC Criterion 53 - Containment Isolation Valves Penetrations that require closure for the containment function shall be protected by redundant valving and associated apparatus.

Page 14 of 21

Enclosure Dose Analyses NMC AEC GDC Criterion 56 - Provisions for test in^ of Penetrations Provisions shall be made for testing penetrations which have resilient seals or expansion bellows to permit leaktightness to be demonstrated at design pressure at any time.

Current Technical Specifications allow the Containment lnservice Purge System to be operated in Modes 1, 2, 3, and 4 providing the isolation valves with resilient seals have been tested to meet containment leakage rate limits and the Containment Ventilation Isolation System is also operable to close the system isolation valves in the event of an accident in containment. Current Technical Specifications require installation of blind flanges on the penetrations when the system is not operating and verification the penetrations meet containment leakage rate limits. This license amendment proposes to require the Containment lnservice Purge System penetrations to be blind flanged at all times during plant operation in Modes 1, 2, 3, and 4 with the penetrations tested to meet containment leakage rate limits. The system isolation valves with resilient seals will not be relied upon to provide containment integrity and the Containment Ventilation Isolation System will not be required to close the valves.

Thus with the changes proposed in this license amendment request, the requirements of AEC GDC 49, 53 and 56 continue to be met and the plant Technical Specifications will continue to provide the basis for safe plant operation.

Criterion 19 - Control room A control room shall be provided from which actions can be taken to operate the nuclear power unit safely under normal conditions and to maintain it in a safe condition under accident conditions, including loss-of-coolant accidents.

Adequate radiation protection shall be provided 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. Equipment at appropriate locations outside the control room shall be provided (I) with a design capability for prompt hot shutdown of the reactor, including necessary instrumentation and controls to maintain the unit in a safe condition during hot shutdown, and (2) with a potential capability for subsequent cold shutdown of the reactor through the use of suitable procedures.

The Prairie Island Nuclear Generating Plant loss of coolant accident dose consequence analysis has been revised to address nonconservative modeling elements and update the analyses for new fuel types and include margin for power uncertainty. These analyses demonstrate that control room personnel doses are within the limits of Criterion 19. The plant main steam line break accident analysis was revised to address nonconservative modeling of equilibrium iodine appearance in the reactor coolant system and demonstrated that the control room dose limits in Criterion 19 are met when the reactor coolant system dose equivalent iodine 131 concentration is limited to 0.5 Page 15of21

Enclosure Dose Analyses NMC

~Cilgm. This LAR proposes concomitant changes to revise Technical Specification 3.4.17, "RCS [Reactor Coolant System] Specific Activity", Condition A statement and Surveillance Requirement limit to 0.5 ~Cilgm and revise the Acceptable Operation range in Technical Specification Figure 3.4.17-1 to 30 pCiIgm for operations at or above 80% of the plant rated thermal power.

Thus with the changes proposed in this license amendment request, the requirements of Appendix A GDC Criterion 19 continue to be met and the plant Technical Specifications will continue to provide the basis for safe plant operation.

Title 10 Code of Federal Regulations Section 100.1 1, "Determination of exclusion area, low population zone, and population center distance (1 0 CFR 100)

(a) As an aid in evaluating a proposed site, an applicant should assume a fission produce release from the core, the expected demonstrable leak rate from the containment and the meteorological conditions pertinent to his site to derive an exclusion area, a low population zone and population center distance. For the purpose of this analysis, which shall set forth the basis for the numerical values used, the applicant should determine the following:

(1) An exclusion area of such size that an individual located at any point on its boundary for two hours immediately following onset of the postulated fission product release would not receive a total radiation dose to the whole body in excess of 25 rem or a total radiation dose in excess of 300 rem to the thyroid from iodine exposure.

The loss of coolant accident and main steam line break accident radiological dose consequences have been reanalyzed to address modeling nonconservatisms and update the analyses for new fuel types and provide margin for power uncertainty.

These analyses demonstrate that the dose limits of 10 CFR 100.1 1 are met when the Technical Specification changes proposed in this license amendment request are included.

Thus with the changes proposed in this license amendment request, the requirements of 10 CFR 100.1 1 continue to be met and the plant design basis analyses and Technical Specifications will continue to provide the basis for safe plant operation.

Regulatory Guide 1.4, "Assumptions Used for Evaluating the Potential Radiological Consequences of a Loss of Coolant Accident for Pressurized Water Reactors",

Revision 2 (Regulatory Guide 1.4)

Regulatory Guide 1.4, Revision 2, describes assumptions acceptable to the NRC Staff that may be used in evaluating the radiological consequences of a loss of coolant accident for a pressurized water reactor to show that the offsite dose consequences will be within the regulatory limits of 10 CFR Part 100. This license amendment request presents revised Prairie Island Nuclear Generating Plant loss of coolant accident analyses using the guidance of Regulatory Guide 1.4 which demonstrate that the Page 16 of 21

Enclosure Dose Analyses 10 CFR 100 limits are met when nonconservative modeling elements are addressed.

Thus the guidance of Regulatory Guide 1.4 is met with this license amendment request.

4.2 Precedent In a license amendment request dated January 30,2006 (Reference 4), the Kewaunee Power Station proposed to modify their radiological accident analyses and associated Technical Specifications to compensate for higher control room unfiltered inleakage.

The NRC reviewed and approved the proposed changes in a Safety Evaluation amendment dated March 8, 2007 (Reference 5). This license amendment request is similar to the Reference 4 Kewaunee submittal in that the Prairie Island Nuclear Generating Plant radiological accident analyses have been revised to compensate for analysis nonconservatisms including higher control room unfiltered inleakage. Like in the Kewaunee submittal, the change in dose results for the Prairie Island Nuclear Generating Plant analyses is sufficient to require submittal of the analyses to the NRC for review and approval prior to implementation into the plant design basis. This license amendment request differs from the Kewaunee submittal in that the Prairie Island Nuclear Generating Plant analyses are based on Technical Information Document 14844 source terms and Regulatory Guide 1.4, and meets the regulatory requirements of 10 CFR 100.

4.3 Significant Hazards Consideration The Nuclear Management Company has evaluated whether or not a significant hazards consideration is involved with the proposed amendment by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of amendment," as discussed below:

1.

Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No This license amendment request proposes implementing revised loss of coolant accident and main steam line break accident dose consequence analyses to address modeling nonconservatisms and update the analyses for new fuel types and provide margin for power uncertainty. These analyses assumed that the containment inservice purge system penetrations are isolated, thus this license amendment request proposes Technical Specification revisions which will require these penetrations to be blind flanged during plant operations; these changes allow the Technical Specification requirements for containment ventilation isolation instrumentation to be removed. This license amendment request also proposes associated more restrictive limits in the Technical Specification for reactor coolant system specific activity since the main steam line break accident analysis assumed lower limits.

Page 17 of 21

Enclosure Dose Analyses NMC The accident radiological dose consequences analyses inputs, methodologies and outputs modified by this request are not accident initiators and do not affect the frequency of occurrence of previously analyzed transients. Likewise, the reactor coolant system specific activity limits are not accident initiators and do not affect the frequency of occurrence of previously analyzed transients.

The containment inservice purge system is not an accident initiator and therefore removal of its Technical Specifications does not involve an increase in the probability of an accident. The Technical Specification changes proposed in this license amendment request require the containment inservice purge system to be blind flanged during Modes 1, 2, 3, and 4, therefore removal of the containment ventilation isolation instrumentation Technical Specifications and other Technical Specification system operating requirements does not involve an increase in the consequences of an accident previously evaluated.

The loss of coolant accident and main steam line break accident radiological dose consequences analyses demonstrated the results are within the applicable regulatory limits and guidance using revised inputs, including the proposed lower Technical Specification reactor coolant system specific activity limits, and methodologies. Thus these changes do not involve a significant increase in the consequences of an accident.

Therefore, the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.

Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No This license amendment request proposes implementing revised loss of coolant accident and main steam line break accident dose consequence analyses to address modeling nonconservatisms and update the analyses for new fuel types and provide margin for power uncertainty. These analyses assumed that the containment inservice purge system penetrations are isolated, thus this license amendment request proposes Technical Specification revisions which will require these penetrations to be blind flanged during plant operations; these changes allow the Technical Specification requirements for containment ventilation isolation instrumentation to be removed. This license amendment request also proposes associated more restrictive limits in the Technical Specification for reactor coolant system specific activity since the main steam line break accident analysis assumed lower limits.

This license amendment request does not involve physical changes to the plant structures, systems or components and there is no adverse impact on component or system interactions due to the proposed changes. The modes of operation of the plant remain unchanged and the design functions of the safety Page 1 8 of 2 1

Enclosure Dose Analyses NMC systems remain in compliance with the applicable safety analysis acceptance criteria. These changes do not create new failure modes or mechanisms and no new accident precursors are generated.

When the containment inservice purge system is not being operated, current Technical Specifications require the system's penetrations to be blind flanged in Modes 1, 2, 3, and 4 to provide post-accident containment integrity. This license amendment proposes to require the system penetrations to be blind flanged at all times during these Modes and prevent operation of the system in these Modes. Since containment integrity is provided with the penetrations blind flanged and this change only extends the time during which the system is in this configuration, these changes do not create the possibility of a new or different kind of accident.

Therefore, the proposed changes do not create the possibility of a new or different kind of accident from any previously evaluated.

Does the proposed amendment involve a significant reduction in a margin of safety?

Response: No This license amendment request proposes implementing revised loss of coolant accident and main steam line break accident dose consequence analyses to address modeling nonconservatisms and update the analyses for new fuel types and provide margin for power uncertainty. These analyses assumed that the containment inservice purge system penetrations are isolated, thus this license amendment request proposes Technical Specification revisions which will require these penetrations to be blind flanged during plant operations; these changes allow the Technical Specification requirements for containment ventilation isolation instrumentation to be removed. This license amendment request also proposes associated more restrictive limits in the Technical Specification for reactor coolant system specific activity since the main steam line break accident analysis assumed lower limits.

The loss of coolant accident and main steam line break accident radiological dose consequences analyses have incorporated revised inputs, including the proposed lower Technical Specification reactor coolant system specific activity limits, and utilized revised methodologies. The results of these revised analyses satisfy the applicable regulatory limits and guidance. There is no adverse effect on plant safety due to this proposed license amendment.

The containment inservice purge system is not credited for mitigation of any accidents or any other safety function, thus, removal of its associated Technical Specifications does not involve reduction in a margin of safety. The containment ventilation isolation instrumentation system is credited for isolation of the Page 19of21

Enclosure Dose Analyses NMC containment inservice purge system following an accident and the valves are assumed to meet containment integrity leakage rate limits. This license amendment request propose to require the containment inservice purge system containment penetrations to be blind flanged during Modes 1, 2, 3, and 4 and the blind flanged penetrations will be required to meet containment integrity leakage rate limits. With these changes, containment integrity is maintained in accordance with the current Technical Specification requirements, thus, this change does not involve reduction in a margin of safety.

Therefore, the proposed changes do not involve a significant reduction in a margin of safety.

Based on the above, the Nuclear Management Company concludes that the proposed amendment does not involve a significant hazards consideration under the standards set forth in 10 CFR 50.92(c) and, accordingly, a finding of "no significant hazards consideration" is justified.

4.4 Conclusions In conclusion, based on the considerations discussed in above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

5.

ENVIRONMENTAL CONSIDERATION A review has determined that the proposed amendment would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or would change an inspection or surveillance requirement. However, the proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluent that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(~)(9). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.

6.

REFERENCES

1.

Westinghouse Nuclear Safety Advisory Letter (NSAL)00-004, "Nonconservatisms in Iodine Spiking Factors", dated March 7, 2000.

Page 20 of 21

Enclosure Dose Analyses NMC

2.

Beaver Valley Power Station, lssuance of Amendment RE: Selective Implementation of Alternate Source Term and Control Room Habitability Technical Specification (TAC NOS. MB5303 and MB5304), dated September 10, 2003, Accession Number ML032530204.

3.

Fort Calhoun Station, Unit No. 1 - lssuance of Amendment (TAC No. MB1221),

dated December 5,2001, Accession Number ML013030027.

4.

Kewaunee Power Station License Amendment Request 21 1, "Radiological Accident Analysis and Associated Technical Specifications Change", dated January 30,2006, Accession Number ML060540217.

5.

Kewaunee Power Station - lssuance of Amendment RE: Radiological Accident Analysis and Associated Technical Specifications Change (TAC No. MC9715),

dated March 8, 2007, Accession Number ML070430020.

Page 21 of 21

ENCLOSURE 1, ATTACHMENT 1 Technical Specification Pages (Markup) 8 pages follow

3.3 INSTRUMENTATION 3.3.5 N-qt-U s

x d

/

n l

d Prairie Island Units 1 and 2 Unit 1 - Amendment No. 4-58 3.3.5-1 Unit 2 - Amendment No. 4-49 Dl C T T W 1 I.JIJ I 1 \\.

TIME 4ketwf h.;---(-mq&-d ru L. !

7

?&&ti%

RCS Specific Activity 3.4.17 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.17 RCS Specific Activity LC0 3.4.17 The specific activity of the reactor coolant shall be within limits.

APPLICABILITY:

MODES 1 and 2, MODE 3 with RCS average temperature (T,,,) 2 500°F.

ACTIONS A.

DOSE EQUIVALENT 1-131 > 0. 5 M pCi/gm.

COMPLETION TIME CONDITION

NOTE ---------------

LC0 3.0.4.c is applicable.

A.l Verify DOSE EQUIVALENT I-13 1 within the acceptable region of Figure 3.4.17-1.

Prairie Island Units 1 and 2 REQUIRED ACTION Once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months A.2 Restore DOSE EQUIVALENT I-13 1 to within limit.

Unit 1 - Amendment No. 424 447 3.4.17-1 Unit 2 - Amendment No. 149 424 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months

RCS Specific Activity 3.4.17 SR 3.4.17.2 NOTE..........................

Only required to be performed in MODE 1.

SURVEILLANCE REQUIREMENTS (continued)

Verify reactor coolant DOSE EQUIVALENT I-13 1 specific activity 5 Ild4-0 pCi/gm.

SURVEILLANCE 14 days FREQUENCY AND Between 2 and 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months after a THERMAL POWER change of > 15% RTP within a 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months period SR 3.4.17.3 NOTE -------- - -- --------- ------

Not required to be performed until 3 1 days after a minimum of 2 effective full power days and 20 days of MODE 1 operation have elapsed since the reactor was last subcritical for > 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months .

Determine E from a sample taken in MODE 1 after a minimum of 2 effective full power days and 20 days of MODE 1 operation have elapsed since the reactor was last subcritical for > 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months .

184 days Prairie Island Units 1 and 2 Unit 1 - Amendment No. 44%

3.4.17-3 Unit 2 - Amendment No. 4-49

RCS Specific Activity 3.4.17 OPERATION ACCEPTABLE OPERATION PERCENT OF RATED THERMAL POWER Figure 3.4.17-1 (page 1 of 1)

Reactor Coolant DOSE EQUIVALENT 1-1 3 1 Specific Activity Limit Versus Percent of RATED THERMAL POWER Prairie Island Units 1 and 2 Unit 1 - Amendment No. -Mg 3.4.17-4 Unit 2 - Amendment No. 449-

Containment Isolation Valves 3.6.3 3.6 CONTAINMENT SYSTEMS 3.6.3 Containment Isolation Valves LC0 3.6.3 Each containment isolation valve shall be OPERABLE.

APPLICABILITY:

MODES 1, 2, 3, and 4.

ACTIONS NOTES.................................................

1. Penetration flow path(~) except for 36-inch containment purge c?.n.d.....l8-inch.

inservice lzurgesystem flow paths may be unisolated intermittently under administrative controls.

2.

Separate Condition entry is allowed for each penetration flow path.

3. Enter applicable Conditions and Required Actions for systems made inoperable by containment isolation valves.

4.

Enter applicable Conditions and Required Actions of LC0 3.6.1, "Containment,"

when isolation valve leakage results in exceeding the overall containment leakage rate acceptance criteria.

Prairie Island Units 1 and 2 Unit 1 - Amendment No. 44%

3.6.3-1 Unit 2 - Amendment No. 4-49

Containment Isolation Valves 3.6.3 COMPLETION TIME ACTIONS (continued)

D. One or more secondary containment bypass leakage not within limit.

CONDITION D. 1 Restore leakage within limit.

REQUIRED ACTION 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months E. Containment purge blind flange or inservice purge blind flange leakage not within limit.

E. 1 Restore leakage within 1 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months limit.

F. Required Action and associated Completion Time not met.

F.l BeinMODE3.

AND 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months 36 hours Prairie Island Units 1 and 2 Unit 1 - Amendment No. 44%

3.6.3-5 Unit 2 - Amendment No. 4-49

Containment Isolation Valves 3.6.3 SR 3.6.3.1 VeriQ each 36-inch containment purge penetration blind flange is installed.

SURVEILLANCE REQUIREMENTS Prior to entering MODE 4 fiom MODE 5 SURVEILLANCE SR 3.6.3.2 Verify each 18-inch containment inservice purge penetration ifblind flanged is instal led^

3,Ct,is-l-.

FREQUENCY j3r ior to ente1i.n.g MOLE 4 from MODE 5+U%i~

4-&kh SR 3.6.3.3............................

NOTE --------....................

Valves and blind flanges in high radiation areas may be verified by use of administrative means.

VerifL each containment isolation manual valve and blind flange that is located outside containment and not locked, sealed, or otherwise secured and required to be closed during accident conditions is closed, except for containment isolation valves that are open under administrative controls.

92 days Prairie Island Units 1 and 2 Unit 1 - Amendment No. 43-8 3.6.3-6 Unit 2 - Amendment No. 1-49

Containment Isolation Valves 3.6.3 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE SR 3.6.3.5 Verify the isolation time of each automatic power operated containment isolation valve is within limits.

SR 3.6.3.7 Verify each automatic containment isolation valve that is not locked, sealed or otherwise secured in position, actuates to the isolation position on an actual or simulated actuation signal.

SR 3.6.3.8 Verify the combined leakage rate for all secondary containment bypass leakage paths is in accordance with the Containment Leakage Rate Testing Program.

FREQUENCY In accordance with the Inservice Testing Program 24 months In accordance with the Containment Leakage Rate Testing Program Prairie Island Units 1 and 2 Unit 1 - Amendment No. 4-58 3.6.3-7 Unit 2 - Amendment No. 4-49

ENCLOSURE, ATTACHMENT 2 Bases Pages (Markup)

(For Information Only) 18 page follows

B 3.3 INSTRUMENTATION B 3.3.5 Not U s

e d

e Prairie Island Units 1 and 2 Unit 1 - Revision 5 Unit 2 - Revision G

RCS Operational LEAKAGE B 3.4.14 BASES (continued)

APPLICABLE Thc

- - sakty analyc ses, w11ich presi~~nc ths occurrcncc_of an RCS iodine SAFETY spike concurrent with an accident? address o_n_erationalI,EAKAGT:.

ANALYSES

'I'he total operational LEAKAGE is used to determine the iodine

'I I

appearance rate. -

lEA#lAL(fE. In is relatedto the safety analyses for LOCA; the amount of leakage can affect the probability of such an event. The safety analysis for an event resulting in steam discharge to the atmosphere assumes the total primary to secondary LEAKAGE is 1 gallon per minute from the faulted SG or is assumed to increase to 1 gallon per minute as a result of accident induced conditions plus 150 gallons per day from the intact SG. The LC0 requirement to limit primary to secondary LEAKAGE through any one SG to less than or equal to 150 gallons per day is significantly less than the conditions assumed in the safety analysis. When the alternate repair criteria discussed in Specification 5.5.8.c.2(c) are implemented for Unit 2 (only), the safety analysis assumes the leakage fiom the faulted SG is limited to 1.42 gallons per minute (based on a reactor coolant system temperature of 578 O F ).

Primary to secondary LEAKAGE is a factor in the dose releases outside containment resulting from a steam line break (SLB) accident. To a lesser extent, other accidents or transients involve secondary steam release to the atmosphere, such as a steam generator tube rupture (SGTR). The leakage contaminates the secondary fluid.

The USAR (Ref. 2) analysis for SGTR assumes the plant has been operating with a 5 gpm primary to secondary leak rate for a period of time sufficient to establish radionuclide equilibrium in the secondary loop. Following the tube rupture, the initial primary to secondary LEAKAGE safety analysis assumption is relatively inconsequential I

when compared to the mass transfer through the ruptured tube.

The SLB is more limiting for site radiation releases. The safety analysis for the SLB accident assumes the total primary to secondary LEAKAGE is 1 gallon per minute from the faulted SG or is assumed Prairie Island Units 1 and 2 Unit 1 - Revision 487 B 3.4.14-2 Unit 2 - Revision 4-87

RCS Specific Activity B 3.4.17 B 3.4 REACTOR COOLANT SYSTEM (RCS)

B 3.4.17 RCS Specific Activity BASES BACKGROUND The maximum dose to the whole body and the thyroid that an individual at the site boundary can receive for 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months during an accident is specified in 10 CFR 100 (Ref. 1). The limits on specific activity ensure that the doses are held to a small fraction of the 10 CFR 100 limits during analyzed transients and accidents.

The RCS specific activity LC0 limits the allowable concentration level of radionuclides in the reactor coolant. The LC0 limits are established to minimize the offsite radioactivity dose consequences in the event of a steam generator tube rupture (SGTR) accident. 7'he IIOSI:

I:QUIVAI,I'N'I' 1-1 3 1 limit has been ---- established to ensure ilrgt&e_control raandose-a~~mmecriteria are ~net foIlo\\ving a main steam line break (MSI,H) when applying the alternative yg l t ~ ~ : ~ i e ~ i g z The LC0 contains specific activity limits for both DOSE EQUIVALENT I-13 1 and gross specific activity. The allowable levels are intended to limit the 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months dose at the site boundary to a small fraction of the 10 CFR 100 dose guideline limits. Except for tile

. -.. DOSE EQIJIVA-L-EN'l; 1-131 limit, tqhe limits in the LC0 are standardized, based on parametric evaluations of offsite radioactivity dose consequences for typical site 1ocations.The LC0 limit for DOSE EQUIVAI,ENT I-13 1 was derived lrorn thc corltrol roonl dose consequence a11allsis fo11o~vi1 a MSLB ass~unin,rr; alternate r g ~ a i : -crilc ria The parametric evaluations showed the potential offsite dose levels for a SGTR accident were an appropriately small fraction of the 10 CFR 100 dose guideline limits. Each evaluation assumes a broad range of site applicable atmospheric dispersion factors in a parametric evaluation._ The-basis f ~ r

_t_he_DOSE EQUIVALENT 1-1 3 1 Iitnit is deriyegJron1 I'rair-I s I L ~ ~ ~ ~

Nuclear Generatii~g I ' l g t Prairie Island Units 1 and 2 Unit 1 - Revision w e

B 3.4.17-1Unit 2 - Re\\,ision

RCS Specific Activity B 3.4.17 BASES (continued)

APPLICABLE The LC0 limits on the specific activity of the reactor coolant SAFETY ensures that the resulting 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months doses at the site boundary will not ANALYSES exceed a small fraction of the 10 CFR 100 dose guideline limits following a SGTR accident with an existing reactor coolant steam generator (SG) tube leakage rate of 1 gpm. B ~ e p t for the DOSE I ~ Q L J l V A l, l ~ N ~ l '

I-13 I Iin~it~tThe values for the limits on specific activity represent limits based upon a parametric evaluation by the NRC of typical site locations. These values are conservative in that specific site parameters of the Prairie Island site, such as site boundary location and meteorological conditions, were not considered in this evaluation (Ref. 2). T h e DOSE EOIJLVALJENrT I-13llimit ishd~scdon a site sgecific dose conscgucncc analysis and ensures t h a ~ t k offsite, as well as the control room. dose limits are rn The limits on RCS specific activity are also used for establishing standardization in radiation shielding and plant personnel radiation protection practices.

RCS specific activity satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii).

The specific iodine activity is limited to 0.544 pCi/gm DOSE EQUIVALENT I-13 1, and the gross specific activity in the reactor coolant is limited to the number of pCi/gm equal to 100 divided by E (average disintegration energy of the sum of the average beta and

, \\

1

?

gamma energies of the coolant nuclides).

I:hIT 1 1 2 1 J L I I 1, I I J 1 L 7

e

. The limit on I>OSI:

- - l<Qt JIVAT.I~N'I' 1-1 3 1 ensures that the thyroid dose to personnel in the_coritrol ro(xli during the Design Basis Accidta will be ~vithin the allowed li~n-iLThe limit on gross specific O H A I

L activity ensures the 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months whole body dose to an individual at the site boundary during the DBA will be a small fraction of the allowed whole body dose.

Prairie Island Units 1 and 2 Unit 1 - f-kvision W M

B 3.4.17-2Unit 2 - Rclaision

RCS Specific Activity B 3.4.17 BASES LC0 The SGTR accident analysis (Ref. 3) shows that the 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months site (continued) boundary dose levels are within acceptable limits. 'I'he MSI,13 accide~~t anal1rsis~R~~~3) s l ~ o \\ ~. s ~ h ~ m n t r o l ~ r o o n ~

dose levels

~ ~ ~ v i t h i n acceptable lin~its assi~n~ing the limiting prin~ary to secondary leakage, Violation of the LC0 may result in reactor coolant radioactivity levels that could, in the event of an SGTR, lead to site boundary doses that exceed the 10 CFR 100 dose guideline limits.

APPLICABILITY In MODES 1 and 2, and in MODE 3 with RCS average temperature 2 500°F, operation within the LC0 limits for DOSE EQUIVALENT I-13 1 and gross specific activity are necessary to contain the potential consequences of an SGTR to within the acceptable site boundary dose values. land MSLR to ~vithia acceptable control roan dose limits.

For operation in MODE 3 with RCS average temperature < 500°F, and in MODES 4 and 5, the release of radioactivity in the event of a SGTR is unlikely since the saturation pressure of the reactor coolant is below the lift pressure settings of the main steam safety valves.

ACTIONS A.l and A.2 With the DOSE EQUIVALENT I-13 1 greater than the LC0 limit, samples at intervals of 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months must be taken to demonstrate that the limits of Figure 3.4.17-1 are not exceeded. The Completion Time of 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months is required to obtain and analyze a sample. Sampling is done to continue to provide a trend.

The DOSE EQUIVALENT I-13 1 must be restored to within limits within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months . The Completion Time of 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months is required, if the limit violation resulted from normal iodine spiking.

Prairie Island Units 1 and 2 Unit 1 - f<evisicx~ -M B 3.4.17-3Unit 2 - Revision-

Containment B 3.6.1 BASES ACTIONS B. 1 and B.2 (continued) based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

SURVEILLANCE SR 3.6.1.1 REQ-Maintaining the containment OPERABLE requires compliance with the visual examinations and leakage rate test requirements of the Containment Leakage Rate Testing Program. Failure to meet air lock, &secondary containment (shield building and auxiliary building special ventilation zone) bypass leakage path 1.,

7

<limits specified in LC0 3.6.2 and LC0 3.6.3 does not invalidate the acceptability of these overall leakage determinations unless their contribution to overall Type A, B, and C leakage causes that to exceed limits. As left leakage prior to the first startup after performing a required Containment Leakage Rate Testing Program leakage test is required to be I 0.6 La for combined Type B and C leakage, and I 0.75 La for overall Type A leakage. At all other times between required leakage rate tests, the acceptance criteria are based on an overall Type A leakage limit of

< 1.0 La. At 2 1.0 La the offsite dose consequences are bounded by the assumptions of the safety analysis. SR Frequencies are as required by the Containment Leakage Rate Testing Program. These periodic testing requirements verify that the containment leakage rate does not exceed the leakage rate assumed in the safety analysis.

Prairie Island Units 1 and 2 Unit 1 - Amendment No. 15 8 B 3.6.1-5 Unit 2 - Amendment No. 149

Containment Isolation Valves B 3.6.3 RASES BACKGROUND In addition to the normal fluid systems which penetrate containment, (continued) two systems which can provide direct access from inside containment to the outside environment are described below.

Containment Purge System (36 inch purge valves)

The Containment Purge System operates to supply outside air into the containment for ventilation and cooling or heating and may also be used to reduce the concentration of noble gases within containment prior to and during personnel access in MODES 5 and

6. The supply and exhaust lines each contain one isolation valve, one isolation damper and a blind flange. The 36 inch purge valves and dampers are not tested to verify their leakage rate is within the acceptance criteria of the Containment Leakage Rate Testing Program. Therefore, blind flanges are installed in MODES 1, 2, 3, and 4 to ensure the containment boundary is maintained.

Inservice Purge System (1 8 inch purge valves)

The Inservice Purge System operates to:supply outside air intcfihe containinent-for ventilation and cooling or heating and may also be used to reduce the concentration of noble gases within containinent prior to and during personnel access in MODES 5 and 6.

a. R c

C 1. 3

.1.

I,

I a

b. ?**

Two containment automatic isolation valves and an automatic shield building ventilation damper are provided on each supply and exhaust r

line.

B 3

e t

I l

n p

p The 18

-- inch purge-valves and dampers are not test-ed to verify their-leakage rate is within the acccptanx critcrk o_flha Containmgnt Leakage Rate Testing Program. _ Tl~creforc~

blind &nges are instalkdin MODES 1,2,3, and 4 to ensurc thc cgn@inn~ent boundai~ is lnaii~t_ainscI.

Prairie Island Units 1 and 2 Unit I - Revision -.

! 5 8 B 3.6.3-3 Unit 2 - Revision

Containment Isolation Valves B 3.6.3 BASES BACKGROUND Inservice Purge

- System (1 8 inch purge

- valves) (continued) h L,

r installed on penetrations 42B and 43A (52 and 53 in Unit 2) and tested to meet the acceptance criteria of the Containment Leakage Rate Testing Program.

APPLICABLE The containment isolation valve LC0 was derived from the SAFETY assumptions related to minimizing the loss of reactor coolant ANALYSES inventory and establishing the containment boundary during major accidents. As part of the containment boundary, containment isolation valve OPERABILITY supports leak tightness of the containment. Therefore, the safety analyses of any event requiring isolation of containment is applicable to this LCO.

The DBAs that result in a release of radioactive material to the containment atmosphere are a loss of coolant accident (LOCA) and a rod ejection accident (Ref. 3). In the analyses for each of these accidents, it is assumed that containment isolation valves are either closed or hnction to close within the required isolation time following event initiation. This ensures that potential paths to the environment through containment isolation valves are minimized.

The safety analyses assume that the 36 inch purge lines 1-8inch i~lservice p-urge liil_es are blind flanged at event initiation.

Prairie Island Units 1 and 2 Unit 1 - Revision B 3.6.3-4 Unit 2 - Revision

Containment Isolation Valves B 3.6.3 BASES APPLICABLE In calculation of control room and offsite doses following a LOCA, SAFETY the accident analyses assume that 25% of the equilibrium iodine ANALYSES inventory and 100% of the equilibrium noble gas inventory (continued) developed from maximum hll power operation of the core is immediately available for leakage from containment (Ref. 3). The containment is assumed to leak at the maximum allowable leakage rate, La, for the first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months of the accident and at 50% of this leakage rate for the remaining duration of the accident.

The containment penetration isolation valves ensure that the containment leakage rate remains below La by automatically isolating penetrations that do not serve post accident hnctions and providing isolation capability for penetrations associated with Engineered Safety Features. The maximum isolation time for automatic containment isolation valves is 60 seconds. This isolation time is based on engineering judgement since the control room and offsite dose calculations are performed assuming that leakage from containment begins immediately following the accident with no credit for transport time or radioactive decay. The 60 second isolation time takes into consideration the time required to drain piping of fluid which can provide an initial containment isolation before the containment isolation valves are required to close and the conservative assumptions with respect to core damage occurring immediately following the accident.

The containment isolation total response time of 60 seconds includes signal delay, diesel generator startup (for loss of offsite power), and containment isolation valve stroke times.

LOCA (Ref.&..

'))

L/ nn

). -,

7...

Prairie Island Units 1 and 2 Unit 1 - Revision -.

5 8 B 3.6.3-5 Unit 2 - lievision

Containment Isolation Valves B 3.6.3 BASES LC0 Vent and drain valves located between two isolation devices are also (continued) containment isolation devices. Test connections located between two isolation valves are similar to vent and drain lines except that no valve may exist in the test line. A cap or blind flange, as applicable, must be installed on these vent, drain and test lines. A cap or blind flange installed on these lines make them "otherwise secured" for SR considerations.

The automatic power operated isolation valves are required to have isolation times within limits and to actuate on an automatic isolation signal. The 36 inch purge valves and 18 inch inservice Durjze valve2 must be blind flanged in MODES 1,2, 3, and 4. The valves covered by this LC0 are listed in Reference 2.

The normally closed isolation valves are considered OPERABLE when manual valves are closed, automatic power operated valves are de-activated and secured in their closed position, blind flanges are in place, and closed systems are intact. These passive isolation valves/devices are those listed in Reference 2.

,- r c

Ssecondary d

l d

i n

g special ventilation zone) bypass valves must meet additional leakage rate requirements. The other containment isolation valve leakage rates are addressed by LC0 3.6.1, "Containment," as Type C testing.

This LC0 provides assurance that the containment isolation valves li l

n r

j w

i l

l perform their designed safety functions to minimize the loss of reactor coolant inventory and establish the containment boundary during accidents.

Prairie Island Units 1 and 2 Unit 1 - Revision B 3.6.3-7 Unit 2 - Revision

Containment Isolation Valves B 3.6.3 BASES (continued)

APPLICABILITY In MODES 1,2, 3, and 4, a DBA could cause a release of radioactive material to containment. In MODES 5 and 6, the probability and consequences of these events are reduced due to the pressure and temperature limitations of these MODES. Therefore, the containment isolation valves are not required to be OPERABLE in MODE 5. The requirements for containment isolation valves during MODE 6 are addressed in LC0 3.9.4, "Containment Penetrations."

ACTIONS The ACTIONS are modified by four Notes. The first Note allows penetration flow paths, except for 36 inch containment purge and 18 inch containment inservice purge system penetration flow paths, to be unisolated intermittently under administrative controls. These administrative controls consist of stationing a dedicated operator at the valve controls, who is in continuous communication with the control room. In this way, the penetration can be rapidly isolated when a need for containment isolation is indicated. Due to the blind flanges on the containment purge and containment inservice purge system lines during plant operation, the penetration flow path containing these flanges may not be opened under administrative controls.

A second Note has been added to provide clarification that, for this LCO, separate Condition entry is allowed for each penetration flow path. This is acceptable, since the Required Actions for each Condition provide appropriate compensatory actions for each inoperable containment isolation valve. Complying with the Required Actions may allow for continued operation, and subsequent inoperable containment isolation valves are governed by subsequent Condition entry and application of associated Required Actions.

The ACTIONS are hrther modified by a third Note, which ensures appropriate remedial actions are taken, if necessary, if the affected systems are rendered inoperable by an inoperable containment isolation valve.

Prairie Island Units 1 and 2 Unit 1 - Revision B 3.6.3-8 Unit 2 - lievision

Containment Isolation Valves B 3.6.3 BASES ACTIONS In the event containment isolation valve leakage results in exceeding (continued) the overall containment leakage rate acceptance criteria, Note 4 directs entry into the applicable Conditions and Required Actions of LC0 3.6.1.

A. 1 and A.2 In the event one containment isolation valve in one or more penetration flow paths is inoperable, except for

-1.,

-secondary containment bypass leakage not within limit, the affected penetration flow path must be isolated. The method of isolation must include the use of at least one isolation barrier that cannot be adversely affected by a single active failure.

Isolation barriers that meet this criterion are a closed and de-activated or mechanically blocked power operated containment isolation valve, a closed manual valve, a blind flange, and a check valve with flow through the valve secured. For a penetration flow path isolated in accordance with Required Action A. 1, the device used to isolate the penetration should be the closest available one to containment. Required Action A. 1 must be completed within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . The 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months Completion Time is reasonable, considering the time required to isolate the penetration and the relative importance of supporting containment OPERABILITY during MODES 1, 2, 3, and 4.

For affected penetration flow paths that cannot be restored to OPERABLE status within the 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months Completion Time and that have been isolated in accordance with Required Action A. 1, the affected penetration flow paths must be verified to be isolated on a periodic basis. This is necessary to ensure that containment penetrations required to be isolated following an accident and no longer capable of being automatically isolated will be in the isolation position should an event occur. This Required Action does not Prairie Island Units 1 and 2 Unit 1 - Revision B 3.6.3-9 Unit 2 - lievision

Containment Isolation Valves B 3.6.3 BASES ACTIONS B.1 (continued)

With two containment isolation valves in one or more penetration 1

flow paths inoperable, except for i i

secondary containment bypass leakage not within limits, the affected penetration flow path must be isolated within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . The method of isolation must include the use of at least one isolation barrier that cannot be adversely affected by a single active failure. Isolation barriers that meet this criterion are a closed and de-activated valve, a closed manual valve, and a blind flange. The 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months Completion Time is consistent with the ACTIONS of LC0 3.6.1. In the event the affected penetration is isolated in accordance with Required Action B. 1, the affected penetration must be verified to be isolated on a periodic basis per Required Action A.2, which remains in effect. This periodic verification is necessary to assure leak tightness of containment and that penetrations requiring isolation following an accident are isolated. The Completion Time of once per 3 1 days for verifying each affected penetration flow path is isolated is appropriate considering the fact that the valves are operated under administrative control and the probability of their misalignment is low.

Condition B is modified by a Note indicating this Condition is only applicable to penetration flow paths with two containment isolation valves. Condition A of this LC0 addresses the condition of one containment isolation valve inoperable in this type of penetration flow path.

Prairie Island Units I and 2 Unit 1 - Revision B 3.6.3-1 1 Unit 2 - Iievision -&

Containment Isolation Valves B 3.6.3 BASES ACTIONS C. 1 and C.2 (continued) necessary since this Condition is written to specifically address those penetration flow paths in a closed system.

Required Action C.2 is modified by two Notes. Note 1 applies to valves and blind flanges located in high radiation areas and allows these devices to be verified closed by use of administrative means.

Allowing verification by administrative means is considered acceptable, since access to these areas is typically restricted. Note 2 applies to isolation devices that are locked, sealed, or otherwise secured in position and allows these devices to be verified closed by use of administrative means. Allowing verification by administrative means is considered acceptable, since the function of locking, sealing, or securing components is to ensure that these devices are not inadvertently repositioned. Therefore, the probability of misalignment of these valves, once they have been verified to be in the proper position, is small.

With the secondary containment bypass leakage rate (SR 3.6.3.8) w

('-1 l not within limit, the assumptions of the safety analyses are not met. Therefore, the leakage must be restored to within limit within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

Restoration can be accomplished by isolating the penetration(s) that caused the limit to be exceeded by use of one closed and de-activated automatic valve, closed manual valve, or blind flange.

Prairie Island Units 1 and 2 Unit 1 - Kevision--

B 3.6.3-13 Unit 2 - Revision

Containment Isolation Valves B 3.6.3 BASES ACTIONS D.1 (continued)

When a penetration is isolated the leakage rate for the isolated penetration is assumed to be the actual pathway leakage through the isolation device. If two isolation devices are used to isolate the penetration, the leakage rate is assumed to be the lesser actual pathway leakage of the two devices. The 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months Completion Time is reasonable considering the time required to restore the leakage by isolating the penetration(s) and the relative importance of secondary containment bypass leakage and containment purge penetration valve(s) leakage to the overall containment fbnction.

In the event containment purge blind flange leakage &(SR 3.6.3.1) or containment inservice purge blind flange leakage*

(SR 3.6.3.2) are not within limits, the leakage rate must be restored within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months to assure containment leakage rates are met. If containment purge blind flange leakage rate or containment inservice purge blind flange leakage rate limits are not met, it could be due to the blind flange not installed or improperly installed. The 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months Completion Time provides a period of time to correct the problem commensurate with I

the importance of maintaining containment OPERABLE during MODES 1, 2, 3, and 4. This time period also ensures that the probability of an accident (requiring containment OPERABILITY) occurring during periods when blind flange leakage exceeds its limits is minimal.

Prairie Island Units 1 and 2 Unit 1 - Revision 468 B 3.6.3-14 Unit 2 - Revision 4-68

Containment Isolation Valves B 3.6.3 BASES ACTIONS F.l and F.2 (continued)

If the Required Actions and associated Completion Times are not met, the plant must be brought to a MODE in which the LC0 does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and to MODE 5 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions fi-om full power conditions in an orderly manner and without challenging plant systems.

SURVEILLANCE SR 3.6.3.1 REQ-Each 36 inch containment purge system penetration is required to be blind flanged when the plant is in MODES 1, 2, 3, and 4. This Surveillance is designed to ensure that the blind flange is installed prior to entering MODE 4 from MODE 5.

Each-7.,,

18 inch containment inservice purge penetrations is rcqujred to bc meblind flanged when the plant isin MODES 1. 2: 3, and 4.

This S~n.eillancc is de2-ignn to e m r e that _t_hc blind flang~is installed prior to cntcfing MODE 4 fromMQDFi5.

Prairie Island Units 1 and 2 Unit 1 - Revision

. l.. 58 B 3.6.3-1 5 Unit 2 - Revision -.

149

Containment Isolation Valves B 3.6.3 BASES SURVEILLANCE SR 3.6.3.5 REQ-(continued)

Verifying that the isolation time of each automatic power operated containment isolation valve is within limits is required to demonstrate OPERABILITY. The isolation time test ensures the valve will isolate in a time period less than or equal to that assumed in the safety analyses. The isolation time and Frequency of this SR are in accordance with the Inservice Testing Program.

Automatic containment isolation valves close on a containment isolation signal to prevent leakage of radioactive material from containment following a DBA. This SR ensures that each automatic containment isolation valve will actuate to its isolation position on a containment isolation signal. This Surveillance is not required for valves that are locked, sealed, or otherwise secured in the required position under administrative controls. The 24 month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power. Operating experience has shown that these components usually pass this Surveillance when performed. Therefore, the Frequency was concluded to be acceptable from a reliability standpoint.

Prairie Island Units 1 and 2 Unit 1 - Revision B 3.6.3-18 Unit 2 - Revision

Containment Penetrations B 3.9.4 BASES BACKGROUND The containment air locks, which are also part of the containment (continued) pressure boundary, provide a means for personnel access during MODES l,2, 3, and 4 unit operation in accordance with LC0 3.6.2, "Containment Air Locks." Each air lock has a door at both ends.

The doors are normally interlocked to prevent simultaneous opening when containment OPERABILITY is required. During periods of unit shutdown when containment closure is not required, the door interlock mechanism may be disabled, allowing both doors of an air lock to remain open for extended periods when frequent containment entry is necessary.

During movement of recently irradiated he1 assemblies within containment, containment closure is required; therefore, the door I

interlock mechanism may remain disabled but one air lock door must always remain closed.

The requirements for containment penetration closure ensure that a release of fission product radioactivity within containment will be restricted to within regulatory limits.

I The Containment Purge and Exhaust System includes two subsystems, Containment Purge and Containment Inservice Purge.

The containment purge subsystem includes a 36 inch purge penetration and a 36 inch exhaust penetration. The second subsystem, a minipurge system referred to as containment inservice purge, includes a 14 inch purge penetration and an 18 inch exhaust penetration.

During MODES l,2, 3, and 4, the two valves in each of the containment purge and exhaust penetrations are

. A 7

<blank flanged.

Prairie Island Units 1 and 2 Unit 1 - Revision Unit 2 - Revision

ENCLOSURE, ATTACHMENT 3 Technical Specification Pages (Retyped) 8 pages follow

Not Used 3.3.5 3.3 INSTRUMENTATION 3.3.5 Not used Prairie Island Units 1 and 2 Unit 1 - Amendment No. 448 3.3.5-1 Unit 2 - Amendment No. 149

RCS Specific Activity 3.4.17 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.17 RCS Specific Activity LC0 3.4.17 The specific activity of the reactor coolant shall be within limits.

APPLICABILITY:

MODES 1 and 2, MODE 3 with RCS average temperature (T,,,) >50O0F.

ACTIONS A.

DOSE EQUIVALENT 1-13 1 > 0.5 pCi/gm.

NOTE ---------------

LC0 3.0.4.c is applicable.

COMPLETION TIME CONDITION A. 1 Verify DOSE EQUIVALENT I-13 1 within the acceptable region of Figure 3.4.17-1.

REQUIRED ACTION AND A.2 Restore DOSE EQUIVALENT I-13 1 to within limit.

I Once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months 48 hours Prairie Island Units 1 and 2 Unit 1 - Amendment No. 458 4-67 3.4.17-1 Unit 2 - Amendment No. 4-49 +53

RCS Specific Activity 3.4.17 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE SR 3.4.17.2 NOTE..........................

Only required to be performed in MODE 1.

Verify reactor coolant DOSE EQUIVALENT I-13 1 specific activity 5 0.5 pCiIgm.

SR 3.4.17.3 NOTE..........................

Not required to be performed until 3 1 days after a minimum of 2 effective full power days and 20 days of MODE 1 operation have elapsed since the reactor was last subcritical for 3 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months .

~etermine E from a sample taken in MODE 1 after a minimum of 2 effective full power days and 20 days of MODE 1 operation have elapsed since the reactor was last subcritical for 2 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months .

FREQUENCY 14 days AND Between 2 and 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months after a THERMAL POWER change of 2 15% RTP within a 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months period 184 days Prairie Island Units 1 and 2 Unit 1 - Amendment No. 4-58 Unit 2 - Amendment No. 449

DOSE EQUIVALENT 1-131 PRIMARY COOLANT SPECIFIC ACTIVITY LIMIT ( pCi/gm)

Containment Isolation Valves 3.6.3 3.6 CONTAINMENT SYSTEMS 3.6.3 Containment Isolation Valves LC0 3.6.3 Each containment isolation valve shall be OPERABLE.

APPLICABILm MODES 1, 2, 3, and 4.

ACTIONS NOTES-------------------------------------------------

1.

Penetration flow path(s) except for 36-inch containment purge and 18-inch inservice purge system flow paths may be unisolated intermittently under administrative controls.

2.

Separate Condition entry is allowed for each penetration flow path.

3. Enter applicable Conditions and Required Actions for systems made inoperable by containment isolation valves.

4.

Enter applicable Conditions and Required Actions of LC0 3.6.1, "Containment,"

when isolation valve leakage results in exceeding the overall containment leakage rate acceptance criteria.

Prairie Island Units 1 and 2 Unit 1 - Amendment No. 458 3.6.3-1 Unit 2 - Amendment No. 4-49

Containment Isolation Valves 3.6.3 ACTIONS (continued)

CONDITION D. One or more secondary containment bypass leakage not within limit.

E. Containment purge blind flange or inservice purge blind flange leakage not within limit.

F. Required Action and associated Completion Time not met.

Prairie Island Units 1 and 2 REQUIRED ACTION D. 1 Restore leakage within limit.

E. 1 Restore leakage within limit.

AND COMPLETION TIME 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months 1 hour 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months 36 hours Unit 1 - Amendment No. 448 3.6.3-5 Unit 2 - Amendment No. 4-49

Containment Isolation Valves 3.6.3 SR 3.6.3.1 Verify each 36-inch containment purge penetration blind flange is installed.

SURVEILLANCE REQUIREMENTS Prior to entering MODE 4 from MODE 5 SURVEILLANCE SR 3.6.3.2 Verify each 18-inch containment inservice purge penetration blind flange is installed.

FREQUENCY Prior to entering MODE 4 from MODE 5 SR 3.6.3.3 NOTE............................

Valves and blind flanges in high radiation areas may be verified by use of administrative means.

Verify each containment isolation manual valve and blind flange that is located outside containment and not locked, sealed, or otherwise secured and required to be closed during accident conditions is closed, except for containment isolation valves that are open under administrative controls.

92 days SR 3.6.3.4............................

NOTE----------------------------

Valves and blind flanges in high radiation areas may be verified by use of administrative means.

Verify each containment isolation manual valve and blind flange that is located inside containment and not locked, sealed, or otherwise secured and required to be closed during accident conditions is closed, except for containment isolation valves that are open under administrative controls.

Prior to entering MODE 4 from MODE 5 if not performed within the previous 92 days Prairie Island Units 1 and 2 Unit 1 - Amendment No. 44%

3.6.3-6 Unit 2 - Amendment No. 449

Containment Isolation Valves 3.6.3 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE SR 3.6.3.5 Verify the isolation time of each automatic power operated containment isolation valve is within limits.

SR 3.6.3.6 Not Used SR 3.6.3.7 VerifL each automatic containment isolation valve that is not locked, sealed or otherwise secured in position, actuates to the isolation position on an actual or simulated actuation signal.

SR 3.6.3.8 Verify the combined leakage rate for all secondary containment bypass leakage paths is in accordance with the Containment Leakage Rate Testing Program.

FREQUENCY In accordance with the Inservice Testing Program 24 months In accordance with the Containment Leakage Rate Testing Program Prairie Island Units I and 2 Unit 1 - Amendment No. 44%

Unit 2 - Amendment No. 149 OFFSITE & CONTROL ROOM DOSE CONSEQUENCES Followed by 36 pages

DOC. NO. 12400604-R-U-01 Rev. 0 PRAIRIE ISLAND NUCLEAR GENERATING PLANT OFFSITE & CONTROL ROOM DOSE CONSEQUENCES LOSS-OF-COOLANT ACCIDENT MAIN STEAM LINE BREAK Prepared for NUCLEAR MANAGEMENT COMPANY June 2008 A

sha;w

Stm & Webster Nuclear

Prairie Island Nuclear Generating Plant TABLE OF CONTENTS Section No. 1 Title Page I INTRODUCTION................................................................................................................. 3 2

REGULATORY APPROACH............................................................................................ 3 2.1 Changes to Current Licensing Basis.......................................................................

3 2.2 Changes to Current Design Basis Input.................................................................

5 2.3 Dose Acceptance Criteria......................................................................................... 6 3

COMPUTER CODES......................................................................................................... 7 4

RADIATION SOURCE TERMS........................................................................................ 7 4.1 Core Inventory............................................................................................................ 7 4.2 Coolant Activity........................................................................................................... 7 5

ACCIDENT ATMOSPHERIC DISPERSION FACTORS (xlQ).................................. 10 5.1 Site Boundary Atmospheric Dispersion Factors 10 5.2 Control Room Atmospheric Dispersion Factors 10 6

DOSE CALCULATION METHODOLOGY 12 7

RADIOLOGICAL ACCIDENT REANALYSES 14 7.1 Control Room Design Information......................................................................... 14 7.2 Loss of Coolant Accident (LOCA)..........................................................................

15 7.3 Main Steam Line Break (MSLB) 21 8

SUMMARY

OF RESULTS AND CONCLUSIONS...................................................... 23 9

REFERENCES.................................................................................................................. 25 DOC. NO.. 12400604-R-U-01 Rev 0

Prairie Island Nuclear Generating Plant 1

INTRODUCTION A design basis verification effort initiated by Nuclear Management Company (NMC) regarding the radiological design basis of Prairie Island Nuclear Generating Plant (PINGP) resulted in the identification of several concerns in the area of post-accident dose consequences. The findings fell into the following categories: (a) some of the plant specific design parameters used in selected dose consequence analyses were not current (e.g., the core and coolant inventory, control room volume and unfiltered inleakage); (b) in some cases, the atmospheric dispersion factors utilized did not address the actual release points / receptor locations; and (c) there were some technical deficiencies in selected analyses (e.g., the Loss-of-Coolant Accident analysis did not address all of the radioactivity release sources, and the activity transport model utilized for the dose assessment in the Control Room was found to be erroneous; the Main Steam Line break did not address the allowable primary coolant leakage when determining the equilibrium iodine appearance rates).

As a result of the corrective action associated with the above findings, the updated methodology and results of the dose consequence analyses for the following design basis accidents are being submitted to NRC for approval via this license application:

1. Loss of Coolant Accident (LOCA)

2. Main Steam Line Break (MSLB)

Included in these revised analyses are the use of updated on-site atmospheric dispersion factors and thyroid dose conversion factors that are representative of current NRC recommendations to the industry, and a request to reduce the primary coolant activity levels in the plant Technical Specifications.

The changes to design inputs for the remaining accidents were deemed minor, and the impact on dose consequences remained within the dose increase limits dictated by the 10CFR50.59 criteria. (Reference 1) 2 REGULATORY APPROACH 2.1 Changes to Current Licensing Basis Provided below is the summary of changes to the current licensing basis proposed by this application.

1. The current licensing basis LOCA and MSLB dose consequence analyses presented in USAR Section 14.9 and 14.5.5.6, respectively, utilize control room atmospheric dispersion factors that are based on methodology developed by James Halitsky which took into consideration a series of wind tunnel tests, the results of which were documented in Meteorology and Atomic Energy, 1968 (Reference 2).

In accordance with NRC current recommendations to the industry, the updated analyses utilize control room atmospheric dispersion factors that are based on ARCON96 (Reference 3) methodology. Usage of ARCON96 methodology at PINGP was previously approved by NRC for the Fuel Handling Accident via License DOC. NO.: 12400604-R-U-01 Rev 0 3 of 36 A

Shaw A slcne & w t w Nudeer

Prairie Island Nuclear Generating Plant Amendment 166 and 156, dated September 2004. (Reference 4) As noted in Section 3.2.1.1 of the NRC SER for License Amendment 166 and 156, the staff performed a quality review of the on-site hourly met data (1993 through 1997) and concluded that the data was an acceptable basis for making estimates of atmospheric dispersion for design basis accidents. The above NRC-approved on-site hourly met data was utilized to develop the ARCON96 on-site atmospheric dispersion factors used in the LOCA and MSLB dose consequence analyses.

2. The current Technical Specifications allow operation of the containment inservice (low flow) purge system during reactor operation. When operating the inservice purge, two containment isolation valves are provided on each supply and exhaust line which receives an automatic closure signal on receipt of a safety injection or containment high radiation signal. The Technical Specification (TS) requirements for the inservice purge are contained in TS 3.3.5 and 3.6.3.

With this application, PlNGP is proposing to remove the containment inservice purge capability from the Technical Specifications which will result in this system not operating during MODES 1, 2, 3, and 4. The radiological LOCA consequence analysis implicitly assumes that the containment inservice purge system is isolated. Therefore, in order to ensure that the release assumption of the analysis can be met, the Technical Specification which allow inservice purge to be operated during plant operating MODES will be removed and the containment inservice purge system will be isolated during MODES 1, 2, 3, or 4.

3. The current licensing basis LOCA dose consequence analysis presented in USAR Section 14.9 utilizes TID 14844 (Reference 5) thyroid dose conversion factors to estimate the thyroid doses at the exclusion area boundary, low population zone, and control room.

The current licensing basis MSLB dose consequence analysis presented in USAR Section 14.5.5.6 utilizes thyroid dose conversion factors based on ICRP 30 methodology to estimate the thyroid doses at the exclusion area boundary, low population zone and control room (Reference 6).

In accordance with NRC current recommendations to the industry, the updated analyses utilize inhalation dose conversion factors based on Table 2.1 of Federal Guidance 11 (Reference 7) to estimate the thyroid doses at the exclusion area boundary, low population zone and control room.

4. The current licensing basis LOCA dose consequence analysis presented in USAR Section 14.9 does not include the dose contribution due to sump water leakage outside containment from components and systems carrying post-accident fluids (called ESF leakage thereafter). Per USAR Section 6.7.2, the PlNGP licensing basis position is that the dose contribution due to ESF leakage is expected to be negligible.

As noted in USAR Section 6.7.2, more significant releases are expected if a major system leak is assumed; however, per NUREG 0800, SRP 15.6.5 (Reference 21), a major system leak 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months into the accident need not be considered if the area is filtered and exhausted via a safety related ventilation system as is the case with DOC. NO.: 12400604-R-U-01 Rev 0 4 of 36 n

ShaYY "

& V&WB ~uclear

Prairie Island Nuclear Generating Plant PINGP. There is no discussion in the USAR regarding Refueling Water Storage Tank (RWST) back-leakage.

In accordance with NRC current recommendations to the industry (as discussed in NUREG 0800, SRP 15.6.5 and Information Notice 91-56 (Reference 19) for ESF leakage and RWST back-leakage, respectively), the updated analysis addresses the contribution of both ESF leakage and RWST back-leakage.

5. The current primary coolant iodine concentrations allowable by PINGP TS 3.4.17 is 1 pCi/gm Dose Equivalent (DE) 1-1 31. Per TS 1.I, the dose conversion factors utilized to limit the iodines in the coolant to 1 pCi/gm DE 1-131 is based on Table Ill of TID 14844.

With this application, PINGP is requesting for a change to PINGP TS 3.4.17 to reduce the limit on the primary coolant iodine concentration from 1 pCi/gm DE 1-131 to 0.5 pCilgm DE 1-131. This change is required to ensure that the dose consequences following a Main Steam Line Break remain within regulatory limits based on the continued use of post-accident steam generator tube leakage due to accident induced conditions of 1 gpm (@ STP) in the faulted steam generator per TS Bases 3.4.14. The dose conversion factors utilized to limit the iodines in the coolant to 0.5 pCi1gm DE I-131 continues to be based on Table Ill of TID 14844.

2.2 Changes to Current Design Basis Input Listed below, by accident, are the principal input parameters whose values have been updated as a result of the corrective action process.

LOCA Core power level: Changed from 1650 MWt to 1683 MWt Core inventory: The fuel design and fuel management schemes utilized by PINGP have changed since the issuance of the current licensing basis LOCA analysis. (See Section 4.1 for further detail.)

Spray lambda for particulate removal : Changed from 0.45 hi' based on WASH 1329 to a calculated value of 5 hi' (see Section 7.2 for detail)

Spray initiation time: Changed from T=l min to T=42 seconds Spray cut-off time: Changed from 14.816 min to 14.516 min Auxiliary Building Special Ventilation System (ABSVS) and Shield Building Ventilation System (SBVS) filter efficiency for elemental and organic iodines: Changed from 90%

to 70% to be consistent with current Technical Specifications Auxiliary Building drawdown time prior to crediting ABSVS: Changed from T=O sec to T=6 min Control Room (CR) volume: Changed from 44,200 ft3 to 61,315 ft3.

Maximum normal operation ventilation air intake prior to CR isolation at 2 min:

Changed from 0 cfm to 2000 cfm CR unfiltered inleakage: Changed from 44 cfm to 165 cfm (both during normal operation and accident) per results of Tracer Gas Test CR unfiltered inleakage due to ingress I egress: Changed from 0 cfm to 10 cfm (both during normal operation and accident)

DOC. NO.: 12400604-R-U-01 Rev 0 5 of 36 A

ShEIW " ~tcne 8 WsW~uclear

Prairie Island Nuclear Generating Plant

> CR Emergency Ventilation filtered recirculation flow rate: Changed from 3000 cfm to 3600 cfm per current Technical Specifications I+

CR Recirculation filter efficiency: Changed from 95% (all iodine species) to 95% for elemental and organic iodine and 99% for particulate iodine per current Technical Specifications MSLB

'I-Primary and secondary coolant Technical Specification concentrations (See Section 4.2 for further detail)

> SG liquid mass: Changed from 109,155 Ibs I SG to 107,100 Ibs I SG

> Primary coolant letdown flow used to determine equilibrium iodine release rates from the core: Changed from 40 gpm at 127F I 15 psig to 90 gpm at l00F 1 15 psig

> Primary coolant leakage utilized to determine equilibrium iodine release rates from the core: Changed from 0 gpm to all leakage addressed in TS 3.4.14 k Dryout time of faulted SG: Changed from 15 min to 2 min I+

Primary coolant volume: Changed from 5227 ft3 to 5290 ft3

> CR volume: Changed from 165,000 ft3 to 61,315 ft3.

k CR unfiltered inleakage due to ingress / egress: Changed from 0 cfm to 10 cfm (both during normal operation and accident) i CR recirculation filter efficiency: Changed from 90% for elemental and organic iodine and 95% for particulate iodine to 95% for elemental and organic iodine and 99% for particulate iodine per current Technical Specifications 2.3 Dose Acceptance Criteria In accordance with current licensing basis, the acceptance criteria for the Exclusion Area Boundary (EAB) and the Low Population Zone (LPZ) doses for the LOCA are based on 10CFR Part 1 00.1 1 (Reference 8):

(i) An individual located at any point on the boundary of the exclusion area for 2-hours immediately following the onset of the postulated fission product release, would not receive a radiation dose in excess of 25 rem to the whole body, or 300 rem to the thyroid from iodine exposure.

(ii) An individual located at any point on the outer boundary of the low population zone, who is exposed to the radioactive cloud resulting from the postulated fission product release (during the entire period of its passage), would not receive a radiation dose in excess of 25 rem to the whole body, or 300 rem to the thyroid from iodine exposure.

Acceptance criteria for the MSLB is based on NUREG 0800, Standard Review Plan 15.1.5, Appendix A (Reference 23). Per Reference 23, the dose limits outlined in 1 OCFRI 00.1 1 remain applicable for the pre-accident iodine spike case, whereas the dose limits for the concurrent iodine spike case have been reduced to a small fraction of that allowed by 10CFR100.11, i.e.,

10% of the 1 OCFRI 00.1 1 limits.

DOC. NO.: 12400604-R-U-01 Rev 0 6 of 36

$ b W S m B W N u c t e a r

Prairie Island Nuclear Generating Plant The acceptance criterion for the Control Room Dose is based on 10CFR Part 50 Appendix A, GDC 19 (Reference 9) and NUREG 0800, SRP 6.4 (Reference 10):

Adequate radiation protection is provided to permit 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.

3 COMPUTER CODES The QA Category 1 computer codes utilized in the dose consequence analyses that support this application are listed below. The referenced computer codes have been used extensively to support nuclear power plant design:

1. Industry Computer Code ARCON96, "Atmospheric Relative Concentrations in Building Wakes" developed by PNL (SS&W Program EN-292, VOO, LOO).

2. Shaw Stone & Webster (SS&W) Proprietary Computer Code, PERC2, "Passive Evolutionary Regulatory Consequence Code", NU-226, VOO, L01.

3. SS&W Computer Code, SW-QADCGGP, "A Combinatorial Geometry Version of QAD-P5A, NU-222, VOO, L02, (Industry Computer Code QAD-CGGP).

4 RADIATION SOURCE TERMS 4.1 Core Inventory The fuel design and fuel management schemes utilized by PINGP have changed since the issuance of the current licensing basis LOCA and MSLB analyses. The inventory of fission products in the PINGP reactor core has since been updated by Westinghouse using current licensing basis methodology and based on maximum full-power operation of the core at a power level equal to the current licensed rated thermal power including a 2% margin for power uncertainty, and current licensed values of fuel enrichment and burnup.

It is noted that the updated core inventory utilized herein is conservative for current operation as it reflects the higher reactor core isotopic inventories between the current OFA fuel and a proposed fufure Heavy Bundle Fuel (HBF).

The PINGP equilibrium core inventory is presented in Table 4.1-1.

4.2 Coolant Activity The design basis (1 % fuel defects) primary coolant activity inventory has also been updated by Westinghouse using current licensing basis methodology to reflect the changes in fuel design and fuel management schemes utilized by PINGP. Consistent with the core inventory, the updated coolant inventory reflects the higher coolant isotopic inventories between the current OFA fuel and a proposed future Heavy Bundle Fuel (HBF)

DOC. NO.: 12400604-R-U-01 Rev 0 7 of 36 n

Shaw sla;e s bwxw Nuclear

Prairie Island Nuclear Generating Plant Technical Specification Primarv Coolant Activity TS SR 3.4.17.2 limits the specific activity for iodines in the primary coolant to 1 pCiIgm Dose Equivalent 1-1 31. Per this application, the specific activity for iodines in the primary coolant will be reduced to 0.5 pCiIgm.

TS SR 3.4.17.1 limits the non-iodine nuclides that make up >95% of the gross primary coolant activity with half-lives greater than 15 minutes to 100 I E,,

pCi1gm.

Development of primary coolant Technical Specification isotopic concentrations take into consideration the following:

s Isotopic compositions are based on the design basis primary coolant equilibrium concentrations at 1% defective fuel.

s Per TS 1. l, iodine concentrations in the coolant are based on thyroid dose conversion factors for 1-1 31, 1-132, 1-1 33, 1-1 34, and 1-1 35 obtained from TID-14844.

s Average beta and gamma energies per disintegration are based on References 12, 13, and 14.

Note that the estimated isotopic primary coolant Technical Specification concentration of fission products is based on the design basis fission product mix adjusted to reflect the more limiting of the "failed fuel percentages" associated with the two Technical Specification limits identified in TS SR 3.4.17.1 and 3.4.17.2.

This approach is reasonable as the mix of isotopes in the primary coolant is determined by the leakage of core activity from the defective fuels and the escape coefficients of the isotopes and its precursors. The above is already factored into the design basis coolant isotopic mix, and the referenced mix is not expected to change (i.e., between iodine vs. non-iodine isotopes) such that both Technical Specification conditions co-exist at the same time.

The primary coolant concentration associated with TS SR 3.4.17.2, i.e., the technical specification for iodines in the primary coolant, is calculated with the following equation:

C(i) x CT,,,.

DE I,,, (i) (pCi l gm) = C I ~ c n x ccol Where DCF(i)

= TID-14844 based Thyroid Dose Conversion Factor per Nuclide (remICi)

F(i)

= DCF(i) I DCF\\-ljl c(i)

= primary coolant equilibrium iodine concentration per nuclide (pCi1gm)

CTTOT

= primary coolant total (DE 1-1 31) TS iodine concentrations (pCiIgm)

DOC. NO.: 12400604-R-U-01 Rev 0 8 of 36 A

s a u SM7eBWNucieal

Prairie Island Nuclear Generating Plant The primary coolant concentration associated with TS SR 3.4.17.1, i.e., the technical specification for non-iodine fission products in the primary coolant is calculated with the following equation:

Non - Iodine Act(i) pCi l gm =

lOOx C(i)

[ ~ ( i )

x {E, (i) + E, (i)}]

E,(i)

= average beta energy per disintegration for isotope i (Mevldisintegration)

Ey(i)

= average gamma energy per disintegration for isotope i (Mevldisintegration)

C(i)

= primary coolant equilibrium non-iodine concentration per nuclide (pCi1gm) 100

= Normalization Factor [(~Cilgm)(MeV/disintegration)]

For PINGP, it was determined that the "failed fuel percentage1' associated with TS SR 3.4.17.2 (i.e., the iodines) was far more limiting relative to allowable concentrations in the primary coolant than that associated with TS SR 3.4.17.1 (non-iodines). Consequently, the estimated isotopic Technical Specification concentration of fission products in the primary coolant is based on the design basis fission product mix adjusted to reflect the limitations imposed by TS SR 3.4.17.2.

For the pre-accident iodine spike associated with the MSLB, CTw is 30 pCi1gm (transient Tech Spec limit for full power operation) or 60 times the proposed primary coolant total iodine technical specification concentration of 0.5 pCi1gm.

The accident generated iodine spike associated with the MSLB activities are based on 500 times the equilibrium iodine appearance rate.

The equilibrium appearance rates are conservatively calculated based on the technical specification primary coolant activities, along with the maximum design basis letdown rate, maximum technical specification primary coolant leakage, and an assumed ion-exchanger iodine removal efficiency of 100%.

Technical Specification Secondaw Coolant Activity The technical specification dose equivalent (DE) 1-131 concentrations per iodine nuclide in the secondary liquid are calculated using the same equation as provided earlier for the primary coolant. The secondary liquid technical specification concentration CTToT is 0.1 pCi1gm DE 1-1 31 per TS 3.7.14, where C(i) is the design basis secondary coolant equilibrium concentrations per nuclide.

The current MSLB analysis assumed that the relative mix of the secondary liquid is essentially the same as the primary coolant. This is a simplified assumption but is deemed non-conservative. The secondary coolant relative mix will not be the same as the primary coolant.

Due to the small allowable TS leakage (1 50 gpd1SG) of primary coolant into the large volume of secondary system liquid and the relatively small blowdown rate (30 gpm1SG) followed by cleanup via a mixed bed demineralizer, the mixture of iodines in the equilibrium secondary coolant will reflect the decay of shorter lived isotopes. Since 1-131 is the longest lived and most radiotoxic iodine isotope addressed in the technical specification, ignoring decay is non-conservative. Assuming that the secondary liquid activity consists of just 1-131 would be overly conservative. Consequently, the equilibrium secondary coolant isotopic concentrations are calculated by taking into consideration the primary coolant concentrations with 1 % fuel defects, the primary to secondary Technical Specification leak rate, the steam generator blowdown, the DOC. NO.: 12400604-R-U-01 Rev 0 9 of 36 Si\\iM S t a w & m N u c l e a r

Prairie Island Nuclear Generating Plant fraction of the blowdown flow that is treated by the mixed bed demineralizers and the efficiency of the demineralizers.

The primary and secondary coolant Technical Specification activity concentrations of noble gases and halogens are presented in Table 4.2-1. The pre-accident iodine spike concentrations and the concurrent iodine appearance rates are presented in Table 4.2-2.

5 ACCIDENT ATMOSPHERIC DISPERSION FACTORS (xIQ) 5.1 Offsite Atmospheric Dispersion Factors The Exclusion Area Boundary (EAB) and Low Population Zone (LPZ) atmospheric dispersion factors (xIQ) for PlNGP remain unchanged by this application and are consistent with current licensing basis.

The EAB & LPZ xIQ values applicable to PlNGP licensing basis are summarized in Table 5.1-1.

5.2 Control Room Atmospheric Dispersion Factors The Control Room air intake and center of Control Room ceiling (called center of control room hence forth) x/Q values for the release points applicable to the LOCA and the MSLB are calculated using "Atmospheric Relative CONcentrations in Building Wakes" (ARCON96) methodology (Ramsdell, 1997, Reference 3).

Shaw Stone & Webster has qualified the computer code ARCON96 for QA Category I use.

Input data consist of hourly on-site meteorological data; release characteristics such as release height, stack radius, stack exit velocity, and stack flow rate; the building area affecting the release; and various receptor parameters such as its distance and direction from the release to the control room air intake and intake height.

ARCON96 methodology has the ability to evaluate ground-level, vent, and elevated stack releases and treats building wake effects and stable plume meander effects when applicable.

This methodology is also able to evaluate area source releases, (as in the case of multiple vents spread over a roof top), using the virtual point source technique where initial values of the dispersion coefficients are assigned based on the size of the area source. The various averaging time period xIQ values are calculated directly from running averages of the hourly xIQ values.

As indicated before, as part of NRC SER for License Amendment 166 and 156 (Reference 4),

the staff performed a quality review of the on-site hourly met data (1993 through 1997) and concluded that the data was an acceptable basis for making estimates of atmospheric dispersion for design basis accidents. The above NRC-approved on-site hourly met data was utilized to develop the ARCON96 on-site atmospheric dispersion factors used in the LOCA and MSLB dose consequence analyses.

All releases are conservatively treated as ground-level as there are no releases at this site that are high enough to escape the aerodynamic effects of the plant buildings (i.e., 2.5 times Shield Building height, per Reference 24). In addition, the stacklvent release flows are not necessarily DOC. NO.: 12400604-R-U-01 Rev 0 10 of 36 1l slaw&bWs&Nudeer

Prairie Island Nuclear Generating Plant maintained throughout the accident period. The Shield Building area assumed to have an effect on the dispersion of the applicable releases is the portion above the Auxiliary Building roof.

The specific release-receptor combinations for which xIQ values are calculated are as follows:

1. Unit 1 and Unit 2 Shield Building Surface to the Unit 1 and Unit 2 Control Room Air lntakes and the Control Room Center

2. Unit 1 and Unit 2 Shield Building Vent to the Unit 1 and Unit 2 Control Room Air lntakes and the Control Room Center

3. Unit 1 and Unit 2 Refueling Water Storage Tank (RWST) Vent (i.e., the Auxiliary Make-up Air Intake) to the Unit 1 and Unit 2 Control Room Air lntakes and the Control Room Center

4. Unit 1 and Unit 2 Main Steam Safety ValvesISteam Generator Power Operated Relief Valve (MSSVs ISG PORVs) to the Unit 1 and Unit 2 Control Room Air lntakes and the Control Room Center (Diffuse Source)

5. Unit 1 and Unit 2 SG PORVs to the Unit 1 and Unit 2 Control Room Air lntakes and the Control Room Center (Point Source)

The following assumptions are made for these calculations:

The distances from the Unit 1 and Unit 2 Shield Building surfaces to the receptors are determined from the closest edge of the Shield Buildings (conservative)

The Shield Building area having an effect on the dispersion of the applicable releases is that portion above the Auxiliary Building roof The Shield Building surface releases are treated as diffuse vertical sources given the large horizontal and vertical surfaces from which the releases emanate The Shield Building Vent releases are treated as point sources The MSSVs ISG PORV releases are from the centroid of a rectangle encompassing the valves and are treated as a diffuse area source only when releases occur simultaneously from both the MSSVs and SG PORVs The SG PORV releases are from the centroid of a rectangle encompassing the valves and are treated as a point source when releases occur only from the SG PORVs The RWST releases are treated as diffuse vertical sources as they enter the environment via the Auxiliary Building make-up air intake louvers that are 9-ft wide and 10-ft high Initial diffusion coefficients for diffuse sources are based on the recommendations in Regulatory Guide 1.A94 (Reference 18)

The wind direction range (90 degrees), wind speed assigned to calm (0.5 mlsec),

surface roughness length (0.2 meter), and sector averaging constant (4.3) are taken from the recommendations in Regulatory Guide 1.I 94 DOC. NO.: 12400604-R-U-01 Rev 0 11 of 36 ShBMI " S M ~

& w

~udear

Prairie Island Nuclear Generating Plant All releases are conservatively treated as ground level releases as there are no release conditions that merit categorization as an elevated release (i.e., 2.5 times Shield Building height) at this site.

The plume centerline from each release is conservatively transported directly over the receptor For the control room centerline receptor, the elevation is chosen to conservatively minimize the distance between the release and the receptor Control Room Unfiltered In-leakage: the xIQ from the accident release point to the center of the Control Room ceiling is utilized for Control Room in-leakage since the above xIQ can be considered an average value for in-leakage locations around the Control Room envelope.

Control Room IngressIEgress: the xIQ from the accident release point to the center of the Control Room ceiling may be utilized for Control Room ingresslegress. The doors to the Control Room are located on the north side (i.e., Turbine Building side) of the Control Room as well as in the northeast and northwest corners. With the exception of the RWST Vents which are located east and west of the Control Room, all release points are located south of the Control Room. Therefore, the distances from these release points to the Control Room center are conservative (i.e., shorter) relative to the Control Room doors. Relative to releases from the RWST Vents, the distance from each of these release points to the center of the Control Room ceiling is reasonably representative of the average distance from each release point to the Control Room doors. Therefore, the xIQ to the center of the Control Room is considered reasonable for ingresslegress.

The bounding xIQ values (taking into consideration an accident at either unit) for the release-receptor combinations applicable to the LOCA and the MSLB are summarized in Table 5.2-1.

6 DOSE CALCULATION METHODOLOGY S&W proprietary computer program, PERC2, is used to calculate doses at the EAB, the LPZ, and the control room.

PERC2 is a multiple compartment activity transport code which utilizes an exact solution analytical computational process and addresses radionuclide progeny, time dependent releases, transport rates between regions, and deposition of radionuclide concentrations in sumps, walls and filters. The decay and daughter build-up during activity transport among compartments, and the various cleanup mechanisms are also included.

The code is configured to evaluate the radiological characteristics of nuclides released from a core and transported offsite and to the control room by way of a confinement and its auxiliary spaces (i.e., a series of regions), and via atmospheric dispersion. The program utilizes inputs which describe the release rates of radionuclides from the core, the path and rate of exchange DOC. NO.: 12400604-R-U-01 Rev 0 12 of 36

~hiiw" Sme.% bWx& ~uclear

Prairie Island Nuclear Generating Plant between regions, including filter efficiencies, deposition rates, atmospheric dispersion, breathing rates, dose conversion and occupancy factors.

The activity transport model first calculates the integrated activity at the location of interest (using a closed form integration solution) for each time interval. Using the integrated activity, the code then calculates the cumulative dose at the location of interest using the dose models and methodology as described below:

Dose Calculation Methodoloav Inhalation Dose -The dose conversion factors by internal organ type are applied to the activity in the air space of the control room, or at the EABILPZ. The exposure is adjusted by the appropriate respiration rate and occupancy factors for the control room dose at each integration interval as follows:

Dose (rem) to organ k from isotope j Integrated Activity (Ci-s/m3)

Inhalation dose conversion factors for organ k and isotope j (mremlpci)

Unit conversion of 1x1 ~ ' ~ p C i / C i C 3 Unit conversion of 1 x 1 0-3 remlmrem CB Breathing rate (m3/s)

CO Occupancy factor (for control room only)

Whole Bodv Gamma Dose - The gamma dose in the control room is based on a finite cloud model that addresses buildup and attenuation in air. The PERC2 gamma dose equation is based on the assumption that the dose point is at the center of a hemisphere of the same volume as the control room. The dose rate at that point is calculated as the sum of typical differential shell elements at a radius r. The equation utilizes the integrated activity in the control room air space, the photon energy release rates per energy group from activity airborne in the control room, and the ANSIIANS 6.1.I-1977 "neutron and gamma-ray flux-to-dose-rate factors" (Reference 15).

The whole body dose at the EAB and LPZ locations are calculated using, the semi-infinite cloud model outlined in Regulatory Guide 1.4 (Reference 16).

yDm (x,y,O) rad - 0.25 EYBAR w(x,Y,O)

EYBAR

=

average gamma energy released per disintegration (MeVIdis)

Y~(X,Y

,o)

=

concentration time integral (Ci-sec/m3) 0.25

=

[ 1. 1 1 x 1. 6 ~ 1 0 ~ ~ ~ 3. 7 ~ 1 0 ~ ~ ] / [ 1 2 9 3 x 1 0 O x 2 ]

where 1.I 1

=

ratio of electron densities per gm of tissue to per gm of air 1.6~1 0-6 (erglMeV)

=

number of ergs per MeV DOC. NO.: 12400604-R-U-01 Rev 0 13 of36 Sh;aRNe

~ l ~ l s

& m ~ u d e a r

Prairie Island Nuclear Generating Plant 3.7x101° (dislsec-Ci)

=

disintegration rate per curie 1293 (g/m3)

=

density of air at S.T.P.

100

- - ergs per gram per rad 2

=

factor for converting an infinite to a semi-infinite cloud Beta Dose - Similar to the whole body dose, the beta skin dose in the control room is calculated using the semi-infinite cloud model outlined in Regulatory Guide 1.4, and the integrated concentrations in the control room air space.

7 RADIOLOGICAL ACCIDENT REANALYSES As discussed in Section 1, as a result of a design verification effort initiated by NMC, and the associated corrective action process, the offsite and control room dose analyses for the following design basis accidents are being updated to reflect corrected design inputs, additional release pathways, and updated atmospheric dispersion and thyroid dose conversion factors developed consistent with current NRC recommendations.

1. Loss of Coolant Accident (LOCA)

2. Main Steam Line Break (MSLB)

The MSLB analysis also reflects a reduction in the primary coolant Technical Specification concentrations (from 1 pCi/gm DE 1-131 to 0.5 pCiIgm DE 1-131) which was required to maintain the current Technical Specification post-accident steam generator tube leakage of 1 gpm at 70°F in the faulted steam generator as a result of accident induced conditions, that was approved by NRC via License Amendment 133 and 125 (Reference 17). It is noted that plant operating data indicates that the actual primary coolant concentrations are well below the proposed reduced Technical Specification limit on primary coolant activity.

7.1 Control Room Design Information PlNGP is served by a single control room that supports both Units. The joint control room is serviced by two ventilation intakes, one assigned to Unit 1 and the other to Unit 2. These air intakes are utilized for the normal operation mode, as well as the accident mode prior to control room isolation.

During normal plant operation, the maximum total supply of unfiltered outside air to the control room is 2000 cfm (1818 scfm + 10%).

The common control room (with a calculated free volume of 61,315 ft3) is located in the Auxiliary Building at El. 735' and is equidistant from both units. The control room envelope includes the chiller rooms located directly above the control room at elevation 755' but does not include the cable spreading room located directly below the control room at Elevation 715', or the Operations Lounge, and Records Room located adjacent to the control room.

Note that operator occupancy and habitability determination is limited to Elevation 735' only.

DOC. NO.: 12400604-R-U-01 Rev 0 14 of36 n

SlhaMt sla7e&We7Nudear

Prairie Island Nuclear Generating Plant Since the PINGP control rooms are contained in a single control room envelope, they are modeled as a single region. Isotopic concentrations in areas outside the control room envelope are assumed to be comparable to the isotopic concentrations at the center of the control room.

To support development of bounding control room doses, the most limiting x/Q associated with fhe release point/receptor for an accident at either unit is utilized.

Prior to isolation, the control room post-accident ventilation model utilized in the dose analysis corresponds to an assumed "single intake" which utilizes the worst case atmospheric dispersion factor (xIQ) from release points associated with an accident occurring at either unit to the limiting control room intake. The atmospheric dispersion factors for the bounding release point 1 receptor applicable for radiological releases following a LOCA or a MSLB at Unit 1 and Unit 2 are provided in Section 5.

The current plant design will automatically isolate the control room and initiate control room filtered recirculation at 3600 cfm (4000 cfm - 10%) via the PINGP Control Room Emergency Ventilation System upon receipt of an SI signal from either unit, or a high radiation alarm from the control room in-duct radiation monitors. At PINGP, for specific operating configurations of the ventilation system, the high radiation signal is relied on as backup to the SI signal in the event of a single failure. The delay in crediting control room emergency isolation 1 filtered recirculation is assumed to be - 2 minutes based on receipt of a high radiation signal. It is expected that this delay is sufficient to address a Loss of Offsite Power (LOOP) that takes into account the delay associated with the diesel generator becoming fully operational (including sequencing delays), damper closure / re-alignment, and the time it takes for the emergency recirculation filtration fans to come up to speed. The control room recirculation filters are 99%

efficient for removing airborne particulates and 95% efficient for removing airborne organic and elemental iodine.

The dose model conservatively assumes that prior to achieving control room isolation the unfiltered intake flow into the control room is equivalent to the intake associated with normal operation, i.e. 2000 cfm. This is conservative since loss of normal ventilation would result in reducing the amount of contaminated air entering the control room via the intake prior to isolation.

A control room unfiltered inleakage of 175 cfm is assumed prior to and during the time it is isolated. The value for control room unfiltered inleakage is based on the results of tracer gas testing in the isolated mode of 165 cfrn, and includes a 10 cfrn unfiltered inleakage due to ingress / egress as recommended by NUREG-0800 SRP 6.4 (Reference 10)

Table 7.1-1 lists key assumptions / parameters associated with the control room design 7.2 Loss of Coolant Accident (LOCA)

PINGP has identified three activity release paths following a LOCA: (a) Containment Leakage, (b) ESF System Leakage, and c) Refueling Water Storage Tank back leakage. At PINGP, during power operations, other pathways between containment atmosphere and the environment (i.e., in-service containment purge, main containment purge, containment vacuum relief, etc.) are isolated during Modes 1 through 4.

DOC. NO.: 12400604-R-U-01 Rev 0 15 of 36 n

Shaw m & W @ N u c l e a r

Prairie Island Nuclear Generating Plant Table 7.2-1 lists some of the key assumptions I parameters utilized to develop the radiological consequences following a LOCA.

Doses due to Submersion and Inhalation SS&W proprietary computer program, PERC2, is used to calculate the Control Room and Offsite dose due to airborne radioactivity releases following a LOCA. PERC2 is a QA Category I code. It utilizes an exact solution analytical computational process that addresses radionuclide progeny, time dependent releases, transport rates between regions and deposition of radionuclide concentrations in sumps, walls and filters.

The inventory of fission products in the reactor core presented in Table 4.1-1 represent a conservative equilibrium reactor core inventory of dose significant isotopes, assuming maximum full power operation at 1.02 times the current licensed thermal power, and taking into consideration approved fuel enrichment and burnup.

Consistent with current licensing basis, and based on TID 14844 (Reference 5), the dose analysis assumes an instantaneous release of 100% of the core noble gases, 50% of the core halogens, and 1% of the core remainder, into the containment following a Loss-of-Coolant Accident.

Containment leakage Consistent with current licensing basis, the noble gases and halogens released from the fuel are assumed to mix instantaneously and homogeneously throughout the free air volume of the primary containment.

Fission product cleanup following a LOCA is accomplished by the Containment Spray System during the injection mode. At PINGP, containment spray does not operate in the recirculation mode.

The PINGP containment is surrounded by an annulus building (Shield Building). In the event of a postulated LOCA, the Shield Building Ventilation System (SBVS) is designed to collect and filter the containment leakage that has entered the Shield Building, and to exhaust the flow to the environment via the Shield Building vent stack.

It is expected that most of the containment leakage will be collected in the Shield Building; however, some of the containment leakage may by-pass the Shield Building and enter the Auxiliary Building where it is collected and filtered by the Auxiliary Building Special Ventilation System (ABSVS) and released to the environment via the Shield Building vent stack.

A small amount of containment leakage has the potential to bypass both leakage collection systems and leak directly to the atmosphere from the Shield Building wall.

Consistent with current licensing basis and in accordance with Regulatory Guide 1.4 (Reference l6), 100 percent of the core noble-gases and 25 percent of the core halogens (i.e., 50 percent of the core halogens that were released into the containment following a LOCA) are instantaneously and homogeneously mixed in the containment atmosphere, while the remaining 25% of the core halogens are assumed to be instantaneously plated out. In addition, per Regulatory Guide 1.4, the chemical form of the airborne iodine in containment after 50%

instantaneous plate-out is 91 % elemental, 5% particulate and 4% organic.

DOC. NO.: 12400604-R-U-01 Rev 0 16 of 36 ShaMI Stm B bW&W Nuclear

Prairie Island Nuclear Generating Plant Isotopic decay, containment leakage and spray removal are credited to deplete the inventory of noble gases and iodines airborne in containment. Containment sprays are initiated at T=42 sec (delay time addresses LOOP coincident with the LOCA). Consistent with original plant licensing basis, 100% of the containment is assumed to be covered by sprays. Also, since the long term sump water pH is controlled to greater than 7, iodine re-evolution is not considered.

Consistent with current licensing basis, the elemental iodine spray removal coefficient is 20 per hr. In addition and consistent with current licensing basis, the maximum allowable spray DF for elemental iodine removal is limited to 100 resulting in a spray cut-off time of 14.516 min after accident initiation.

The particulate iodine spray removal coefficient is developed in the updated analysis using NUREG 0800, SRP 6.5.2 methodology. (Reference 20) Per SRP 6.5.2, there is no limit on particulate iodine removal; however, the particulate iodine spray removal coefficient is reduced by a factor of 10 when the DF reaches 50.

Consistent with current licensing basis and based on plant technical specifications, the containment vessel is assumed to leak at 0.25 weight percent per day for the first day, and 0.125 weight percent per day for the remainder of the 30-day period. The Containment Vessel is assumed to leak to the environment via the following pathways: During the initial 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , of the total 0.25%/day containment leakage, 0.14%/day leaks into the Shield Building, O.lO%/day leaks into the Auxiliary Building Special Ventilation Zone, and O.Ol%/day leaks directly to the outside environment. These leak rates are consistent with the containment leak rate testing program requirements.

From 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to 30 days, of the total 0.125%lday containment leakage, 0.07%/day leaks to the Shield Building, 0.05%/day to the Auxiliary Building Special Ventilation Zone, and 0.005%/day directly to the outside environment.

Consistent with current licensing basis, prior to establishing negative pressure in the Annulus and Auxiliary Building Special Ventilation Zone, the containment leakage is assumed to be released directly to the environment without filtering. After the Shield Building Ventilation System and the Auxiliary Building Special Ventilation System have established a negative pressure, activity is assumed to be filtered prior to release.

The initiating signal for the SBVS and the ABSVS is the Safety Injection (SI) signal. Startup of the ABSVS results in the automatic termination of the normal operation Auxiliary Building Ventilation System and closure of the associated exhaust dampers.

It is determined by calculation that the normal operation Auxiliary Building Ventilation System exhaust dampers are closed prior to environmental release of any airborne activity in the Auxiliary Building due to containment leakage.

Containment Leakase that is collected in the annulus (Shield Buildina):

0 to 36 seconds Consistent with current licensing basis, immediately following the accident (time t=O), the Shield Building annulus pressure increases due to containment shell expansion and heat transfer from the containment shell to the annulus air.

The activity leaked from the containment into the annulus is assumed to be uniformly mixed in 50% of the Shield Building free volume. The Shield Building Ventilation System's two redundant exhaust fans DOC. NO.: 12400604-R-U-01 Rev 0 17 of36 n

ShillUY SlcneBbWsw Nuaear

Prairie Island Nuclear Generating Plant are not active during this time period. During this time period, no credit is taken for filtered exhaust from the Shield Building Ventilation System. It is assumed that 600 percent per day of the Shield Building volume is released directly to the atmosphere as a diffuse source from the Shield Building wall, without filtration.

36 seconds to 4.5 minutes Consistent with current licensing basis, one Shield Building Ventilation System fan begins to draw air from the Shield Building annulus at 36 seconds. Both filtered and unfiltered releases occur during this time period. After this time period all releases are filtered. 600 percent per day (1558.3 cfm) of the Shield Building volume is assumed to be released directly to the atmosphere without filtering. The Shield Building atmosphere is filtered 1 exhausted to the environment via the Shield Building Ventilation System at a rate of 6000 cfm.

4.5 mlnutes to 10 minutes Consistent with current licensing basis, after 4.5 minutes, a negative pressure of approximately -2 inches of water with respect to the atmosphere is achieved in the Shield Building annulus; thus direct out leakage from the Shield Building is assumed not to occur during this time period. It is conservatively assumed that no filtered recirculation takes place during this time interval.

Equilibrium exhaust flow to maintain the required negative pressure through the Shield Building is not established as yet, and the Shield Building atmosphere is filtered / exhausted to the environment via the Shield Building Ventilation System at a rate of 3000 cfm.

10 minutes to 20 minutes Consistent with current licensing basis, it is conservatively assumed that no filtered recirculation takes place during this time interval. Direct out leakage from the Shield Building is assumed not to occur during this time period since the Shield Building annulus is at a negative pressure with respect to the atmosphere. Equilibrium exhaust flow to maintain the required negative pressure through the Shield Building is not established as yet, and the Shield Building atmosphere is filtered 1 exhausted to the environment via the Shield Building Ventilation System at a rate of 1300 cfm.

20 minutes to 30 days Consistent with current licensing basis, equilibrium exhaust flow to maintain the required negative pressure through the Shield Building is established at 20 minutes. The Shield Building atmosphere is filtered / exhausted to the environment via the Shield Building Ventilation System at a rate of 1000 cfm. Credit for filtered recirculation (4000 cfm) within the annulus is taken during this interval.

Containment Leakage that is collected in the Auxiliary Building Special Ventilation Zone:

0 to 6 minutes The updated analysis recognizes that negative pressure with respect to atmosphere is achieved in the Auxiliary Building Special Ventilation Zone within 6 minutes after accident initiation. No credit is taken for filtration of releases via the ABSVS during this period.

During this time period, the containment leakage is assumed to be released directly to the environment from the Shield Building wall.

DOC. NO.: 12400604-R-U-01 Rev 0 18 of36 n

ShaYY s M w & m N u d e a r

Prairie Island Nuclear Generating Plant 6 minutes to 30 days Negative pressure with respect to atmosphere is achieved in the Auxiliary Building Special Ventilation Zone. Consistent with current licensing basis, during this period, although no mixing is credited, releases from the containment into the Auxiliary Building Special Ventilation Zone are filtered by the ABSVS before release via the Shield Building Stack.

SF Leakage Fifty percent of the core iodines and 1% of the core remainder are assumed to be homogenously mixed in the minimum post-LOCA sump water volume of 250,874 gals.

Equipment carrying sump fluids and located outside containment are postulated to leak at 1106 cclhr into the Auxiliary Building Special Ventilation Zone. This value represents one-half of the maximum (design) leakage possible from one operating system as documented in USAR Table 6.7.2. It is noted that the "as found" leakage is an order of magnitude less than that utilized in this analysis. ESF leakage is postulated to start at initiation of the recirculation mode which, at PlNGP is conservatively assumed to be at 16.25 minutes (assuming two train operation). Note that due to the long term nature of this release, minor variations in the start time of this release will not significantly impact the resultant doses.

The peak sump water temperature occurs at 6600 seconds and is 252.g°F. The fraction of total iodine in the liquid that becomes airborne is assumed to be equal to the fraction of the leakage that flashes to vapor. The flash fraction, associated with this temperature is calculated to be less than 10%. Consequently, 10% of the halogens associated with this leakage is assumed to become airborne and is exhausted (without mixing and without holdup) to the environment via the Shield Building Vent Stack after being processed by the ABSVS filters. In accordance with regulatory guidance, the 1% core remainder remains in solution and is not released to the environment.

RWS T Back-lea kage Sump water back-leakage into the RWST (tank is located in the Auxiliary Building) is postulated to occur at a rate of 5 gph and released directly to the environment via the Auxiliary Building Make-Up air intake louvers which is the closest opening in the Auxiliary Building located near the RWST. It is noted that a significant portion of the iodine associated with the RWST back-leakage is retained within the tank due to the equilibrium iodine distribution balance between the RWST gas and liquid phases which results in a time dependent iodine partition coefficient.

The assessment of the postulated iodine releases due to back-leakage of sump water into the RWST is similar to that performed by SS&W and accepted by NRC for other nuclear power plants, and takes into consideration the thermal hydraulic effects as well as the chemical equilibrium of iodine in the gas phase with either the incoming leakage or the iodine inventory contained in the residual RWST liquid, whichever is greater. Noble gases and their particulate daughters formed by decay of the halogen inventory in both the RWST liquid and gas phases collect in the gas phase. The RWST vent rate calculated at each time step determines the rate at which iodine and the isotopes formed by halogen decay are released from the RWST into the Auxiliary Building atmosphere via the RWST vent and is then dispersed into the environment.

To maximize the release rate, the temperature and relative humidity (RH) of the gas initially in the RWST, and of the air drawn into the RWST from the Auxiliary Building, were parametrically DOC. NO.: 12400604-R-U-01 Rev 0 19 of 36 n

~

u d

e a

Prairie Island Nuclear Generating Plant varied from 32OF to 120°F and 0% to 100% RH. The initial liquid temperature in the RWST was set at 120°F to be consistent with DBA LOCA initial conditions. The pH of the RWST liquid was also calculated at each time step to address the addition of boric acid, water and NaOH (the sump water buffering agent), to the RWST liquid. The pH of the RWST liquid, initially at 4.3, increased to 6.04 at 30 days. For all the cases studied, the gas phase iodine equilibrium concentration in contact with the incoming leakage was higher than that with the RWST liquid; this was due to the assumption that the sump liquid temperature of 165OF, that was reached at 1 E6 sec, was constant thereafter.

The RWST gas space vent rate is a maximum of -50% of its volume / day within the first 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months of the accident, -43% of its volume 1 day at 2 hrs, -23% of its volume / day at 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months ,

-1 1 % of its volume / day at 24 hrs, -4% of its volume / day at 96 hrs and -0.4% of its volume 1 day at 30 days. The maximum iodine release calculated during each time period was:

Fraction of Iodine Entering the RWST Vented to the Environment During Each Time Period Time Period 0 - 2 hrs 0.021 0%

2 - 8 hrs 0.01 93%

8 - 24 hrs 0.0129%

1 - 4 days 0.0095%

4 - 30 days 0.0033%

For purposes of the dose analyses, a conservative value of 0.1% of the iodine contained in the RWST back-leakage is assumed to be released to the environment and all of the daughter products formed by decay are assumed to be released.

For purposes of establishing dose consequences, a simplified conservative model is used which is very similar to the ESF leakage model. RWST back-leakage is expected to start at initiation of the recirculation mode which at PlNGP is conservatively assumed to be at 16.25 minutes. One tenth of a percent of the iodine in the back-leakage fluid is conservatively assumed to be airborne, released out of the RWST via the vent, and dispersed to the environment. Environmental airborne iodine activity resulting from RWST leakage is assumed to be elemental.

Control Room Dose due to Direct Shine from the External Cloud and Contained Sources:

The dose contribution in the Control Room due to direct shine from the external cloud and from contained sources is addressed. The external cloud contribution includes containment leakage, ESF leakage and RWST back-leakage. The contained sources include shine from the Containment Structure and the Shield Building, and the control room emergency ventilation filters. The remaining post-LOCA radiation sources (components carrying ESF fluids, the ABSVS and SBVS ventilation filters, and the RWST) are located such that there is sufficient distance or intervening shielding that would render the associated dose contribution at the Control Room negligible.

DOC. NO.: 12400604-R-U-01 Rev 0 20 of 36 A

ShaMI S r a w & W N u & a r

Prairie Island Nuclear Generating Plant SS&W Computer program PERC2 is used to calculate the radiation source term in post-LOCA airborne source in the containment and in the Shield Building, in the external plume passing the Control Room due to containment leakage, ESF leakage and RWST back-leakage, and accumulated in the control room emergency ventilation filters. SS&W Computer program SW-QADCGGP is used to calculate the dose inside Control Room by modeling the source-shield-detector configuration. PERC2 models used to calculate the external shine source term are the same as those used to calculate the Control Room inhalation and submersion doses. The flux to dose rate conversion factors used in the SW-QADCGGP calculations are from ANSIIANS 6.1.I-1 997 (Reference 15)

Dose Assessment The EAB, LPZ and Control Room dose following a LOCA is presented in Section 8.

7.3 Main Steam Line Break (MSLB)

SS&W proprietary computer program, PERC2, is used to calculate the Control Room and Offsite dose due to airborne radioactivity releases following a MSLB.

The MSLB assessment supports the implementation of Alternate Repair Criteria (ARC) as defined in USNRC GL 95-05 (Reference 22) and previously approved in PlNGP License Amendment Number 133 and 125. (Reference 17) In accordance with GL 95-05, the MSLB dose assessment utilizes the maximum allowable accident induced leakage that results in dose consequences that are just within the most limiting of the regulatory limits associated with the EAB, LPZ and the Control Room.

Note that subsequent to license amendment number 1331125, Unit 1 has implemented replacement steam generators and per TS 5.5.8, ARC is now applicable only to Unit 2.

However, usage of ARC methodology for Unit 1 remains conservative.

Thus ARC methodology is utilized herein for both Unit 1 and 2 As indicated previously, per this application, the specific activity for iodines in the primary coolant is being reduced from 1 pCiIgm DE 1-131 to 0.5 pCiIgm DE 1-131, to ensure that the dose consequences remain within regulatory limits with continued use of a 1 gpm (@ STP) post-accident leakage into the faulted steam generator as a result of accident induced conditions.

Table 7.3-1 lists some of the key assumptions 1 parameters utilized to develop the radiological consequences following the MSLB.

The updated radiological model used for the MSLB assessment conservatively assumes an almost immediate dry-out of the faulted SG following a MSLB (vs the previously assumed 15 minutes) resulting in a release of all of the contents of the steam generator. It is noted that for this release pathway, due to SG dryout, 107,100 Ibs of secondary liquid is released to the environment in 2 min. The initial concentration of iodine in the steam generator liquid is assumed to be at Tech Spec levels. The radioactivity concentration at the release point (i.e.,

the break point) is assumed to be that in the steam generator liquid after adjustment to reflect the change from liquid to vapor phase (i.e., assuming that at the release point, the release is saturated air at 212F and at atmospheric conditions of 14.696 psia). The analysis takes no DOC. NO.: 12400604-R-U-01 Rev 0 21 of 36 ShSEUV Stcne & Wr Nuclear

Prairie Island Nuclear Generating Plant credit for atmospheric dispersion. It conservatively assumes that the concentration entering the Control Room is the same as the steam concentration at the break exit point.

Consistent with current licensing basis, a simultaneous Loss of Offsite Power is assumed rendering the condenser unavailable, and environmental steam releases are postulated via the MSSVs I SG PORVs of the intact steam generator until shutdown cooling is initiated at T=8 hrs.

The elevated iodine activity in the primary coolant due to a postulated pre-accident or concurrent iodine spike, as well as the noble gas (at the proposed Technical Specification concentrations for the primary coolant), leak into the faulted and intact steam generators, and are released to the environment from the break point, and from the MSSVs 1 SG PORVs, respectively. It is noted that at PINGP, the ADVs are not utilized to cool the plant following the MSLB.

In accordance with current licensing basis, the steam releases from the intact steam generator continue until shutdown cooling is initiated via operation of the Residual Heat Removal (RHR)

System at T= 8 hrs, resulting in the termination of environmental releases via this pathway.

Additionally, and consistent with current licensing basis, the releases from the faulted SG due to primary to secondary leakage is also terminated at T=8 hrs after the accident.

In accordance with the guidance provided in GL 95-05, increased primary-to-secondary leakage (i.e., in addition to that allowed by the Technical Specification) is postulated to occur via pre-existing tube defects as a result of the rapid depressurization of the secondary side due to the MSLB and the consequent high differential pressure across the faulted steam generator. In accordance with the referenced guidance and current licensing basis, the MSLB dose analysis is performed to establish a maximum allowable accident-induced leakage, against which the cycle leakage projections can be compared.

Source terms Since there is no postulated fuel damage associated with this accident, the main radiation source is the activity in the primary and secondary coolant system. For the primary coolant, two spiking cases are addressed: a pre-incident iodine spike and a coincident iodine spike.

a)

Pre-incident spike - the initial primary coolant iodine activity is assumed to be 60 times the proposed Technical Specification Limit of 0.5 pCilgm DE 1-131 which is the transient Technical Specification limit for full power operation. The initial primary coolant noble gas activity is assumed to be at Technical Specification levels.

b)

Coincident spike - In accordance with current licensing basis, immediately following the accident the iodine appearance rate from the fuel to the primary coolant is assumed to increase to 500 times the equilibrium appearance rate corresponding to the Technical Specification coolant concentrations. The duration of the assumed spike is 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . The initial primary coolant noble gas activity is assumed to be at Tech Spec levels.

The secondary coolant iodine activity, just prior to the accident is assumed to be at the Technical Specification limit of 0.1 pCilgm DE 1-1 31.

The primary and secondary coolant source terms used to determine the dose from a postulated MSLB accident are discussed in Section 4.2.

DOC. NO.: 12400604-R-U-01 Rev 0 22 of 36 ShaMl m & W N u d e a r

Prairie Island Nuclear Generating Plant Release Path and Activity Transport Following a MSLB, primary and secondary coolant activity is released to the environment via two pathways, i.e., via the break point of the faulted SG, and via the MSSVs 1 SG PORVs of the intact SG.

Faulted Steam Generator The release from the faulted Steam Generator occurs via the postulated break point of the main steam line.

The faulted steam generator is conservatively assumed to dry-out almost instantaneously (- 2 minutes) following the MSLB, releasing all of the iodine in the secondary coolant that was initially contained in the steam generator. The secondary steam activity initially contained in the faulted steam generator is also released; however, this contribution is not included in this analysis since the associated radioactivity is insignificant compared to the other contributions.

In accordance with current licensing basis, the primary to secondary tube leakage in the faulted SG is assumed to increase from 150 gpd to 1 gpm (at 70°F) as a result of accident induced conditions. All iodine and noble gas activities in the referenced tube leakage are released directly to the environment without hold-up or decontamination. Per the current licensing basis, the primary to secondary leakage continues for 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . Release of the SG inventory (including primary-to-secondary leakage) is via steam line break point.

Intact Steam Generator The releases from the intact steam generator occur via the SG MSSVsI SG PORVs. The iodine activity in the intact SG liquid is released to the environment in proportion to the steaming rate and the partition coefficient. Steam releases from MSSVs ISG PORVs, terminate within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months of the DBA due to initiation of RHR. Per the plant Technical Specifications, the primary to secondary leakage in the intact SG is 150 gpd. Steam releases to the environment occur via the MSSVs & SG PORV during the cool down phase. The analysis assumes that the release point is either the MSSVs or the SG PORV, which ever has the worse atmospheric dispersion factor. It is noted that that the SG PORVs are located in the same area as the MSSVs.

The EAB, LPZ and Control Room dose following a MSLB is presented in Section 8.

8

SUMMARY

OF RESULTS AND CONCLUSIONS The accidents listed below have been re-analyzed for dose consequences at the exclusion area boundary, low population zone and Control Room.

1. Loss of Coolant Accident (LOCA)

2. Main Steam Line Break (MSLB)

The dose at the EAB during the first 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , and the dose at the LPZ "for the duration of the release" is presented in Table 8.1-1. These dose values represent the post accident dose to the public due to inhalation and submersion for each of these events. In accordance with current licensing basis, due to distancelplant shielding, the dose contribution at the EABILPZ DOC. NO.: 12400604-R-U-01 Rev 0 23 of 36 n

ShiMi" B ~(3bster~udear

due to direct shine from contained sources is considered negligible for all accidents. The associated regulatory limits as discussed in Section 2 are also presented.

Per regulatory guidance, the dose at the Control Room is integrated over 30 days. The calculated doses address the fact that for events with durations less than 30 days, the CR dose needs to include the remnant radioactivity within the CR envelope after the event has terminated. The 30-day integrated dose to the control room operator, due to inhalation and submersion, is presented in Table 8.1-2.

In accordance with current licensing basis, the Control Room shielding design is based on the LOCA which represents the worst case DBA relative to radioactivity releases. The dose contribution due to direct shine from post LOCA contained sourceslexternal cloud is identified and included in the Control Room doses reported for the LOCA.

It is concluded that the dose consequences at the EAB, LPZ and Control Room following a LOCA and MSLB remain within regulatory limits outlined in 10CFR1OO.ll and 10CFR 50, Appendix A, GDC 19.

DOC. NO.: 12400604-R-U-01 Rev 0 24 of 36 A

SjhaMT SMw &-Nuclear

Prairie Island Nuclear Generating Plant REFERENCES Code of Federal Regulations Title 10, Part 50.59, "Changes, Tests and Experiments" TID 24190, D.H. Slade, ed. Meteorology and Atomic Energy, 1968.

Ramsdell, J. V. Jr. and C. A. Simonen, "Atmospheric Relative Concentrations in Building Wakes". Prepared by Pacific Northwest Laboratory for the U.S. Nuclear Regulatory Commission, PNL-10521, NUREGICR-6331, Rev. 1, May 1997.

NRC SER issued September 10, 2004,

Subject:

Selective Implementation of Alternative Source Terms for Fuel Handling Accidents, Amendment No. 16611 56 (Accession Number ML042430504).

TID 14844, "Calculation of Distance Factors for Power and Test Reactor Sites", 1962 ICRP, "Limits for Intakes of Radionuclides by Workers, "ICRP Publication 30, 1979.

Federal Guidance Report 11, EPA-52011-88-020, Environmental Protection Agency, "

Limiting Values of Radionuclide Intake and Air Concentration and dose Conversion Factors for Inhalation, Submersion and Ingestion", 1988 Code of Federal Regulations Title 10, Part 100.1 1, "Determination of Exclusion Area, Low Population Zone and Population Center Distance" Code of Federal Regulations Title 10, Part 50, Appendix A, GDC 19, "Control Room" NUREG 0800, SRP 6.4, Revision 2, "Control Room Habitability System" NUREG-0737, Supplement 1, "Clarification of TMI Action Plan Requirements -

Requirements for Emergency Response Capability," December 17, 1982 NUREG 76-6521, "Radiation Signature Following the Hypothesized LOCA, Sandia National Labs (SAND76-0740), September 1977 DOEITIC-11026, Radioactive Decay Data Tables - A Handbook of Decay Data for Application to Radiation Dosimetry and Radiological Assessments", Kocher, David C.,

1981.

Lawrence Berkeley Laboratory, University of California, Berkeley, "Table of Isotopes,"

Seventh Edition.

ANSIIANS 6.1.l-1977, "Neutron and Gamma-ray Fluence-to-dose Factors" USNRC Regulatory Guide 1.4, "Assumptions Used for Evaluating the Potential Radiological Consequences of a Loss of Coolant Accident for Pressurized Water Reactors," Revision 2.

DOC. NO.: 12400604-R-U-01 Rev 0 25 of 36 ShilYY -&=Nudear

Prairie Island Nuclear Generating Plant NRC SER for Amendment No. 1331125, "Incorporation of Voltage Based Steam Generator Tube Repair Criteria", November 18, 1997 (Accession Number ML022260614).

Regulatory Guide 1.194, "Atmospheric Relative Concentrations for Control Room Radiological Habitability Assessments at Nuclear Power Plants", June 2003.

NRC Information Notice 91-56, "Potential Radioactive Leakage to Tank Vented to Atmosphere", September 19, 1991 NUREG-0800, 1988, Standard Review Plan, "Containment Spray as a Fission Product Cleanup System," Section 6.5.2, Revision 2.

NUREG-0800, SRP 15.6.5, Rev.1, July 1981, Loss-of-Coolant Accident resulting from Spectrum of Postulated Piping Breaks Within the Reactor Coolant Pressure Boundary" Appendix B, "Radiological Consequences of a Design Basis Loss-of-Coolant Accident:

Leakage from Engineered Safety Feature Components Outside Containment".

NRC Generic Letter 95-05, August 3, 1995, "Voltage Based Repair Criteria for Westinghouse Steam Generator Tubes Affected by Outside Diameter Stress Corrosion Cracking".

NUREG-0800, Standard Review Plan 15.1.5, Appendix A, Rev.2, "Radiological Consequences of Main Steam Line Failures Outside Containment of a PWR".

Regulatory Guide 1.145, "Atmospheric Dispersion Models for Potential Accident Consequence Assessments at Nuclear Power Plants", Revision 1 DOC. NO.: 12400604-R-U-01 Rev 0 26 of 36 6

SfiiaMf &m&WWNuclear

Prairie Island Nuclear Generating Plant TABLE 4.1-1 PINGP Equilibrium Core inventory (Power Level: 1683 MWt)

Nuclide AM-241 BA-1 39 BA-140 BR-83 BR-84 BR-85 CE-141 CE-143 CE-144 CM-242 CM-244 CS-134 CS-I 36 CS-137 CS-138 1-129 1-1 30 1-1 31 1-1 32 1-1 33 1-1 34 1-1 35 KR-85 KR-85M KR-87 KR-88 LA-? 40 LA-141 LA-1 42 MO-99 Activity (Curie) 6.54E+03 8.06E+07 7.73E+07 5.09E+06 9.08E+06 1.08E+07 7.39E+07 6.76E+07 6.25E+07 2.79E+06 6.1 3E+05 1.49E+07 3.1 2E+06 7.88E+06 8.46E+07 1.90E+00 1.45E+06 4.50E+07 6.51 E+07 9.1 3E+07 1.02E+08 8.72E+07 7.1 5E+05 1.08E+07 2.12E+07 2.82E+07 8.1 3E+07 7.26E+07 6.99E+07 8.31 E+07 Nuclide NB-95 ND-147 NP-239 PR-143 PR-144 PU-238 PU-239 PU-240 PU-241 RB-86 RB-88 RH-105 RU-103 RU-105 RU-106 SB-127 SB-129 SE-79 SR-89 SR-90 SR-91 SR-92 TC-99M TE-127 TE-127M TE-129 TE-129M TE-131 M TE-1 32 XE-131 M Activity iCurie)

Nuclide Activity (Curie)

DOC. NO.: 12400604-R-U-01 Rev 0 27 of 36 A --

& Wbs& ~ucleer

Prairie Island Nuclear Generating Plant TABLE 4.2-1 Primary and Secondary Coolant Technical Specification Iodine and Noble Gas Concentrations )

Primary Secondary Coolant Coolant Nuclide

@CiIgm)

(~Cilam)

Note 1: Primary and secondary coolant is assumed at 0.5 pCi/gm 1-131 DE and 0.1 pCilgm 1-131 DE, respectively TABLE 4.2-2 Primary Coolant Pre-Accident Iodine Spike Concentrations and Concurrent Iodine Appearance Rates Pre-Accident Iodine Spike Activity Concentration Nuclide

@CiIam)

Activity Appearance Rates (Cilsec)

DOC. NO.: 12400604-R-U-01 Rev 0

Prairie Island Nuclear Generating Plant TABLE 5.1-1 PINGP Offsite Atmospheric Dispersion Factors (sec/m3)

Exclusion Area Boundary Averaging Period Release Point All Releases Release Point All Releases Low Po~ulation Zone Averaging Period 0-2 hr 0-8 hr 8-24 hr 1-4 day 4-30 day 1.77E-04 1.77E-04 3.99E-05 7.12E-06 1.04E-06 DOC. NO.: 12400604-R-U-01 Rev 0 29 of 36 A

Shannr - sKne ti WB Nuclear

Prairie Island Nuclear Generating Plant TABLE 5.2-1 Control Room Atmospheric Dispersion Factors (sec/m3)

Averaging Period ReleaselRece~tor Combination 0-2 hr 2-8 hr 8-24 hr 4-30 d Shield Bldg WalllCR Air Intake 2.36E-03 Shield Bldg WallICR Inleakage 1.45E-03 1.13E-03 4.94E-04 3.78E-04 3.05E-04 Shield Bldg. Stack1 CR lnleakage 2.39E-03 1.98E-03 8.74E-04 6.12E-04 4.71 E-04 Aux. Bldg. M.U. Air Intake1 CR lnleakage 2.19E-03 1.89E-03 8.49E-04 5.95E-04 5.28E-04 SG PORVs & MSL Break Point/ CR Air 3.07E-02 Intake SG PORVs & MSL Break-Point1 CR 5.01 E-03 4.09E-03 lnleakage Note: PUFF Release of Secondary System TS Activity from the faulted SG following a MSLB assumes no credit for Atmospheric Dispersion DOC. NO.: 12400604-R-U-01 Rev 0 30 of 36 n

ShirVY" ~lcne

& Wixxf ~uctear

Prairie Island Nuclear Generating Plant TABLE 7.1-1 Analysis Assumptions & Key Parameter Values Control Room Control Room Parameters Free Volume Unfiltered Normal Operation Intake Emergency Intake Rate Emergency Recirculation Rate Emergency Intake Filter Efficiency Emergency Recirculation Filter Efficiency Normal/Accident Unfiltered Inleakage NormalIAccident IngressIEgress Occ~~pancy Factors Operator Breathing Rate Control Room emergency isolation Signal Delay in CR Isolation/Recirculation 61,315 ft3 1818 cfm + 10%

NA 4000 cfm f 10%

NA 99% (particulates) 95% (organic & elemental iodine) 165 cfm 10 cfm 0-24 hr: 1.O 1 - 4 d: 0.6 4-30 d: 0.4 0-30 d: 3.47E-04 m3/sec SI / Control Room Radiation Monitors 2 minutes (total)

DOC. NO.: 12400604-R-U-01 Rev 0 31 of 36 C1?

ShaMl Stcnesw81tJttaear

Prairie Island Nuclear Generating Plant TABLE 7.2-1 Analysis Assumptions & Key Parameter Values Loss of Coolant Accident Containment Leakage Parameters Power Level Free Volume Sprayed Fraction Spray flow rate Spray fall height Spray Period Mixing Rate Containment Leakrate (0-24 hr)

Containment Leakrate (1 -30 day)

Maximum DF for Elemental Iodine Sump/Recirculation Spray pH Fuel Activity Release Fractions Fuel Release Timing Chemical Form of Iodine Released Spray Removal Constants Core Activity

% Containment Leakage Released To:

1683 MWth 1.32Et-6 ft3 100 %

1200 gpm 150 ft 42 seconds to 16.25 minutes NA 0.25% volume fractions per day 0.125% volume fractions per day 100 2 7 Per Regulatory Guide 1.4 PUFF 5% elemental 9 1 % elemental 4% organic 20 h i ' (elemental) 5 hr" (particulate) 0 hi' (organic)

Table 4.1-1 T=O to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 0.14% /day (Shield Bldg.)

0.1% /day (Aux. Bldg.)

0.01% /day (Direct to Environ)

T=l day to 30 day 0.07% /day (Shield Bldg.)

0.05% /day (Aux. Bldg.)

0.005% /day (Direct to Environ)

Shield / Aux. Building Leakage Parameters Shield Building Free Volume 3.74E+5 ft3 Mixing Credited in Shield Bldg.

5 0%

Mixing Credited in Aux Bldg.

None ABSVS and SBVS Filter Efficiency 99% (particulates) 70% (organic & elemental iodine)

Period of time that the Shield Building is considered pressurized 0-36 seconds Shield Building Drawdown Period 36 seconds to 4.5 minutes DOC. NO.: 12400604-R-U-01 Rev 0 32 of 36 A

Sh;;wv stcne 8 WE#

Nuclear

Prairie Island Nuclear Generating Plant Auxiliary building drawdown period via the ABSVS

< 6 minutes Shield Building Leakage Rate 0 to 4.5 minutes 4.5 minutes to 30 days Shield Building Filtered Recirc. Rate 0 to 20 minutes 20 minutes to 30 days Shield Building Filtered Exhaust Rate 0 Sec to 36 Sec 36 Sec to 4.5 Min 4.5 Min to 10 Min 10 Min to 20 Min 20 Min to 30 Days Margin on all Ventilation Flows ECCS / RWST Leakage Parameters Sump Volume (minimum)

ECCS Leakage Rate RWST Back-Leakage Rate Leakage Period ECCS Iodine Release Fraction RWST Iodine Release Fraction Chemical Form of Iodine Released ECCS Release Point Fraction of ECCS that is Filtered RWST Release Point Fraction of RWST that is Filtered 600% /day or 1558.3 cfrn 0% /day 0 cfrn 4000 cfrn (Range 4500 - 5500 cfm) 0 cfrn 6000 cfrn 3000 cfrn 1300 cfrn 1000 cfrn 10%

250,874 gallons 1 106 cclhr 5 gph 16.25 min - 30 days 0.1 0.001 Assumed to be elemental Shield Building Vent Stack via ABSVS 100%

Aux. Bldg. Make-Up Air Damper 0%

CR Emergency Isolation1 Recirculation: Initiation SignallTiming Initiation time is 2 minutes (SI 1 CR Radiation Monitors)

DOC. NO.: 12400604-R-U-01 Rev 0 33 of 36 a

Shaw sw & W~EX Nudear

Prairie Island Nuclear Generating Plant TABLE 7.3-1 Analysis Assumptions & Key Parameter Values Main Steam Line Break Power Level Primary Coolant Mass Total Post-Accident Leakage to Affected SG Leakrate to Intact Steam Generator FailedIMelted Fuel Percentage RCS Tech Spec Iodine Concentration RCS Tech Spec Noble Gas Concentration RCS Equilibrium Iodine Appearance Rates Pre-Accident Iodine Spike Activity Accident Initiated Spike Appearance Rate Duration of Accident Initiated Spike Secondary System Release Parameters Iodine Species released to Environment Tech Spec Activity in SG liquid Iodine Partition Coefficient in Intact SG Fraction of Noble Gas Released from Intact SG Fraction of Iodine Released form Faulted SG Fraction of Noble Gas Released from faulted SG Post-Accident SG Liquid Mass per SG Steam Releases from Intact SG Dryout of Affected SG Termination of release: Faulted SG Termination of release: Intact SG Release Point: Faulted SG Release Point : Intact SG 1683 MWth 5290 ft3 @ at 560°F and 2235 psia 1 gpm @ STP 150 gpd @ STP 0%

Table 4.2-1 (0.5 pCi/gm DE-1131)

Table 4.2-1 (limiting of 1 OO/EBA~

and 0.5 pCi/gm DE-131)

Table 4.2-2 ( 0.5 pCi/gm DE-I13 1)

Table 4.2-2 (30pCilgm DE-I13 1) 500 times equilibrium 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months Assumed to be 100% elemental Table 4.2-1 (0.1 pCi/gm DE-113 1) 100 (all tubes submerged) 1.0 (Released to Environ without holdup) 1.0 (Released to Environ without holdup) 1.0 (Released to Environ without holdup) 107,100 Ibm (nominal at 100% power) 254,400 lbs (0-2 hr) 486,000 Ibs (2-8hours)

< 2 minutes 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 8 hours Portion of line between containment and MSIVs SG PORVs CR Emergency Isolation Recirculation: Initiation SignalITiming Initiation time (signal) 2 minutes (assumed)

DOC. NO.: 12400604-R-U-01 Rev 0 34 of 36 A

S ~ W sw B WWC-Z Nuclear

Prairie Island Nuclear Generating Plant TABLE 8.1-1 Exclusion Area Boundary and Low Population Doses EAB Dose (rem)

LPZ Dose (rem)

Accident LOCA Main Steam Line ~reak"'

Note 1 : Concurrent Iodine spike: CIS Pre-Accident Iodine spike: PIS Thyroid 20.5 6.4 (CIS) 3.9 (PIS)

Accident DOC. NO.: 12400604-R-U-01 Rev 0 Whole Body 2

c 0.1 (CIS)

< 0.01 (PIS)

Thyroid Regulatory Limit 300 (Thyroid) 25 (Whole Body) 30 (Thyroid) 2.5 (Whole Body) 300 (Thyroid) 25 (Whole Body)

Whole Body LOCA Main Steam Line Break "'

Regulatory Limit 2

< 0.1 (CIS)

< 0.01 (PIS) 16 22.4 (CIS) 3.3 (PIS) 300 (Thyroid) 25 (Whole Body) 30 (Thyroid) 2.5 (Whole Body) 300 (Thyroid) 25 (Whole Body)

Prairie Island Nuclear Generating Plant TABLE 8.1-2 30 Day Integrated Control Room Operator Dose (rem) 'l' NOTES:

I GDC 19 Limits as clarified by NUREG 0800, SRP 6.4:

30 rem thyroid 5 rem whole body 30 rem beta skin Accident LOCA Main Steam Line Break DOC. NO.: 12400604-R-U-01 Rev 0 Whole Body 1.5

< 0.01 (CIS)

< 0.01 (PIS)

Thyroid 28.5 27.2 (CIS) 6.9 (PIS)

Beta 16.5

< 0.1 (CIS)

< 0.1 (PIS)

v t e Letter Sequence Request
TAC:MD9140, (Approved, Closed)
Initiation Request, Request, Request Acceptance... Supplement, Supplement Administration Questions 27 April 2009 26 January 2009 Answers Results Approval, Approval Other:
MONTHYEAR2009-01-26 [Table View]

L-PI-08-052, License Amendment Request (LAR) to Revise the Loss of Coolant Accident (LOCA) and Main Steam Line Break (MSLB) Accident Dose Consequences Analyses and Affected Technical Specifications (TS)

Text

Enclosures:

SUMMARY

2.2 Background

SUMMARY

2.2 Background

SUMMARY

SUMMARY

Subject:

Navigation menu

L-PI-08-052, License Amendment Request (LAR) to Revise the Loss of Coolant Accident (LOCA) and Main Steam Line Break (MSLB) Accident Dose Consequences Analyses and Affected Technical Specifications (TS)

Text

Enclosures:

SUMMARY

2.2 Background

SUMMARY

2.2 Background

SUMMARY

SUMMARY

Subject:

Navigation menu

Search